ML20148T015

Requests Commission Grant Licensee Exemption from Early Warning Alert Sys Requirements.Encl Sets Forth Position Re Adequate Protection for Public in Evacuation Plume Zone

ML20148T015
Person / Time
Site:	Fort Saint Vrain
Issue date:	01/09/1981
From:	Millen C PUBLIC SERVICE CO. OF COLORADO
To:	Ahearne J NRC COMMISSION (OCM)
References
P-81009, NUDOCS 8102130005
Download: ML20148T015 (4)

Text

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2053

SUBJECT:

Fort St. Vrain Unit No. 1 Emergency Planning Early Warning Alert System

Dear Dr. Ahearne:

In our letter of April 1,1980, to Mr. Brian Grimes (P-80066) which is included as Attachment I,

we set forth our evaluation, and justification for recucing the EPZ and Ingestion Pathway distances from 10 and 50 miles respectively, to 5 and 30 miles.

Our l

justification was accepted and the EPZ and Ingestion Pathway was subsequently recuced along with those of LWR's 250 MW and smaller.

In our transmittal of our emergency response plan on March 18, 1980, we took exception to the early warning alert system requirement on the casis that Attachment I clearly indicates that for a LOFC accident (DSA #1) at Fort St. Vrain we do not exceed the Protective Action Guidelines (PAG's) at the exclusion area boundary for at least twelve (12) hours and do not exceed the PAG's at one (1) mile for at least twenty (20) hours.

These studies were made utili::ing a persistent plume trajectory with Pasquill Category F meteorological conditions, and conservative source term estimates.

(See excerpt of our resconse from P-80083 in Attacnment II.)

Following the initial submittal of nur emergency response plan, we suomitted our revised emergency response plan on August 23, 1980, and again we referenced the technical justification for eliminating the early warning alert system.

(See excerpt of response from P-80288 in Attachment III.)

On December 10, 1980, we met with Mr. Brian Grimes along with other members of the emergency preparedness group and NRR staff in an attempt to.*esolve some of the major issues of our emergency response plan as well as attemot to justify the major differences in HTGR Tecnnology versus LWR tecnnology.

We were unable to resolve any of the majo-issues at the December 10, 1980, meeting, primarily cue to tne fact that our transmittal letter (P-3028E) was apparently misplaced and never received a staff review.

From all indications,

however, it was apparent t hi.: Brian Grime's group was not in a position to provide relief from the early warning alert system.

Recogni::ing that our initial evaluations ( Attachment I) were based on extremely conservative source term estimates, we performed an 810 218 0CCS9 gjSdG Y

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l additional evaluation (see Attachment IV). While we removed some of the extreme conservatism of our initial EPZ study we. have still i

maintained a considerable factor of conservatism in our alternate-i scenario.

Based on this alternate scenario,.the'following conclusion can be summarized:

l 1.

The whole body gamma dose cumulative over a thirty (30) day I

period for an LOFC accident at Fort St. Vrain never exceeds i

the PAG for shelter protection and never approaches the PAG l

for mandatory evacuation, i

2.

The maximum whole body gamma dose for a 30 day period at the exclusion area is 800 millirem with the conservative meteorological plume trajectory.

(See Figure 2 of Attachment IV.)

j 3.

The most restrictive PAG, Thyroid Inhalation, is not I

exceeced at the exclusion area boundary-for fi fty (50) i

hours, (see Figure 3 of Attachment IV), and the PAG for mandatory evacuation is never exceeded.

4.

None of the PAG's for the closest population center, Platteville, is exceeded or even approached.

On the basis of the conservative estimates represented by either l

Attachment I or Attachment IV, more than adequate time exists to warn i

the general public and institute protective action.

Most certainly a 15 minute time limit cannot be justified, in light of the potential doses to the general public.

I Based on the characteristics of Fort St. Vrain and the comparison of time for accidents to develop, we cannot see the justification for i

imposing requirements developed for LWR Technology upon Fort St.

l Vrain. We have been_ unable, to date, to obtain any relief based on j

the technical justification provided, and apparently the emergency

}

preparedness ~ group is not in a position to judge the requirements j

based on technical merit.

i Si-we have been unable to obtain any definitive guidance from the l

review groups, we are appealing to the Commission for consideration of relief from the early warning. alert system requirements, and i

vaintain that our position set forth in Attachment III provides adequate protection'for the public in the EPZ.

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Because of the significant delays we have already experienced, in obtaining some definitive guidance, prompt consideration of 'his appeal will be greatly appreciated.

Should you requi-; any additional information, please contact Mr. Don Waremboure, Manager, Nuclear Production, (303) 571-7436, Extension 200.

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i Very truly yours, M+

W,nh.

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C. K. Millen Senior Vice President CKM/dkm

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PLATTEVILLE, COLOR ADO 80651 April 1, 1980 For: S:. Vrain Uni: No. 1 P-80066 Mr. Brian K. Gri=es Director, E=ergency Preparedness Task Group Office of Nuclear Regulation U. S. Nuclear Regula ory Co==1ssion Washing:en D.C.

20555

SUBJECT:

For: St. Vrain Uni: No. 1 E=ergency Planning RITERENCES:

1) P-79205 2)

P-80011

Dear Mr. Grimes:

In Referen:a (1), ?-79205, we expressed our concern with :he ex:en-sion of the EP :o a ten (10) =11e radius and :he inges:1on pa:hway

o a fif:7 (50) =ile radius. We did not receive any acknowledge-

=ent of this letter, however, on January 8, 1980, we =e wi:h Messrs.

Roe, Ku:=yc:, and Willia =s in Be:hesda :o discuss our concerns. As a resul: of this =ee:ing we agreed :o evalca:e :he I?: for For: S:.

Vrain with reference to :he guidelines of :TERIG-0396 to provide fur-

her justifica:icn for s= aller E?Z and inges: ion pathway radii for For: S:. Vrain.

We have ce=pleted our evaluation, and we are ::ans=1::ing herew1:h five (5) copies of :his evaluation for your review.

The attached evaluation is self enplanatory and fully supports our original con-l encion :ha: :he I?Z for For: S:. Vrain can be considerably smaller than that for a 1000 M'a'e LWR which was utili:ed as basis f or devel: ping the guidelines of NUREG 0396.

As a brief st==ary our evalua:1on indica:es : hat the co= parable I?:

f: Fort C:. 7:21: wou d be abou: five (3h =iles and a c:= parable in-gestion pa:hway would be abcut thir:7 (30) ni;es.

The evaluatice is based on c:nserva:ive es:ima:es of source :er= releases, wors: case

=e:eorological condi: ions and ex:re=ely conservative PCR7 (Pres:ressed Concre:e i ::.or Vessel) leakage ra:es.

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P D V i)( bF Q Q ( l [;,( D "s.? o!

Mr. Brian K. Grimes April 1, 1980 On the basis of our evaluation it is requested that :he Fort St. Vrain EPZ and ingestion pa:hway be established as five (f) and thirty (30)

=iles, respec:ivsly.

It should be noted fur:her that we have beep hrough our bases wi:h State officials, and we have cumplece a cr.uust wi:h :he Sca:e con-s cerning the inherent safety of For: St. Vrain versus a LWR. We have also discussed our position with the local FEMA representatives, and again have concurrence that there is no justification for penali:ing For: St. Vrain with water reactor criteria.

I: is requested that you give this =acter your i=nediate attention.. As you are aware we are in the process of finali:ing our Emergency Response Plan with the State as well as the emergency';1anning for station. Our planning is presently based on :he smaller I?! and 1.nges:1on pathway exposure radii and any changes in :his criteria could have a significer.t i= pac: on final acceptance of our plans, both for :he s a: ion and :he State.

Very truly yours, w/7 4

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Don Warenbourg Manager, Nuclear ?roduction Fort St. Vrain Nuclear Generating.S:s ion.

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EVALUATION OF EMERGENCY PLANNING :',0NE. DISTANCES APPLICABLE TO THE FORT ST. VRAIN NUCLEAR GENERATING STATION l

.j 1.

Introduction

]

sin an'NRC 1stter.to all reactor licensees dated November 29, 1979, a t

request vas made for information regarding estimates for, evacuation of, various,_

areas around nuclear power reactors. NUREG-0396, IPA 520/1-78-016 specified, based upon' studies of a 1000.W(e) LWR, that the emergency planning zone (IPZ) for plume exposure pathway should be about.10 mile radius aoout the plant and 1

about 50 mile radiun for the ingestion exposure pathway. Both design basis accidents with various active engineered safety features, and the accident release categories of the Reactor Safety Study (Wash 1400) were considered in determination of the specified distances. A factor in establishing the about 50 mile rad 1us ingestion exposure pathway is that distance has been evaluated

' i and discussed in detail in Chapter 2 of Facilities Safety Analy;is Reports.and in Environmental Reports.

The purpose of the evaluation presented herein is to determine the appropriate EPZ distances for Fort St. Vrain which provide the general public equivalent pro-taction that would be afforded by the 10 mile and 50 mile EPZs for a similar sited 1000.W(e) LWR. At the onset of the investigation it was believed the principle differences would be a smaller source term d.na to site, 330 MW(e) at 39% efficiency versus 1000 MW(e) LWR, and time to perfoca energency actions due to the relatively slow heacup rate provided by the large graphite core of an HTGR.

It was confirned by the detailed study that the FSV smaller source term coupled with unique slov and

gradual core heacup characteristic of the plant, did in fact justify sharply re-duced EP: distances for Fort St. Vrain and significantly longer eines to perforn offsite emergency actions.

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a 2.

Conduct of Study This section describes the. methodology, parameter assignments and assump-tions used in the conduct of this EPZ study.

Section 2.1 describes the accident scenario used as the basis for FSV EPZ determination as well as the description of what we.1 assumed to be a comparable accident scenario for the LWR.

Section 2.2 describes the conservatis=s applied to the accident scenario.

Section 3 l

describes the conservanium.mateorlogical conditions assu=ed to prevail during l

the duration of the EPZ determining event. Dose computation =echodology and parameter assignments and assumptions appear in Sections 4 and 5 respectively.

I The appropriate E?Z distances for FSV are presented in Section 6.

Conclusions

[

t of F.his study are presented in Section 7.

t 2.1 Accident Scenario Used to Establish the I?Z for FSV The 'FSV Design 3 asis Accident #1 (DBA #1) which was the only accident iden-i tified and analyted in the ?SAR that results in extensive core damage, was selected i

as the representative scenario used to establish the plant's EPZ.

This event is postulated to be initiated by a "non-mechanistic" per=anent loss of forced cir-culation while operating at full pewer.

The reactor is scram =ed by the plant

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protection system and all atta= pts to restore forced circulation using the i

multiple heat sinks, circulators, and motive power for the circulators fail. When f

it becomes apparent to the plant operators that the loss of forced circulation t

i is permanent, e.g. when restoration of cooling would cause steam generator damage, the pri=sry coolant system would be depressurized to storage in a controlled sanner through the heliu= purifiestion syste=.

The reserve shutdown system would i

i be operated after this initial period to assure an adecuate shutdown margin.

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s 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 de-pressuri:ation 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 a= ended by PSC letter F-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 redue:1on in time was due to the re-evaluation of the capability of the helium purification system to process primary coolant during the planned depressurization and release of the clean primary coolant to the reactor building ventilatica stack. Thus, the de-pressurization of the PCRV is now initiated af ter 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).

The fuel is slow to heat up due to the large heat sink provided by the core graphite. A peak average active core te=perature of 5400 7 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 vaporization tempera:ure of graphite is 6900 F.

Peak ac:1vi:y 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.

PCRV conta1= ment in:egrity is =aintained through heat removal by the liner cooling system in the " redistribution = ode" which -avd-4 es cooling in the top head of the PCKV.

Leakage of primary coolant from the PCRV is assumed :o occur at a conser-vatively high leakage ra:e of 0.2" of the pri=ary coolant inventory per day.

The reactor building ventilation system main:ains continuous venting and processing of the reac:or building environ =ent at 1.5 building volunes/hr during

he entire period of the accident.

-5 9000 857.5

-7 9000 2

2 W = 1.13 x 10 x

9'

.5

- 12.5 )

10 12.5 10

= 0.43 + 602 = 600 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 in transcribing from equation 14 to 15 since equr.cion 13 has the 9.1 x 10 coefficient.-

Equation 15 should read:

-5 9000 857.5

-6 2

2 W = 1.13 x 10 x

1"

+

x 0 9000 (857.5 - 12.5 )

10 12.5 10

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

The second item is that the i?/iPo term has been dropped in going from egn. 14 to egn. 15, which is significant if it is assu=ed that these equations are appropriate for evaluating the leak rate at F

=,5 psig. Including this f acter, 1

we find:

LEAK RATI ressure ?

1 lb /dav

%/dav (psig) qn 14 15 15 Revised 14 15 15 Revised Given App D; A=end 9 Question D.2 5

.0019

.13

.30

.001

.07

.17

.20 Amend 9 Questien D.2 2

.0003

.046

.107

.0001

.025

.059

.08

Since equation 14 is the appropriate equation, the 0.2*/ day leak race is conservative by a factor of 200.

For purposes of this evaluation, the historic 0.2 / day is assumed to exist as an upper limit of all potential contaminated primary coolant leakage ir.cluding permeability through the PCRV concrete.

This is judged to be conservative since the primary coolant with any.significant activity is contained witnin the PCRV or helium purification components contained in walls within the PCRV.

For this study, the PC3V leak rasu of 0.2*/ day is conservatively assumed to apply for the total accident duration period beginning at t = 0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />.

In l

actuality, this arbitrarily conservative estimate of PCRV. leakage would not apply during the period of controlled PCRV depressuri:ation or in the subsequent post depressuri:ed period prior to an assumed liner failure. However, for con-servatism this PCRV leak rate will be applied uniformly for all time.

2.2.2 Radionuclide Source Terms for DBA-1:

i As previously stated, the fuel vithin the graphite core is slow to heatup 1

during 3BA#1. After about 7.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />, some portions of the slowly heating core will have reached the FSAR fuel particle coating failure temperature of 1725 C (3137 ?).

Once reaching the FSAR' fuel particle coating failure temperature, the fission products are assumed, for purposes of this evaluation, to be released in a time dependent fashion normalized per the TID-14844 core release assu=ptions.

yor release to the primary coolant within the PCRV, the TID-14844 conditions assu=e a release of 100" of noble gases, 50* of the iodines and 1" others.

Consistent with TTD-14844 release assu=ptions, 50" of the iodines plateout within the. primary coolant system resulting in a depletion of the iodir.e to 25" of core inventory in the resultant PCRV leakage to the teactor building atmosphere.

O'

.Th s3 TID-14844 ralsacco and pistecut assumptions are vary conscrvativa for'an RTCR but never the less have been utilized in this evaluation. This' l-

. conservatism in release and plateout assumptions is particularly evident when i

the halogen release fractions estimated in the FSV FSAR (Appendix D) are compared to TID-14844 values. The FSAR estimates a core halogen available leakage frac:fon gi of 5.5% which is a factor of 4.5 below the !!D-14844 effective core release fraction of 25%. The FSAR fuel particle coating failure temperature of 1725 C is also conservative compared to test data simulacing a core heacup accident.

Figure 2-1 i

illustrates the margin from the test data to the 1725 0 f ailure temperature used in this evaluation. Also shown is a NRC model (NUREG-0111) for. TRICO UC E***i"1" 2

which is similar to the FSV FSAR model.

l 3.

Meteorologv r

3.1 Site Seecific Meteoroloey The Fort St. Vrain site meteorology was described in deta'l in the FSV FSAR Section 2.8.

The most likely condition on an annual basis (Table IVII of FSAR) are vinds from the N sector of the wind rose which occur 17% of the time. 'The most likely stability category is D (Table IIIV of FSAR), and the i

average wind speed at 200 ft. elevation averaged over all direc:1ons is 6.51 sph (2.9 m/s) as given in Table II of the FSAR.

Figure 17 in See:1on 2.8.3 of the FSAR gives the probability that the dilution factor will exceed a given value.

Using the dashed curve shown in Fig. 17 (FSAR Sec. 2.8.3) which includes :he i

effect of using corrected ane=ometer wind veloci:y indicates that a dilution factor

-3 of about 3.5x10 s/s, at 590= will be exceeded only 5% of :he time (the 5%

cu:-off). This correc:ed cut-off corresponds to stabill:y categery F with a I

wind speed of about 2.4 m/s. Thus the dilu:1on factor for stab 111:7 category F with 1 m/s wind speed will be exceeded considerably less :han 5% of the :i=e and represents a " worst case" type condition.

3.2 NRC Staff Meteorologv from SER l

The Fort St. Vrain Safety

• Evaluation Report (SER) (Ref. 4) prepared by the than AEC staff includes the following meteorology and breaching rate assumptions for accident condition calculations at the 1.?Z (16,000m) l 3

Time Period Dilution Factor (s/m )

Breathing Race ( )

-6 Most Unfavorable 8 hrs.

7.8x10 3.47x10 Next Most' Unfavorable

-6 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> 3.9x10 1.75x10

~0

~4 '

1 to 4 days 2.4x10 2.32x10

~7 4 to 30 days 7.8x10 2.32x10 30 to 180 days 4.6x10 2.32x10"'

~

Plotting these dilution factor values against the time period over which they represent an average value (71g. 3-1), it appears that there is approximately k

l at time dependence, as would be expected, when allowing for wind meander and change of direction. This can be also be seen from the equations for averaging i

4 o

• 03 the dilu: ion factor for wind meander over a 22.5 sector ($ = M for ground c,x u i

level release) and for.an annual average, (Een. 6, Sec. 2.8.3 of FSAR) both of which depend inversely on e, ( the vertical diffusion parameter).

Since at long

imes the diffusion process is dominated by random walk effects, c2 is erpec:ed a

k to be proportienal to :., and so $ wot'id be proportional to :

3.3 Extention of SRC Staff Meteorologv to other Distances To derive the ti=e dependence of the dilu: ion fac:or at other distances,

he :i=e dependence of :he NRC staff neceorology was esti=a:ed from a curve dra.m through :he 16,000m points frc= :he SER as shown in Fig. 3-1.

This gave a

~*

dependence.

~he reference poin: used for the other dis:ances was :he 8 hr. value, which was calculated from o(Shr) =, o e _ with C = 1 m/s and u

73 e: for all distances :aken for stability category F.

This is equivalent :o

?^

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l,

.taking the short ter= dilution factor for ground level release, 1

I naa G ya 7

1 2

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as the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> value, and then extrapolating to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> by the factor \\g = 0.5.

l Table 3-1 gives the values at other ti=es obtained from 9(t) =

(8)

}0.493 j

• /

where c is in hours. The breathing rates used were taken fro = the SER and are a

given above in 3.2.

3.4 Conservative Meteorologv Assumptions i

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Listed below are the conservatis=s used in obtaining the dilution factor in 1

Table 3-1.

a)

Worst case type =eteorological conditions (Category F, G = 1 =/s) used are likely to ec ur =uch less than 5% of the time, b)

Downwash conditions are assu=ed to persist over the entire ti=e.

c)

The effective stack height is taken to be cero (i.e., ground level release).

d) 3euyancy affects of He relative to air are neglected.

Ii e)

Dilution by the building wake is neglected.

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4.

Dose Coccutation Methodolorv-i e

Plu=a exposure doses were e.alculated using the ce=puter progra= IDAC (Ref.1).

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TDAC is a computer progra= which utilites analytical =ethods to calculate the ti=e-I dependent radiological i= pact due to rouette oc accidental release of radionuclides from nuclear power plants or other potential sources of radioactivity release, j

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I The basic odel used by TDAC is comprised of eight volu=es which are in-1 terrelated by a network of flow paths.

The progra= calculates the a=ount of activity in each volu=e at specified ti=es and deter =ines radiological doses are I

specified distances based on activity entering volu=e 8 and specified neteorology.

The calculated doses are external whole body ga==a, external whole body beta, and 1

1 the internal inhalation com=1t=ent doses to the sxeletal syste=, thyroid, lung, gastro-intestinal tract and total body. The recuired input da:a corsists of

I three types:

(1) flow model data such as flow rates and filter efficiencies, (2) nuclear parameters such as decay constants, branching ratios, and disintegration energies, and (3) radiological dose input such as vindspeed, at=ospheric dilution Sinters, distances, and breathing races.

The calculation of activity in each volume is based on the assumption of instantaneous homogeneous mixing.

The calculation of radiological doses is based on the semi-infinite cloud approximation. All calculations are based on the analytical solution of the coupled linear differential equations governing the activity in the different volumes.

I The TDAC code, with appropriate parameter assign =ents was also used to evaluate the plume exposure doses for the Reference L*JR of this study.

For this study, it was obsetved that the dose categories of whole body l

gamma external plu=e exposure and thyroid inhalation and ingestion dose control and for this reason other organ dose categories are omitted fro = chis presentation.

5.

Parameter Assien=ents and Assu=rtions Table 5-1 provides. a listing and su==ary of the parameter values and analysis assumptions of this study. The character 4.stics of the FS7 siting event, the DBA #1, and'the reference L*JR event, DBA/LOCA are described.

These events are felt to bound the envelope of emergency preparedness planning.

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6.

-Apo11 cable EPZ Distances for Fort St. Vrain 6.1 Evaluation of Evacuation IPZ for FSV The'EPZ (Emergency Planning Zone) determination for Fort St. Vrain is based-

.on the estimated accident doses in the environment and their comparison to the re-l comended Protective-Action-Guides-(?AGs) contained in Ref, 3, Table 5.2.

1 i

These PAGs for the general _ population consist of dose ranges in the event of an emergency condition, along with recomended actions to be taken by the.

l responsible local authorities.

Figures 6-1 and 6-2 show the comparison between FSV and a reference 1000 ST(e)

LWR of the cumulative dose vs. distance. Also shown on Figure 6-1 are the 1 and 5 Rem whole body gamma PAG dose levels.

Salow 1 Ram, no protective action is required according to the PAG, while the state may issue an advisory, and environmental radiation levels are monitored.

For doses of 1 to less than 5 Ram, the general population should seek shelter and avait further instructions, evacuation of particularly susceptible elements of the general population should be considered, access should be controlled, and environmental radiation levels monitored.

For 5 Ram or higher whole body gam a doses, =andatory evacuation of the population in predetermined areas is recomended.

Figure 6-2 is similar to Figure 6-1 except that it is for inhalation thyroid dose and the PAG 1evels are 5 and 25 Rem.

Clearly, thyroid dose is the controlling factor since the distance to which protective action is required is larger.

From Fig. 6-2, the O to 30 day inhalation thyroid dose for 757 of 25 Rem occurs at about 4100m (2.5 mi.).

The above data is also su=arited in Table 6-1.

1 For a 10 mi. !.WR evacuation radius (16,000=), the dose for the reference 1WR is about 11 Rem inhalation thyroid. The distance fer FSV for the sa=e dose as the re:erencil LWR at 10 mi. is about 7500m (4.7 mi.).

Therefore an evacuation radius of about 5 mi. for FSV would afford a comparable degree of public protection at j

1 2

e tha 10 mi. LWR distance..

There is also a considerable difference in the time of the activity released from FSV comp'ared to the reference LWR as shown in Fig. 6-3 for the inhalation thyroid. This figure shows that for the LWR high thyroid doses (100 rem) have already been received at 1 mi. in the first hour,.while 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> should be avail-able even at 590m for FSV, before the lower PAG (5 ram thyroid) is reached al-lowing an. orderly and organized evacuation to be done if necessary.

6.2 Ingestion Pathway EPZ Evaluation In addition 'to direct. plume exposure, the population surrounding the nuclear power plant can be subjected to pathway doses from contamination of the sur-rounding creps, =11k, and food stuffs subsequent to the atmospheric release of radioactivity in the accident.

The protective action guides for exposure from food stuffs or water (Ref. 3) define the guidance to be exercised by the utility in its asseosment and control of the post accident ingestion pathways.

The most severely restrictive ingestion pathway for FSV and the reference LWR accident examined was found to be the iodine-grass-cow-milk-infant thyroid pathway.

The predicted thyroid inhalation dose vs. distance for both the reference LWR and FSV is displayed in Fig. 6-2.

The iodine-grass-cow-cilk-infant pathway may be related to I-131 plume exposure dose by using an approx 1= ate factor of 300 (Ref. 2) and recognizing the fact that the thyroid dose reported in Fig. 6-2 is nearly all due to I-131 exposure.*

lFhe plume thyroid inhalation dose, which is proportional to ingestion pathway dese, for the reference LWR is 0.35 Ra=.**The distance at which this dose level is reached for FSV is about 30 miles and is consistent with FSAR Chpater 2 discussions of demographic studies at distances up to 30 miles from the site.

• I-131, at 50 miles, contributes 89" of the reference U m plu=c inhalation dose commitment and correspondingly 97% for FSV.

• • 'Ref.

2, Table I-2 extended to 50 mile radius.

4 7.

Conclusions It is concluded from this I?Z study tha: a sig:iifican:1y smaller evacuation radius can be jus:1fied for the Fort St. 7:ain plant than for a reference 1000 MW(e)

L'a'A.

A further conclusion from this study is that the unique characteristics of :he l

(

FSV plant provide a slow gradual heacup of de core al1owing consi[ierable ti=e

~

(on the order of tens of hours) for orderly notificacion, monitoring and eva-cuacion.co proceed. uso, a smaller radius for the controlling ingestion pathway, the grass-cow-milk-infant thyroid, can be derived which offers the public ':he same protection as a similarly sited L*nA.

In both cases, thyroid dose is the limiting cons :aint. These conclusions are based on si ing event analyses using generally conservative core release frae:1cas and meteorology. Table 7-1 su=-

=ari:es the applicable emergency planning :enes for :he plume exposure pachway and the ingestion exposure pathway for For: St. Vrain providing equivalent pro-eaction :o the general public.

References 1.

"D AC - An Analytical Cc puter ? ogra= :o Calculate the Ti=e Dependen:

Radiological Iffec:s of Radionuclide Release", D. W. Buckley, General A:omic Co. Repor: GA-D13476, 1976.

2.

NURIG-0396, " Planning Basis for the Development of State and Local Goviirn=ent Radiological Isergency Response ?lans in Suppor: of Ligh: Wate: Nuclear Plants".

3.

" Manual of ? ocective Action Guides and Protective Actions for Nuclear Incidents", Environmental ? otection Agenc'/ Report, E?A-320/1 001, 1975.

4 FSV Safety Evaluacion Report, Division of Reac:or Licensing, TSAEC, Jan. 20, 1972.

l

,-i

i.

o.

TAllLE 3-1 a

AlliOSPilERIC OILUTION FACTORS

• FOR FSV DBA #1 AND REFERENCE LWR DDA/LOCA BASED ON NRC STAFF VALUES USED IN Till FSV SER

~

DISTANCE (H)

TiltE (ilR) 590 1600 4800 8000 16,000 24,140 32,200 48,300 64,400 80,500 8

7 2-4 1.43-4 2 76-5 1.65-5 7.89-6 5 09-6 3 77-6 2.5-6 1.9-6 1.5-6 16 5.1-4 1.02-4 1.96-5 1.17-5 5.61-6 3.62-6 2.68-6 1.8-6 1.3-6 1.0-6

~

I 24 4.1-4 8.35-5 1.61-5 9.60-6 4.59-6 2 96-6 2.19-6 1.4-6 1.I-6 8.6-7 34 3.5-4 7.03-5 1.35-5 8.09-6 3.87-6 2.50-6 1.85-6 1.2-6 9 1-7 7 2-7 40 3 2-4 6.49-5 1.25-5 7.46-6 3.57-6 2 30-6 1.70-6 1.1-6 8.4-7 6.7 52 2.8-4 5 7-5 1.1-5 6.56-6 3 14-6 2.02-6 1.50-6 9.8-7 7.4-7 5.8-7 58 2.65-4 5.4-5 1.04-5 6.21-6 2.97-6 1 92-6 1.42-6 9 3-7 7 0-7 5.5-7

.l 100 2.05-4 4.13-5 7.95-6 4.75-6 2.27-6 1.47-6 1.08-6 7.l-7 5.4-7 4.2-7 1

400 1.03-4 2.09-5 4.01-6 2.4-6 1.15-6 7.4-7 5.47-7 3.6-7 2.7-7

.2.1-7 720 7 7-5 1.56-5 3.0-6 1.8-6 8.59-7 5.54-7 4.1-7 2.7-7 2.0-7 1.6-7 1500 5 3-5 1.09-5 2.09-6 1.25-6 5.90-7 3.86-7 2.85-7 1 9-7 1.4-7 1.I-7 4320 3.2-5 6.45-6 1.24-6 7.42-7 3.55-7 2.29-7 1.69-7 1.1-7 8.4-8 6.6-8

'Vdines in units of S/m are listed in exponential form such that, e.g., 7 2-4 means 7.2 x 10-3 i

I i

---

=:- -- - --

Table 5-1 EPZ Study - Assumptions and Parameter Selections Parameter Description,

FSV Source Term Power level, MW(t) design 879 MW(c)

Accident scenario assumed DBA #1 (FSV FSAR Appendix D, page D.1-56)

Release of fission products frem Time dependent release

• over a 475 the FSAR fuel particle hour period Release frsction from fuel to TID-14844 fractions enployed primary coolant system Noble gases 100%

Iodines 50%

Others l'

Plateout reduction factor TID-14844 reduction of 50" applied to (Iodine only)

Iodine transported from the primary coolant system to the reactor con-tainment building.

FCRV leak rate 0.2" volume / day (total duration of accident)

Reactor building ventilation system characteristics Purge rate 1.5 volunes/ hour (en for total duration of the accident)

Filter efficiencies Noble gases 0:

Halogens 90" Particulates 95:

Additional plateout and gravita-Ignored tional settling of condensible nuclides in the confine =ent building i

• Time-dependent source ters identified by designation FSVTID*FUELREL. and discussed in engineering fila C-70-001, F. S. Domoek.

m r-

~_

~

- u t

Table 5-1(Cont.)

i Parameter Description-Reference L7R*

Source term Power level, MW(c) 2958.

Accident scenario assumed DBA/LOCA Release of fission products from.

Near instantaneous release of the core matrix to the reactor TID-14844 core fractions building Core inventer,r, available for TID-14844 assumptions (including.

release from containment iodine reduction due to plateout)

Noble gas 100%

Iodine 25:

Available for leakage from the reactor contain=ent i==ediately Isotoons after LOCA (C1)

I-131 1.9+07 I-132 2.7+07 I-133 4.2+07 I-134 4.8+07 I-135 3.8+07 Kr-83M

1. 3%7 Kr-83M 3.2+07 Kr-85 7.3+05 Kr-87 6,.3+07 Kr-88 8.8+07 Kr-89 1.1+08 Xe-131M 6.5+04 Xe-132M

4. 0M6 Xe-133 1.7+08 Xe-135M 4.5+07 Xe-135 2.8+07 Xe-137 1.5+08 Xe-138 1.5-08

• ' Ails not specifically identified as the Sun Desert Nuclear Power Plant, the reference LWR of this study has many characteristics similar to this current PWR design.

7 rw w-tfw r

v

-*-r'er yyi+y e-we 14

• r'*t-%d

--*==Wt--*-v

~

J o.

,,y.

1 Table 5-1(Cont. )

Parameter Descrip tion Iodine composition Elemental 91:

Particulate 5%

Organic 4%

Contain=ent building leak rate 0-24 hou:'c.

0.2%/ day

>24 hour 0.1%/ day Containment spray characteristics Initiation time 150 see Sprayed region 70%

Unsprayed region 30%

Mixing race sprayed to unsprayed 2 hr-1' region Elemental Iodine removal rate 10 hr-1 Parriculate Iodine removal rate 0.45 hr-1 Organic Icdine re= oval rate 0

Iodine DF (elemental) 100 Resultant single volume Lk'R 3.0 hr-1 spray re= oval constant used this study Post LOCA containment purge

.i characteristics (to reduce combustiole gas cencentration)

Purge interval 336 te 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> j

I Purge rate 50 SCFM Purge filter efficiency Elemental & Methyl Iodine 70%

Radionuclide leaknge frem ESF Ignored area

g.

c

' - i Table 5-1(Cont. )

Parameter Description Specific Site and Dose Parameters (Used for both FSV & Reference UJR dose estimates)

Atmospheric dilution factors See Table 3-land Figure 3-1 employed Breathing rates Reference U4R 0-8 hour 3.47 x 10-" =3/see 3

8-24 hour 1.75 x 10-4 m /sec 3

24-720 hour 2.32 x 10-4 m /sec TSV*

3 0-40 hour 2.32 x 1C-4 m /sec 3

40-52 hour 3.47 x 10-" m /sec 52-58'heur 1.75 x 10-4 =3/sec 3

58-720 hour 2.32 x 10-" m /sec Average vind speed 1 m/see Dese code utill:ed TDAC - GA-D13476 Decay & buildup esti= ate Considered Fallout deposi:Lon enroute :o Considered for Iodine only (see dose receptor Reg. Guide 1.111 Figure 6)

Conversion fae:or for infant-300 (Ref. NUREG-0396 Appendfx 1,

-dik thyroid Iodine ingestion page I-19) pathway (ratio of thyroid dose commi:=ent factor for milk pathway to the inhala: ion plume exposure pathway)

• Note:

FSV breaching rates and a:=cspheric dilution factors have been per=uted to nae -ize :he off-si:e dose, i.e.,

he ti=es of highes:

d breathing rate and poorest atmospheric dilution correspond to :he peak in ti=e dependent release of fission produe:s from :he fuel :o :he primary coolant.

i i

Table 6-1 Summary of Derived EPZ Evacuacion Distan'ces for FSV and a.

Reference 1000 MW(ei LWR Whole Bodv Gamma - Direct Plume Excesure l

No Protective Action (PAG <1 rem whole body)

[

EPZ Distance

}

FSV LWR 0-8 Hour

< EAB(l}

Y.T Mi.

0-24 Hour

< EA3 2.5 M1.

l 0-30 Day 0.6 Mi.

3.0 M1.

f Evacuatien Required (PAG),5 ram whole body) l EPZ Distance l

FSV LWR f

0-8. Hour

< EA3 0.9 Mi.

l 0-24 Hour

< EA3 1.0 Mi.

f 0-30 Day

< EA3 1.2 Mi.

l Thvreid Inhalati:n - Direct Plu=m E::eesure j

i No Protective Action (PAG <5 res Thyroid)

[

EPZ Distance

{

FSV L'4R t

I 0-8 Hour

< EA3 7.5 Mi.

I 0-24 Hour 1.2 M1.

9.0 Mi.

0-30 Lay 8.1 M1.

16.0 M1.

l t

Evacuation Required (PAG 2: 25 ram Thyroid) i EPZ Distance l

FSV LWR 0-8 Ecur

<EAB 3.1 Mi.

0-24 Hour 0.5 M1.

3.5 M1.

0-30 Day 2.5 M1.

5.3 M1.

i (1)

FSV EA3 (Exclusien Area Boundary Distance) 590 meters.

l i

1 l

i i

1 i

h

4 Table 7-1 FSV EPZ Distance Compared to Generic LWR E?Z Dis:ance of NUREG-0396(1) i Accident Phase Generic LWR Emergency Comparable FSV Emer-Planning Zone Distance gencyPlangingZone NUREG-0396 Distance ll i

1 Plu=e Exposure Pathway About 10 =iles radius About 5 miles (3)

Whole body (external)

Thyroid (inhalacion)

Other organs (inhalation)

Ingestion Pathway About 50 mile radius About 30 miles (')

Thyroid, whole body, bone = arrow (ingestion)

(1)

NCREG-0396, E?A S20/1-78-016, "?lanning Basis for the Development of State and Local Gover==en: Radiological E=e:gency Response ?lans in Support of Light Water ?cwer Plants", Join: Report USE?A and USNRC.

Collins, H. E.

et. al., Dec. 19 3.

(2)

FSV E?Z distances were obtained by (1) assu=ing that :he Reference LWR of this s:udy =eets the intent of NCREG-0396 for design basis accidents, (2) esti=a:ing the Reference LWR dose at 10 miles (plu=e) and 50 miles (ingestion) for :he DEA /LOCA of this study and then. (3) determining the corresponding equi-dose distance for the TS7-DBA #1 event.

(3)

Thyroid inhalation con::els with about a 5 mile E?Z.

E?Z distance for whole body is approximately 2 miles and other organ doses are negligible.

(4) The grass-cow-nilk-infant :hyroid exposure pathway dominates for the FSV Snvironmentally released radionuclide source term.

l P

l

F v

f 1.0 FSV FSAR Hodel f

1

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I Figure D.1-13 j

s

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0.8 -Il0 REG-0111 Hodel,

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THISO UC2 Particle with j

c 3

lii gh Burnup -----------

f 5

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,t 0.6 _FSV TRISO (4 Th, U)C

• k-2 3 /

i 18.2% FIHA I

Temperatures Increased

/

/

from 1015* to 2400 C as

]as Over Period of 9 Ilours

/

0.4. (Ref. GA-A14744,

/

December,1977) 0 0

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m b

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0.2

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n n i n i n n nnnn i

900 1100 1300 1500 1700 1900 2100 2300 2500 TEMPERATURE (*C)

Figure 2-1 Comparison of Models for Fuel Failure as a function of Temperature versus Observed Release fraction on Fissile A FSV Production Fuel

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Integration intervals of 8 'hr,1 day, and 30 day:. Comparison of FSV-UBA-1 Event and i

1000.MW(c) LUR-DBA/LOCA sitin's event.

NRC'meccorology (from CER) for both.

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FIGURE 6-2 Chmulative plume exposure Inhalation thyroid dose vs. downwind distance.

Inte-gration, intervals of'8 hr, 1 day. and 30 days. Comparison of FSV-0BA 1. Event and 1000 MW(c) LWR DBA/LOCA siting tvent. ~ tlRC ncteorology (f ron SER) for botn.

w 1

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PLATTEVILLE, COLORADO 80651 March 18, 1980

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Fort St. Vrain P-80083 Mr. Brian K. Grimes

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Director Emergency Preparedness Task Group Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D.C.

20555

SUBJECT:

Fort St. Vrain Unit No. 1 Radiological Emergency Response Plan O

Dear Mr. Grimes:

We are transmitting herewith two (2) copies of Radiological Emergency Response Plan in draft f o rm.

There are some detailed areas of the plan which have not been completed as yet, but the basic plan should be complete enough for you to begin your review.

The plan has been developed on the basis of the acceptance criteria published as a part of NURIG 0654, and generally meets the intent of NUREG 0654 with the following exceptions.

1.

Near Site Emercency Operations Center Our near site emergency operations center (Forward Command Post) has been designated as the Fort Lupton Municipal Building which is located approximately ten (30) miles south-j southeast of the plant.

As indicated in our letter P-79205 our Forward Command Post location was developed with the State / Local agencies and has been equipped with necessary.

communication systems, supplies, etc.

This facility was utilized during our co=bined drill with the State on Febru-ary 28, 1980, and proved to be more than satisfactory in all respects in accomplishing the coordination and direction of emergency activities.

We can see absolutely no justification for establishing an emergency operations center within one (1) mile of the site as required by NUREG 0654.

In fact such a center within the EPZ presents problems of access, personnel control, possibility of evacuation and other problems which are not experienced at the Fort Lupton location.

We clearly demonstrated during the VUj)l OA

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1. tHarch l'8, 1980 Mr. Brian K., Grimes. '

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1. -Near Site Emergency Doerations Center (continued)

Februaiy 28,.1980,- drill that an emergenc'y 'can be bandled at 9

the Tort Lupton facility with'a minimum of.. problems, and in

^

our opinion with'.far less problems than emergency center with.

'in one'(1) mile'of the site.( The services availab'le in Fort J ~

Lupton would not be available'at a near' site center, and we' feel moving the center from Fort Lupton would only serve as

a. detriment to our overall plan.

Based on.'the 'results of our drill and the lack of any justi-fication for being closer to the site we feel that adequa.:y of o.ur Forward' Command Post location has been justifie'd.

~

2.

Plant Staffing-hTREG 0654. set forth certain minimal requirements for plant staffing. in Table B-1, which in effec't requires ten (10) people on shift and the capability of having an additional i

twenty-six (26) people on site within thirty (30) minutes. - As indicated in our previous responses to the TMI-2 Lessons Learned requirements and as confirmed by Nuclear Regulatory Commission acceptance of these requirements (undated letter, from T. Speis to J. Fuller, March,1980), emergency situations at Fort St. Vrain develop very slowly as compared to = a.. water reactor.

As a. result, operator and emergency response times are considerably longer.

Our emergency plan has been developed cn this basis, and therefore, the require =ents of Table B-1, NUREG 0654, are' considered inapplicable to Fort St. Vrain.

3.

Early Warning Alert-System As indicated in our recently transmitted study (see P-80066) concerning the EPZ for Fort St. Vrain we do not exceed any of the PAG's for cumulative whole body gan:ma dose for the first twenty-four (24)' hours following a LOFC accident, and never reach the PAG's requiring evacuation.

For thyroid inhalation exposures we do not exceed the PAG for "no protective action required" for up to twelve hours at the exclusion area boundary (22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br /> at one mile) and we do not exceed the " shelter, monitor *and control access" PAG for up to twenty (20) hours at the exclusion area boundary (50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> at one mile).

Based on these times more than adequate time e>;ists for notifica-l cion of the public and for initiating. evacuation of the public i

without the need for an early warning alert syste=.

It should be noted that these exposures are based on extremely conserva-o tive releases of Iodine and extremely conservative reactor vessel leak rates. More realistic release rates of Iodine would be a f actor of ten (10) below those assumed for the EP2 study.

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" Mr. Brian K. Gribes.

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4.

Meteorological Systems

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j Our present plan does not meet t$e requirements of NURIG 06,54 concerning backup meteorological instrumentation nor the re-quirement for remote interro.gation c.apabilities.

We are pres-ently investigating other alternatives.

5.-

Tso-Dose Curves The kso-Dose curves have not been included in the draf t plan.

as we have been unable to determine from the Nuclear Regulatory Co= mission just what is required in this area.'

6.

Letters of Agreement Letters of agreement with various off site support 'organiza-tions have not been included at this time as we are still in the process of developing these agreements.

In general, how-ever, our State plan is designed as a general agreement for all participating agencies, and those agencies that are signa-tory to the State Plan will not be called upon for separate letters of agreement.

o tTUREG 0610 The guidance provided by NUREG 0610 has been modified to best fit HTGR technology versus water reactor technology.

The re-quirements for " Unusual Event Notification" are included in an agreement between Public Service Company and the State which was. originated prior to NUREG 0654.

Since this agreement meets the intent of NURIG 0610 and has the concurrence of both Public Service Co=pany and the State we feel it is adequate to fulfill the notification requirements for unusual events.

Other than minor differences we believe the plan meets the intent of NUREG 0654 acceptance criteria.

If you should have any problems or questions we would appreciate hearing from y u prior to the May 21, 1980, meeting date.

Very truly yours, Wcuud w~

Don W.

Warembourg 67 Manager, Suclear Pr oduction Fort St. Vrain Nue. lear Generating Station j

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Fort St. Vrain Unit No. 1 F-3^238 r

Mr. Robert L. Tedesco, Assistant Director Division of Licensing U. S. Nuclear Regulatory Commission Washington, D.C.

20555 i

SUBJECT:

Fort St. Vrain Unit No. 1 Emergency Response Plan

REFERENCE:

NRC Letter July 23, 1930

)

Dear Mr. Tedesco:

We are transmitting herewith three (3) copies of our revised i

emergency response plan.

This revised plan includes changes as a result of the May 21,

1980, plant site review meeting as well as certain cha1ges that.resulted from ccmments contained in your j

July 23,

1980, letter.

In addition to the revised plan, we are provi :ing our response as Attachment A to this letter to address your July 23, 1930, letter.

l r

As we indicated in the May 21, 1930, meeting as well as in various corresponcence submitted as_a part of the TMI-2 Lessons Learned j

Tasks, we believe Fort St. Vrain is a completely different reac:ce concept.

This reactor concept coupled with the si:e of the reactor negates many of the requirements setforth by NUREG's 0554 and 0610 which were developed primarily on the basis of 1,000 K4(e) ligh:

i water reactor technology.

It is imperative, therefore, that our Emergency Plan be evaluated on the basis of our reactor design and l

size, and that-generic requirements be evaluated on,the basis of i

specific technical, safety, and environmental diff erences*.

t We have had to essentially develop our own criteria for Fort St.

i Vrain utilizing water reactor criteria setforth by various Nuclear Reg ul a t o ry Commission documents.

On this basis our criteria is ne:essarily different from that published and we have taken I

ustifia'ie excection to the NURE3's.

These exce;; ices,ere o

s.oportec in various correspondence (see reference list attached) and i

are further supported by Attachment A of this letter.

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In the May 21, 1980, meeting we were informed by the Nuclear Regula Ory Ccemission review team that many. of the items were a matter of policy, and that the review team did not have the authority to make.ex:eptions on policy -

regardless. of the tecnnical matters justification.

We canno a: ept this position, and we recues :na:

i as soon as you nave had tne Opportunity review our revised emergency plan anc our respohse that we ce_given tne opportunity to meet with you and other personnel who de have tne authority to evaluate and/or at:ept our positions on the basis of the :eennical

- justification provided.

In the. interest of time it is requested that such a meeting De establishec at ne earliest possible cate so na we may finali:e our plans to -meet the various c:mmitmen ca:es se forth.

We will be available to meet with you at your ::nvenience and are looking f o rwa re :: nearing from you snortly.

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Very truly yours,

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Don W. Warem:ourg 4

Manager. Nuclear Precu::icn Fort S. Vrain Nuclear Generating Station OWW/alk i

A tachments 4

4 5

4 i

4 4

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PSC CORRESPONDENCE LISTING LETTERS TO NRC INVOLVING TMI-2/ EMERGENCY DLANNING/EMERGEN;Y RESPONSE PLANS Cor*escencen:e I"ummer Date Subject 1.

P-79130 June 15, 1979 Gaseous Effluen: Monitors 2.

P-79205 September 10, 1979 Emergency

Planning, For:

St. Vrain 3.

P-79239 October 17, 1979 Followup Action TMI-2 4.

P-79249 October 29, 1979 Followup Action Resulting From NRC Reviews Regarding the TMI-2 Accicent 5.

P-79290 Novemoer 30, 1979 NUREG-0610 6.

P-79298 December 12, 1979 Fort 5:. Vrain, Unit No. 1, TMI Lessons Learnec 7.

P-79299 Decemoer 12, 1979 Revised Followup Actions Resulting From NRC Reviews Regarcing TMI-2 Ac:icent

~~

8.

P-79305 Decem er 18, 1979 Sucpl ementa ry

Resconse, 1:em 2.2.1.0, Lessons Learned Task Force, TMI-2 9.

P-79312 Decemoer 28, 1979 Additional Information Regarcing June 1,

1980, Action Items Resulting from TMI-2 10.

P-80011 January 29, 1980 Recuest for Evacuation Times 11.

P-30028 February 20, 1980 Accitional Information Resulting from TMI-2 NRC Review Team Site Visit, January 21-22, 1950 12.

P-300a1 March 3, 1980 Recuest for Evacuation Times

'. 3.

P-30053 Mar:n 16, 1950 Fort St. Vrain., Uni; h:: 1, Raciciogical Eme gency Re>conse Pian 14 P-30066_

Ac-il 1,

• 980 Fort St. Vrain, Uni: No. 1, Emergency Planning

ATTACHMENT A PSC RESPONSE TO NRC COMMENTS FSV EMERGENCY RESPCNSE PLAN i-1.

NRC Ouestion/ Comment D*an

.cu s t ' be revised to establish a principal anc an alternate

~

EOF.

Both facilities should meet the requirements of I

Darrell G. Eisenhut's letter.

of April 25,

1980, subjee:

" Clarification of - NRC ' Requirements for Emergency

Response

Facilities at Each Site."

PSC Resconse J

We cannot acdress the principle and alternate EOF as we have never received the April 25, 1980, letter which you reference.

It is our uncerstanding that new criteria will be publisnec as a part of NUREG-0696.

Upon receipt anc evaluation of this cocument we wil.1 modify our emergency olan.

In the in erim we intenc :o continue with.our plans to utilize :ne Fort Luoten Municipal Building for the EOF as stated in our letter'P-ECOS3.

As we understand the new criteria being develocec uncer NUREG-0696, a distance' of appreximately 10 miles from ne rea::or would be a::eptable for ne

EOF, Depending on the criteria specified for the EOF and an alternate EOF we will re-evaluate our position at the -ime NUREG-0696 is publisnac.

2.

NRC Ouestion/ Comment Plan must ce revised to take into consideration :ne plant staffing in Table 5-1 of NUREG-0654 There must ce some augmentation of on-site personnel within 30 minutes.

Mus; icentify position that will not ce filled and provice rationale for not having 10 personnel on shif: at all times.

PSC Resconse Consiceration was given to the plant staffing in our 'Acril draf t of :ne RERP.

Figures 5.1-1 nrougn 5.2-6 'of tne RERP depict botn ne normal and the emargency staffing for the clant.

Figure 5.1-2 provices the normal c:erating staff (9 tersonnel plus a Lead Security Officer) for :ne pian; and fulfilis the on-snift reovirements of Table 5-1, NUREG-0554, witr

ne exce;; ion :na we ce not have a Rac/ Chem Tecnnician on snift.

The on-snift Heal n Physics Te:nni: tan nas sufficient training to cerform ne necessary initial surveys and raciciogical assessmen s to orete:: in-clant eersonnel.

Tne coerating staff has sufficient training and cro:ecures to evalua e :ne off-si e effe: s.

We can see no immedia e recuirement for the Rad / Chem Tecnnician' esoecially since Our ac:icents ceveico a a mu:n slower rate then comoarable water ea: or ac:icents (see NRC le er Inemis Speis to J. Fuller, Maren,198C, Acce0:ance of Category A TMI-2 Recuirements).

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4 i

With reference to the augmented staff callec for in Table B-1, NUREG-065a, we have justificd delaying the resporse time of the Shif: Technical Advisor (STA) (based again on the rate in which our accidents. develop) from 10 iainutes to one'(1) hour (see PSC letters P-79249, October 29, 1979; P-79299, December':12, 1979; P-79305, December. 18, 1979; P-79311, Decemoer 28, 1979).

The accioen: time frames and the associated response times were accepted by the Nuclear Regulatory Commission by the above referenced letter (Themis Speis to J. Fuller, March,1980) in the overall acceptance of the Category A TMI requirements.

Since the Nuclear Regulatory Commission found the resoonse time of the STA to be acceptable we maintain :nat the 30 minute' l

augmented staff time called for in NUREG-0654, Table B-1, is not applicable to Fort St. Vrain. On the basis of the slow time in wnich accidents cevelop and the one (1) hour resonse time of the Technical Advisor we committed in our RERP (Section 5.2) to have One emergency organi:ation activated within 90 minutes which would include an augmented staff equivalent to Table B-1, NUREG-0654 This staff auomentation is ' consistent with Tecnnical Advisor response time anc is certainly consistent with a::icent analysis and the ac:ident development time frames.

3.

NRC Ouestion/ Comment The plan must (in acdition to other NUREG-0610 notification recuirements) specify that when a " general" emergency is ce:lared that the off-site authorities resconsible for imolementation of protective measures will be notifiec ey the

" Plant Emergency Director" and aavised of recommencec protective actions within 15 minutes of tne direction of the emergency

. condition.

The pian must specify the content of this initial message to include:

a.

Class of emergency b.

Whether a release is taking place c.

Affected areas a

d.

Protective measures NOTE: The protective measures recommended in the initial message off-site may be "go inside - turn on radio" (30 minutes) croviced a followuo message indicating more detail protective measures casec on cose projections.

PSC Resconse Per your re:uest the notification time of fifteen (15) minutes af ter cetermination :nat a "gener61" emergency exists nas ceen acted to Table 4.1-4 of :ne RERP.

Samole notification messages as well as followuc messages have been included in Section 6 of the RERP (see Figures 6.1-1

nrougn 6.1-3).

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NR; Ouestion/ Comment t

Your clan must describe the public notification sys em to 1

incluce:

a.

The initial off-site contact who will be responsible for not'i fyi ng the affected copulation (either the specific j

organi:ation or individual).

1 b.

The capability for 24-hour per-day notification >(to off-site authorities).-

c.

The physical alerting system to be used:

sirens, NOAA weather of emergency alert, telephone automatic dialers, aircraft with loudspeakers (wnien will be used to alert public).

(10 db above average caytime ambient background is a target level for design of an adeouate siren system.)

Distance

% Notified in 15 Minutes 5 miles 100%

d.

The basis for any exceptions (e.g., for extenced water areas with transient boa s or remote hiking trails) must be documented.

e.

Eve ry year you must take a statistical sample of -he residents of all areas within the 5 mile E.DZ to assess the public's awareness of tne promet notification system anc the availacility of information on what to de in an emergency.

Plan must also incluce a provision for corrective measures to provice reasonable assurance that coverage accroa:ning the cesign obje:tives is maintained, f.

The provisions for use of a public media system (radio, TV) to provide clear instructions to the public.

a.

Twenty-four hour station - total plume coverace.

b.

Include in the plan the messages to be ransmi ted to ne puoli: (cover a range of prote::ive actions).

It is ne coerator's responsioility to ensure na: :ne means exist for no-ifying an: provicing p-omo: instru icns o :ne puolic.

It is the resconsibility of ne State and local governments to activate tne system.

t

.3 g DSC Resconse c >-

1-a.

The County and the State have the resconsibility for notifying the affected peculation. Additional clarification has been added to Section 6.1 of the RERP.

b.

The caoability of 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> per cay notification to off-site authorities is incluced in Section 7.2 of the RERP.

c.

and d.

The physical alerting systems proposed both by the Port St. Vrain RERP and the State RERP is tne use of the amergency broadcast system (radio) and the dispatch of County personnel-in vehicles and loud speakers to warn the general public.

As indicated in our le::er P-80083, March 18,.1980, we have more than adecuate time to effect necessa ry protective actions prior to exceecing any of the Prote:tive A:: ion Guidelines.

During the Feoruary 2S,1980, drill we tes ed the State Emergency Resconse Plan and cetermined that the general puolic in tne affected :ene

culd 'be notified within 90 minutes.

As indicated in our letter P-30066 we nave uo to twelve (12) nours to notify the cuali: prior to exceeding tne most restrictive Prote::ive A: icn Guiceline (iodine innalation) at :no exclusion area bouncary. The plan for public notification inciuces the usa of several emergency breac:ast bands and ne use of

21evisi on augmented, of curse, with dispatched oersonnel.

- On the basis of the time f*ames associatec with our ac:icent analyses and the use of radio anc televisien media we can i

see no justification for an early warning alert sys em, e.

As a part of the oubli: information :rogram a s atistical samole of the resicents wi-hin the five (5) mile EPZ will ce taken to assess the public's awareness of what to co in an emergency.

The details of the public information program are still being worked out between PSC anc the State.

Once this program is cefinec Section 3.1.1.d of the RERP will be ceveloped to define the program.

f.

Use of the media systems to provide instru:-ions to the public is the responsibility of State,

County, and local authorities.

The State RERC which is accreved by DSC has provisions for the use Of media sys ems na; mee; tne recuirements of NUREG-0654 Sec: ton 6.4.1.c of :ne RERP ideresses tnis suoje:.

5.

NRC Cues-ion /Commen:

Uncer PSC actions, Table 4.1-1, eage 1, :ne firs: ham.

ne 2-nour time limit for no-ificatien of an o::urren:e of an Unusuai Event should be changec to "as soon as ciscoverec".

i Similarly, the four:n item uncer PSC a:: ion snoui inciuce a recuiremen-for a written summary ni:nin 24 nours.

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The Kinc Furnace fuel-to-core release fractions vere employed for iodine release in the alternate scenario, i.e., 5.5j of the core idoine is availabla for leakage from the 20R7 circuit versus 255 for the 0:0-lhSLL release previously e= ployed.

t 3

A FORY leak rate of 0.15/ day was e= ployed, consistent with large ET33 containment buildin5 leah rates-k.

Ua ural deposition of condensible fission products in the reactor building was allowed.

A reduction of about LOS for iodines was empicyed for the alternate scenario.

In the previous Ref.1 ( Attach =ent :) analysis, TID-lh3hh assveptiers on icdine reduction were employed, i.e., 505 of the fuel-borne iodine is assumed to be released from the fuel to the PCRV circuit and of this 305, a further reduction of 50% vas employed within the PCRV to eccount for platecut', thus resulting in an effective source of 255 Of the fuel iodine inven ory svailable for leakage from the PCRV.

fo credit for further plateout of the iodine once released to the reactor building was empicyed in the Ref. 1 (A :achment I) analysis (the current alternate scenario e= ploys about a -05 reduction due to reactor building platecut phenomenen).

5.

All her conditiens were iden-'-C o ??C I?: assumptiens as given in u...,,

Ia%..a,.w.

e..

.s,.

Table i sumarizes the assumptions emplored in the current alternate scenario analysis, as well as provides the reader with a statement of some of the additi:nal conservatisms felt Oc be present in the Ref.1 ( Attachment :)

analysis, but not modified at this time.

Eesults The analysis =e-hods e= ployed in this current study are those same methods employed in the Ref. 1 (Attachment :) verk.

Figures 1 and 2 su=marite the cumulative (to 30 days) thyroid inhalati n dose and whole bcdy gamma exposure, respectively, for the DBA-1 alternate scenario of this study.

Additionally, for c =parison purposes, the previous FS7 I?: study results as well as 1000 T4(e)

LWR results are also displayed.

t is observed that beyond the FSV EA3 of 590 meters, neither whole body gan=a nor thyroid inhalation dose exceeds the FAG for sandstory evacuation on the basis of D3A-1 predicted doses using the alternate scenario.

Applying the Ref.1 ( Attachment :) approach of comparinz ;he 737 dese versus iista. e curves for the 02A-1 scena-i: vi-h a " reference" 1:0 0 Ti( e ) 1WR, ::

the alternate scenario dese curves :f Firs.1 and 2 results in an I?: of abcut

ne mile for the plume submersion pathwa:.

Usine this same method, five miles was derived for the incestion exposure pathway.

These res"'-e an crier cf a-a narnitude icver than the " reference" 17: I?:'s which are 10 miles f:r plume excesure and 50 r.iles for the inrestion f:cd pathvay.

t is noted that the 330 subsequently has ruled tha; the ;?.ute exposure and ingestion pathway distances are respectively 5 and 10 r'.les ':r Fort St. 7 rain and LWR's of 250 T/fe i or less in aize.

9

e.

3-

.n.

4 Fic. 3.: resents the time history of the inhalalation thyroid dose at the LI.O (590 meters) and at the Platteville boundary distance of three miles.

1 "his fizure shows graphically the slov buildup of dose at the dose receptor coint,;chara:teristic of the ETGR.

References

.l.

"FSV Emergency Pier.ning Zone (I? ) Assessment," PSC to NRC Letter

?-80066..

2.

.FSV Technical Specifications, "SR 5 5 3 - Reactor Buildins Ex.haust Filters, Surveillance," Amendment No. 18.

3 FSV FSAR, Amend =ent 27,-p.'6-1-8.

L '.

" Nuclear Power Plant Air Cleaning Units and Components," Table 5-1 Test 5d., ele = ental iodine retention, AUS!/ASME Standard N509-1976.

5

" Design,' Testing, and Maintensnee Criteria for Post-Accident Engineered-Safety-?eature Atmosphere Cleanup System Air Filtration and Adsorption Units of Lizht *.~ater-Cooled Huelear ?cuer Plants," U. S. Regulatory Cuide 1 52, Eev. 2, March 1978.

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