ML20198G504

Forwards Accident Analysis Branch SER Input.Several Items Requiring Resolution Noted,Including Redesign of Spray Additive Injection Sys & Failure of Control Room Dual Inlet Design to Meet Acceptance Criteria

ML20198G504
Person / Time
Site:	Washington Public Power Supply System
Issue date:	03/04/1975
From:	Harold Denton Office of Nuclear Reactor Regulation
To:	Moore V Office of Nuclear Reactor Regulation
References
CON-WNP-1050 NUDOCS 8605290631
Download: ML20198G504 (34)

Text

o O

.YM 1 1375 V. A. Moore, Assistant Director for Light Water Reactors, Group 2, AL l

WASHUGTON PU3LIC POWER SUPPLY SYSTEM - WPPSS NOS.1 & 4, SER i

INPUT BY THE ACCIDCff ANALYSIS BRANCH PLANf NAME: WPPSS Nos. 1 & 4 LICENSING STAGE: CP DOCKI*r NUMBER: 50-460 50-513 2ESPONSIBLE BRANCH: LWR 2-3; T. Cox, LPX REQUESTED COMPLETION DATE: January 23, 1975 REVIEW STATUS AAS Review Completa Euclosed is the Accident Analysis Branch (AAS) SER input for the Washington Public Power Supply System's nuclear plants, Nos.1 and 4 (WNF-1 and WNP-4). There are several items which will require resolution. These are a need for redesign of the spray additive injection system, failure of the control roosa dual inlet design to meet our acceptance criteria, a need to have the area where the ECCS equipment is located served by ESF filters, a need for a comunisment limiting the quantity of explosives shipped by rail past the site, and a need for confirmation, by CGC, that the applicant has reasonable aanurance of control over the exclusion area.

This review was coordinated by L. Soffer, Site Analyst, of the Accident Analysis Branch. An advance copy was given to the LPH on February 19s 1975.

Harold R. Denton, Assistaat Director for Site Safety Division of Technical Raview Office of Nucisar Reactor Regulation Enclosures SER Inpet cc: See next page 0 %.,. u - %

I I

I 0605290631 750304 PDR ADOCK 00000460 E

PDR

- " " " " ' " ~ " ' - -

pate w N foran AEC ll8 (Rev. 9 95) AIChi 0240

1. w. e. eevemament reentene erriete 1 94 888184

/

9

,~

i I

':D

. 1315 V..A. Moore I l

sa w/o anelosures i

A. Giambusso j

W. MsDonald J. Emmaarella es w/eselosures

8. Hanauer F. Schroeder R. Boyd R. Elecker i

S. Varga D. Eisenhut TR A/D's SS B/C's TR T/C's A. Salnesacar T. Cam i

K. Murphy W. Pasedag i

K. Campe J. Read L. Soffer DISTRIBUTION:

I Docket Files NRR/RDG AAB/RDG I

i i

i

)

s AAB/TR, b,. [

. AD/S '/ER...

i eme s,

LS er/ tab,

. pKGri s.

..llRDenton.

e u.a.u s.,

2 /.28 / 75..

..263[75

• t/_1/15. _

ee,.

foram ABC 318 (Rev. 9 9)) ABCM 0340 W u. eleevenmuent passteme eepeese se,e.ese eee

WNP 1 and 4 SER Input 2.0 SITE CHARACTERISTICS 2.1 Geography and Demography 2.1.1 Site Location The site for the Washington Public Power Supply System's Nuclear Projects numbers 1 and 4 (abbreviated as WNP-1 and WNP-4, respectively) is located in Benton County, Washington within the former U. S. Atomic Energy Commission's Hanford Reservation. The site is about 8 miles north of the city of Richland, 2.5 miles west of the Columbia River at river mile 352, and 1 mile east of the WNP-2 facility, presently under construction.

Figure 2-1 shows a general view of the site location.

2.1.2 Site Description The site is relatively flat and semi-arid with sage brush interspersed with desert grasses.

The 972 acre site for WNP-1 and WNP-4 adjoins the 1089 acre parcel currently leased for WNP-2.

Figure 2-2 shows a view of the site itself.

2.1.2 Exclusion Area The applicant has selected as an exclusion area the envelope of the two circles each having a radius of 1930 meters (6400 feet) and centered on both the WNP-1 and WNP-4 con-tainments. Since WNP-4 lies some 3000 feet northwest of WNP-1, this produces an irregularly shaped exclusion area, as shown

2 t

on figure 2-2.

The applicant is making arrangements with the Energy Research and Development Administration (formerly part of the U.S.A.E.C.) for the lease of 972 acres of land lying within the exclusion area, and upon which the principal station structures are to be located. In addition, the applicant currently is leasing 1089 acres for the WNP-2 facility, which is contiguously located.

Part of the exclusion area lies outside the leased acreage and remains under the ownership of the Energy Research and Development Adminis tration (ERDA). The applicant has committed him-self to obtain a written agreement from ERDA, similar to one ob@ fined for.WNP-2, wherein the AEC stated that it recognized the existence of an exclusion area and would undertake no activities within this area which would interfere with or restrict the applicant's right to fully comply with this condition of the license. With the possession of a lease and this letter of intent from the AEC or its successor, we conclude that there will be reasonable assurance that the applicant has the authority, within the meaning of 10 CFR Part 100.3 (a), to determine all activities within the designated exclusion area. (This requires confirmation by OGC.)

The exclusion area is traversed by the mainline track of the Hanford Reservation railroad system, operated by the AEC, a railroad spur line leading to the FFTF f acility, and

\\

l 3

access roads leading to the WNP-2 and WNP-1 and 4 facilities.

None of these routes are open to the general public. The WNP-2 facility, presently under construction, is located within the exclusion area as is the proposed H. J. Ashe substation and the ( tisting Wye burial ground, a nine acre radioactive waste burial facility, located about a mile away. The only activities unrelated to plant op-erations within the exclusion area will be travel along the above routes and activities at the above-mentioned facilities.

2.1.3 Population and Population Distribution The nearest residents are located about 3 miles east of the site, on the bluffs overlooking the e,ast bank of the Columbia River. Table 2.1 shows the 1970 resident cumu-lative population as a function of distance from the site out to 5 miles.

Table 2.1 - 1970 Cumulative Population Distance, miles 0-1 0-2 0-3 0-4 0-5 Popula tion 0

0 0

38 138 Figure 2-3 shows the 1970 and projected 2020 cumulative population surrounding the site, as a function of distance from zero to fif ty miles.

For reference, the cumulative pop-ulation corresponding to a moderately populated region of

O 4

500 people per square mile-is also shown. Comparison of the curves indicates that the site area is not heavily populated. The 1970 population within 50 miles was 183,878 persons. The applicant projects this to increase to 295,538 persons by the year 2020. This corresponds to a population inctease of almost 10 percent per decade, as compared with a projected increase of 1.1 percent per decade for the recent Series E projections made by the U. S. Department of Commerce, Bureau of Economic Analysis (BEA) for BEA Area Number 156.

(See figure 2-4).

I?. S.

Census data indicate that Benton County, Washington had an 8.8 percent gain in population from 1960 to 1970. The applicant's projected population growth for the area within 50 miles of the site may be conservatively high.

The applicant has selected a low population zone (LPZ) radius of four miles, and the nearest population center, as defined by 10 CFR Part 100, is the city of Richland, Washington, located 8 miles south.

In 1970 there were 38 residents within the low population zone, while the population of Richland was 26,290 persons.

In addition to the resident population, there is projected to be a transient population within the low population zone of up to about 900 agricultural and industrial workers. About half of this transient population works either at the proposed

m e

e S-facilities or at the WNP-2 or FFTF facilities located within the low population zone. In addition, about 2000 workers pass through the low population zone twice per day traveling to and from their jobs in other areas of the Hanford Reservation.

2.1.4 Conclusions On the basis of the 10 CFR Part 100 definitions of the population center distance, the exclusion area and low population zone, our analysis of the on-site meteorological data from which atmospheric dilution factors were calculated (see Section 2.3 of this report), and the calculated potential radiological dose consequences of design basis accidents (discussed in Section 15.0 of this report), we have concluded that the exclusion area radius and low population zone and population center distances meet the guidelines of 10 CFR Part 100 and are acceptable if revisions to the containment spray system and containment leakage specification are provided as indicated in Section 6.2.3 and Section 15.1.

2.2 Nearby Transportation, Industrial & Military Facilities 5

~.

-6*

l 2.2.1 Nearby Transportation Routes The nearest road to the WMP-1 and WNP-4 sites is the Hanford Reservation Route 4 which is located 1.5 miles southwest of s

the site at its closest approach. The mainline trcck of the Hanford Reservation railroad systes passia abou; 2500 feet southwest of both proposed f acilities.

Postulating an explosion of the probable maxicum cargo of 7

high explosives carried by a truck located on Route 4 at its closest point of approach, we find that the rssulting peak overpressures are much lower than those produced by the design basis tornado. Postulating an explosiod

[

of the probable maximum cargo contained in one railroad boxcar located at its closest approach of 2500 feet would produce a peak reflected overpressure of 1.6 pai which might exceed that produced by the design basis tornado for Region III (as defined in Regulatory Guida 1 76).

e No explosives are presently transported on the Hanford i

L Reservation railroad system. However, in view of the proximity of the mainline track from the WNP-1, WNP-2, and WNP-4 facilitics, we will require that the applicant make errangements with the i

l railroad system to be notified of eny intent to ship oxplosives regularly or to ship more than several hundred pounds of high explosives by rail past the site, and to provide an analysis of any such change in usige to the NRC.

t

[

. _. _ _ _ _ _ _ _ _. _ _ J

O

' Other transportation routes in the vicinity include the Columbia River, located 2.5 miles east of the site, and State Highway 240, located seven miles southwest, We conclude that, due to the distances of these routes from the facilities, no significant hazards to the plant exists from these sources.

2.2.2 Airports There are no airports or commercial airways within 10 miles of the site.

The nearest airport, the North Richland airport, is located about 11 miles so.uth of the site. This field has hard surface runways, but no commercial facilities.

The nearest airport with commercial facilities is the Pasco airport located about 15 miles southeast of the site.

In addition, the llanford Reservation presently lies in an aircraft restricted zone (R-6715) with no private or commercial traffic permitted at altitudes below 10,000 feet without special permission. On the basis of the distances of the existing airfields from the site and previous staff studies we conclude that the WNP-1 and WNP-4 plants need not be designed with special provisions to protect them against the effects of an aircraft crash.

8-2.2.3 Military Facilities There are no military bases or high speed, low altitude military training routes within 10 miles of the site. The nearest military facilities are the Othello Air Force sta-tion located about 30 miles northeast of the proposed plants and'the Yakima Firing Range located more than 30 miles northwest of the site. On the basis of the re-moteness of these facilities from the proposed plants, we conclude that the activities at these facilities need not be considered in the design of the plaats.

2.2.4 Industrial Facilities and Pipelineo The only industrial facilities located within 5 miles of the proposed plants are those associated with other activ-ities on the Hanford Reservation. These include the WNP-2 plant, presently under construction, located about 1 mile away, the Fast Flux Test Fecility (FFTF) located about 3 miles away, a substation located about 1 mile away, and 2 small solid radioactive waste disposal burial sites, the nearer of which is located about a mile away. Based upon the distances of those facilities from the plants, and 6

(

0 9'

the types and quantities of materials stored there, we con-clude that the effect of an industrial accident or an in-advertent release need not be considered in the design of the proposed plants. We concluded earlier, at the time of the WNP-2 review, that an accident occurring at the FFTF would not result in doses which would prevent the safe shut-down of the WNP-2 facility. Based upon the similarity of control room designs between the WNP-2 plant and those for the WNP-1 and 4 plants, and, based upon the greater distance of WNP-1 and 4 from the FFTF than WNP-2, we conclude that design basis accidents at the FFTF facility would not result in doses which would prevent a saf'e shut-down of the WNP-1 and WNP-4 facilities. The WNP-2 facility has a postulated design basis accident source term which is similar to those postulated for the WNP-1 and 4 facilities and in view of the separation distance, interaction of these facilities is not a consideration.

There are no gas or petroleum pipelines or commercial storage facilities within 5 miles. There is on on-site fuel oil storage tank of 300,000 gallon capacity located about 350 feet from the reactor containment.

Based on its distance from other safety related structures, and staff analyses in similar cases, we conclude that the heat

r O,

from a fire at:this tank will not constitute a problem, while the redundant air intakes and isolation capability of the control room (to be verified in section 6.4) should be effective against smoke from such a fire.

Because the fuel oil unloading station is more distant from safety-re-lated structures than the tank, accidents associated with resupply of the tank should not present a more severe hazard than a fire in the tank itself. Finally, a massive rupture of the tank and its surrounding dike, caused by a common mode event, such as a seismic event, would cause no spilled oil to flow into the safety related structures since the tank is at a lower grade level than the surrounding structures. We conclude that the presence of the on-site fuel oil storage tank will pose no hazard to the plant.

2.2.5 Conclusions We have concluded that based upon the nature of'thIe nearby transportation, industrial and military facilities, in-cluding such factors ac their locations and distances relative to the plant, types and quantities of materials transported and stored, and the nature and frequency of activities carried on there, that potential hazards from these facilities are sufficiently remoto that they need not be considered in the design of the WNP-1 and WNP-4 plants.

O g

C A N A D

A

• f A

9tLLsNGHAM W

A, S

H I

N G

T 0 N l* "3

.4 (Vl#fff COWIOm

,}

,C,,,4 i

$1AffLt h

T, g

wtNATCHtt CHtte

[

CLI ILUM O

lb E

Mo$tlLArt l,$:.

3 giggy,ggg

=

(LLtNS8URO

.'I, LINO mount 3.s O

CthtLLO SANWR A V

l 4

i V,-

CONNtLL

/D

,AawA.

l e

)

v",C = ' 8 k'e "

.E iu~4ht.

1. • .T,t?,^0 4

p,aN, ohc;10518a,

)

't" ADAM 5

,,,Ca w

$tNNtw Cx

. agga g,gga f

COLDIN0Att fe.

• UsblLA /

k VANCOUVit (C

0" e

PCtftAND -

$0 8,

PIN 0Lif 0N DALLil O

R E

G O N AMENDMENT 8 JULY 1974 WASHINGTON PUBLIC POWER SUPPLY SYSTEM WPPSS NUCLEAR PROJECT NO,1 LOCATION OF !!ANFORD RESERVATION Preliminwy Safety Analysis Report i

U i

I I

l' I I

ii I: i l

li l' l' i ' lli l' g!

g i

8 i'

i-

!; I i;

l' i

.t.i 0

t.

s.i I

:I :

.r.

g g

41 a

a a

a et a

e a

i e,

.e,

.o

.m i

}.. -

i d

e m'

6 e

i e

d e;

e e

a e

e a

e e

e a

e a

n.

e e

a e

e r

a e

e a

e dl d

.I a

e l

l

.= i

\\..

)

l *7

)-

i

.....l.

i r

t

=..

t

,,,a. ;.

j j

7..-

e

..w.....a..

I I

t

{

9 a

i

\\

i i

i I

I i

l

\\.

-

• f --

I i

e.

i.

ev5 N.ih I

i

,..r'......,

.3....'.

'E

< % -.....no., t.

s e

1 l

m.w..

./.-

... i

...,a

--vr-

......,......- l g,

, pd.

%,n g.,,,,,,,,,;

-- ) g,,,g,.3 -

,..,g.

I il

/ s X

. t ~,.~,..

4 t

?p '. ---

. - ~ --

' p

• --+- - - -.. ;,.;,.; ;,c" - )s.

o

- ~

l

,[.

_,.,., 4..

t...i....

s-

\\. 1

... -4g

\\...

---.--t.

~,..t

~. e

4. ;

y-g 3 g.

9,,,.,.,.,,

l l

1' l

. u. I C....E.r[.

l l

f

.s.

l

{sh "'

\\

i g

i

'd i

l i

8

~.

...._M_PJ 1.a i

l N.

.W a

g e

(

. /i \\,_

j.

. \\. !

I I

/

t k =- l

'.5.

1

..\\.\\,t,/

t

'l l

l)_s 1 4.,..

..n.

.avn - m._

n-

/

1 A

l w.......

.i l

u _.i.- !

e i

-l I

/

4 r-r -y, -

- r-=

f i T

f. '.' *....\\

t l

[

~

..~. y...W g

l i;

I

\\

N I

. N

.....e 5

i I

....c......

/

r".....n'.=.*...

i l

l

\\

l l

s.

l'

/

l

\\--

i

,1

=

),.,..,.

. n..

9 s,

. b.s

...I l

l I

l 7

a.&..

I

,l,...

...-y..

.......g...

6

)

l e

i l

i i

g t

i t

l j

l I

i i

WASHINGTON PUOLIC PO\\

WPPSS NUCLEArt P Preliminary Safety 4

/

I

^

1ii f

'_iiIJ:

b

~

j

/

l n

/

/

/

/

/

./

6 10

/

~

/

w p.

7

/

/l-

/_

l

/'

oj

/

f

/

/

c e.

1b6

/

m y

/

s

/

~

s 3o f

/

[

l i

4 10 * ~l l

l 7

f

-l t

6 5

4

,i I

3 l-

,' 2 k

(

~

5

^

u.

I g i

I8II Il' lu3 to 20 30 40 50 DISTAf4CE F ROP.1 PLAtJT, MILES FIGURE f,3 CU550'LATIVE PO?ULATION DISTRIBUTION

~

4%

y

9

%J

~~~

Y Y

f

~

m I

%)/

i

, M

• f'

~

a,

~.,

,_1

~,

~-

~

~. ~

/\\

aYs.'r i

g y

,p s s.

~.

9-N')

)I~a

?

' N./..

75i!

Q

. ~

!. o t

p

~.

j.

E m

%/

(

u

-~

b

}

J f

s,.

s l

l 4

5, l

l 1.5S_.

f

• • m L

~

~~

.~

3

_p$

170

~

s

~ * * * *

• s 1

s

~~

l 60

~

S

)

• • 9 8eg ij gf

%, i C~m..

}.:

~ ~.

_ b '~ ( '

i N,

,x-tg

~~

Egwu 2.4 EE/}

A r4

d.. I C 6 t

Atl G.'! 7 tj t4 WPPSS #1 SER INPUT 3.5 TOP.NADO MISS1f.ES-

.s

!!c have reviewc41 t.he informatica t.upplied by the appliccnt in the l'SAR concerning thcJ r analysis of ternado generated missiles, and agree thct the spectru. of missiles J.icted below is e.cceptabJe ??o. ure in the design of the WPPSS 1 and 4 Station. The missile spectrum complies with that descr1 bed in WASH-1361.

MJculle Dittcacioal lleight Velocity A.

Uood PJcnk 4 in. x 12 in. x 12 ft.

200 lb 28') ft/sec E.

Steel pipe 3 in. diemoter, 10 ft. lonm 78 lb 234 ft/see schedula /,0

'. C. Steel rod 1 in, dia: toter x 3 Lt. long 3 lb 221 ft/see D.

Steel pipe 6 in. dinacter, 15 ft. loam 300 lb 210 Et/sec schedule 40 L

Staci pipe 12 in, diar.eter, 15 ft. long, 750 lb 175 ft/sec sche.dulo 40 v.

Uti1Jty pole 13.5 1.,.

diameter : 35 ft. long 1500 lb 130 ft/sec

?

G.

At:to. obile 20 ft' frontal area 4000 lb 70 ft/sec Missiles A through E are to be considered at all altitudes, and Missiles F and G at altitudes 30 feet above all grade levels within 1/2-mile of the facility structures. Our review indicates that these missiles can strike exposed surfaces at all orientations, horizontal as well as vertical.

t

o 3.5 TURBINE MISSILES The turbine generators on Units 1 and 4 have a peninsular orientation with respect to their containments. All safety-related systems are located outside of the low trajectory missile strike zone. In response to the staff's request, the applicant has performed an analysis with respect to high trajectory turbine missiles. We agree that for this

~

facility, the applicant's damage probability of 1.4 x 10 for each unit on the basis of an annual turbine design overspeed failure

-4 probability of 10 is approximately correct. Most vital equipment is protected by separation of redundant components and control areas are located under multiple floors of concrete.

In addition, we are currently performing a generic study on the matter of turbine missiles which may result in additional requirements for such items as the overspeed protection system and valve testing procedures. This aspect will be handled as a post-CP item.

i e

L i

i l

9 0

AIR CLEANUP SYSTEMS SECTION 6.2.3 The containment spray system is used for iodine removal from the contain-ment atmosphere following a postulated LOCA. Sodium hydroxide is added to the containment spray solution to enhance the iodine scrubbing function of the system.

The system proposed would add sodium hydroxide to all safety systems (including ECCS) taking suction from the Borated Water Storage Tank (BWST).

An. inadvertent actuation of this system would cause the addition of sodium hydroxide to the core internals, resulting in highly undesirable consequences.

The applicant has p'roposed manual overrides for "certain automatically operated devices" in the spray and/or spray additive system. Our concern is that these overrides might be used by an operator to defeat injection of the spray additive in the event of an actual accident.

If the additive is properly injected, we judge the spray would be effective in reducing iodine concentrations in the containment, and could have maximum removal constants ( A) of 10.0 hr for elemental, and 0.36 hr~

for particulate

~

6 iodine, in an effective volume of 2.47 x 10 cubic feet for off-site dose calculations. Without additive injection, the elemental removal constant

~1 would be decreased to 0.5 hr We have calculated off-site doses in the event of a postulated LOCA assuming effective sodium hydroxide injection (see Section 15.0).

However, in view of our concerns discussed above, we will require that the applicant redesign the system to remove the manual overrides, and redesign the piping so that sodium hydroxide would be added to the spray only, or propose any other such design that will satisfy our concerns. We will review this redesigned system and report our evaluation in a future supplement.

L

l

)

i 6.4 HABITABILITY SYSTEMS The following discussion is related to the emergency protection provisions of the control room with respect to General Design Criterion 19, Control Room, Appendix A, to 10 CFR Part 50, and to Regulatory Guide 1.75 and 1.95 that relate to toxic gas protection for centrol room operators. Other aspects of the control room ventilation systems are discussed and evaluated in Section 9.4 of this report.

6.4.1 CRITERION 19 The applicant proposes to meet Criterion 19 by use of concrete shielding and by installing a dual fresh-air inlet system containing a redundant charcoal filter train that supplies the control room with 1500 cfm of filtered air for pressurization. The filter also supplies 8000 cfm of filtered air to the switchgear and battery rooms.

The dual inlets consist of a set of local inlets that are on the north and south sides of the General Services Building at Elevation 517.

Make-up air is normally admitted to the control room through one of these local inlets.

e 4

9

-n.

--~

--.ww

t 2-A remote inlet is located about 1200 feet away from the containment structure, directly south in the case of Unit 1 and directly north in the case of Unit 2.

In the event of an activity release a radiation monitoring system, having detectors in all three inlets, is used to automatically select the inlet having the lowest activity concentration.

The other inlets are isolated and the make-up air that is drawn in by the operating inlet is diverted to the charcoal filter system and then into the control room for pressurization.

The design of the dual inlet system fails to meet our acceptance criteria for the following reasons:

1.

A failure of one of the automatic isolation dampers to open or to close will either prevent the supply of make-up air from the selected inlet or allow contaminated air to enter from one of the

" isolated" inlets. To ensure both flow and no-flow in each inlet requires a parallel set of two dampers in series or equivalent.

2.

Because the local inlets are positioned on a structure that is contiguous to the containment, there is no guarantee that they will remain free of contamination when the wind is carrying contamination in the direction of the remote inlet. This is due to wake affects that will cause general contamination of the areas adjacent to the containment. Because of this possibility, full m

-3_

credit for the dual inlet system cannot be given. We estimate the X/Q's for a long-term accident to be the following:

Time After Accident X/Q (sec/m )

-3 0-8 hr 1.5 x 10

-4 8-24 hr 9.5 x 10 1-4 days 3.1 x 10 '

4-30 days 1.1 x 10-The use of the above X/Q's result in unacceptably high thyroid and beta skin doses to the control room accidents after a LOCA.

Changes in the charcoal, filter system and/or inlet design are required to meet the dose guidelines of Criterion 19.

6.4.2 T0XIC GASES Co,ntrol room habitability following a postulated toxic gas release is essential to ensure that operators can shut down the plant and maintain it in a safe shutdown condition..

Chlorine has been identified as the only material that, if released, would pose a potential operator hazard. Provisions such as quick acting chlorine detectors and self-contained breathing apparatus will be provided to protect the operator against a chlorine release. We have reviewed these provisions against the requirements in draft Regulatory Guide 1.95 t

4-and have found the following deficiencies:

1.

The applicant has not specifically indicated the number of self-contained breathing apparatus nor the quantities of breathing air to be available for operator use.

2..

Amendment 11, page 6.4-2(a) indicates that the local inlets will be isolated upon a high chlorine alarm and the remote inlet operated to provide pressurization air. The amendment does not indicate how the system is controlled if all inlets are contaminated.

We conclude that the provisions to control postulated toxic gas con-centrations in the control room are unacceptable.

e e

15.0 ACCIDENT ANALYSIS The accidents analyzed by the staff in evaluating the effectiveness of the engineered safety features for the WNP-1 and WNP-4 plants include the hypothetical loss-of-coolant accident, a pump seal failure in the RHR system following a LOCA, a hydrogen purge of the containment, post-LOCA, a fuel handling accident, a rod ejection accident, and a rupture of a waste gas decay tank.

On the basis of our experience with the evaluation of the steam live break and the steam generator tube rupture accidents for PWR plants of similar des'ign, we have concluded that the consequences of these accidents can be controlled by limiting the permissible primary and secondary coolant system radioactivity concentrations so that potential offsite doses are small. We will include appropriate limits on primary and secondary coolant activity concentrations in the technical specif-ications at the time of the operating license review. The radioactive waste gas decap tanks are designed as Seismic Category I.

Therefore, the total failure of these tanks is sufficiently improbable that 10 CFR Part 100 guideline doses are applicable. Our calculations indicate that doses for failure of these tanks would be within Part 100 guidelines. Appropriate technical specifications will be placed on the maximum activity that can be stored in one tank at any time such that single failures of active components, such as lifting or sticking of a relief valve, will not result in radiological con-sequences that exceed small fractions of 10 CFR Part 100 guideline doses.

II

_i The rod ejection accident calculations indicate that substantial We will fractions of the Part 100 guideline doses could be realized.

reexamine this accide'nt at the operating license review stage and may impose restrictions on primary to secondary leak rates to insure that rod ejection ace,ident doses are well wi hin the exposure guide-t lines of 10 CFR Part 100.

15.1 LOSS-OF-COOLANT ACCIDENT The containment model used to describe the dose mitigating effects of the engineered safety features proposed for the WNP-1 and WNP-4 structure surrounding the plants includes a single containment reactor and a sodium hydroxide additive injection system operating The purpose of in conjunction with the containment spray system.

the sodium hydroxide additive injection system is to increase the iodine The removal capability of the spray following the hypothetical LOCA.

assumptions we used in evaluating the consequences of this accident are The results of the calculation indicate that with given in Table 15-2.

a reduced primary containment leak rate of 0.15%, the potential radio-logical consequences are within the guideline values of 10 CFR Part 100 As a result of and Regulatory Guide 1.4 and are, therefore, acceptable.

to reduce the primary this analysis, we will require the applicant leak rate from 0.20% per day to 0.15% per day so that the containment doses will conform to the guideline values of Regulatory Guide 1.4.

?

s 3

As part of the loss-of-coolant accident (LOCA), we and the applicant have also evaluated the consequences of leakage of containment sump water containing radioactive fission products which is circulated by the ECCS outside the containment after a postulated LOCA.

.We have assumed the sump water to contain a mixture of iodine fission products in agreement with Regulatory Guide 1.7.

If a source of leakage should develop, such as from an RHR pump seal, a portion of the iodine would become gaseous and would exit to the outside atmosphere. The offsite doses resulting from such a sequence of events depends upon the temperature and magnitude of the assumed leakage.

If the leakage occurred when the water temperature was below 2120F, a leak rate of about 20 gpm over a period of one-half hour would result in doses (without filters) which could exceed the guide-

.line values of Regulatory Guide 1.4 from this source alone.

If the leakage occurred when the fluid is near its peak temperature of 270 F, then part of the leaking water would flash to steam, leading to additional iodine release.

In this case, about 3 gpm leakage for 30 minutes would result in doses (without filters) which could exceed the Regulatory Guide 1.4 values.

If the area where the ECCS equipment is located is served by filters effective in removing iodine, the offsite doses from possible pump leakage in this area would be a small contributor to the LOCA dose, even for substantial amounts of leakage. Based on the above analysis,

4 we will require that the area where the ECCS equipment is located be served by filters which are effective in removing iodine and which conform to the requirements of engineered safety feature (ESF) filters.

15.2 HYDROGEN PURGE DOSE ANALYSIS The applicant has provided redundant recombiners for the purpose of controlling any formation of hydrogen after a design basis LOCA.

In the event of failure of both recombiners, the applicant has

~

provided a backup purging mode. Purging would be provided to the Containment Purge Filtration system to minimize the radiological con-sequences of purging. We have evaluated the additional dose an individual might receive due to purging the containment after the design basis accident. Our assumptions are listed in Table 15-3 and the calculated doses are listed in Table 15-1.

We find the calculated doses are within the guidelines of 10 CFR Part 100 even without the use of filters.

15.3 FUEL HANLIING ACCIDENT We have evaluated the radiological consequences of a fuel handling accident. Our assumptions are listed in Table 15-4 and the calculated doses are listed in Table 15-1.

We find the calculated doses to be well within the guidelines of 10 CFR Part 100.

P 5

15.4 CONTROL ROD EJECTION ACCIDENT We have evaluated the consequences of a rod ejection accident.

The assumptions used by the Regulatory staff to calculate offsite doses from a control rod ejection accident were:

Case I 1.

Power level of 3760 Hut.

2.

46% fuel failed in transient.

3.

10% of iodine and noble gas inventory in gap of failed fuel.

4.

Release of total gap activity in failed fuel to containment building.

5.

50% plate-out of radioactive iodines.

6.

Containment building sprays are not initiated.

7.

Containment building leak rate of 0.20%/ day for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and one-half of this value thereaf ter.

8.

Standard ground level release meteorology and dose conversion factors.

Case II 1.

Power level of 3760 Mwt.

2.

46% fuel failed in transient.

3.

10% of iodine and noble gas activity in gap of failed fuel.

4.

Release of total gap activity in failed fuel to primary coolant.

5.

Primary to secondary coolant operational leakage is 1.0 gpm.

l 6.

Loss of offsite power so that steam is released from secondary side relief valve.

7.

Primary-secondary coolant equilibrium reached at 30 minutes af ter l

the accident.

6 15.4 CONTROL ROD EJECTION ACCIDENT (Cont'd.)

8.

Standard steam line release meteorology and dose conversion factors.

The calculated doses are listed in Table 15-1.

We find the cal-culated doses to be within the guidelines of 10 CFR Part 100.

15.5 CAS DECAY TANK RUPTURE We have evaluated the consequences of gas decay tank f ailures and determined that the doses are within the whole body dose guidelines of 10 CFR Part 100.

The assumptions utilized are found in Table 15-5.

t t

TABLE 15.1 Potential Offsite Doses Due to Design Basis Accidents Two-Hour Course of Accident Exlcusion Boundary Low Population Zon (1950 meters)

(6440 meters)

Thyroid Whole Body Thyroid Whole Body Accident (rem)

(rem)

(rem)

(rem)

Loss-of-Coolant 150 7

51 2

Post-LOCA Hydrogen Purge Dose 82 9.2 Fuel Handling 5

2

<1

<1 Rod Ejection

• Case I 100

<1 54

<1 Case II 120 3

Gas Decay Tank Rupture Negligible 20 Negligible 2 See Section 15.4 of this report for the different assumptions used for Cases I and II.

I I

l r

u-

I P

TABLE 15-2 Assumptions Used in the Estimate of Design Basis Accident Doses LOSS-OF-COOLANT ACCIDENT Power Level 3760 MWt Operating Time 3.0 years Reactor Building Leak Rate (0-24 Hours) 0.15%

~

( >24 hours) 0.075%

Iodine Composition Elemental 91%

Particulate 5%

Organic 4%

Minimum Site Boundary Distance 1950 m Low Population Zone 6440 m 3

X/Q Values (sec/n )

0-2 @ 1950 meters 3.0 x 10-0-8 @ 6440 meters 2.8 x 10

-5 8-24 @ 6440 meters 1.9 x 10-6 24-96 @ 6440 meters 8.3 x 10-6 96-120 @ 6440 meters 2.5 x 10 SPRAY EFFECTIVENESS Max. Elemental Iodine Decontamination Factor 100 0

Effective Volume 2.47 x 10 ft Elemental Iodine Removal Coefficient (Injection phase) 10.0 hrs-Particulate Iodine Removal Coefficient 0.36 hrs-Organic Iodine Removal Coefficient 0

E e

s TABLE 15-3 Hydrogen Purge Dose Input Parameters Power Level (MWt) 3760 0

Volume of Containment (f t )

3.09 x 10 Purge Duration (days) 30 Holdup Time in Containment (days) Prior to Purge Initiation 15 Purge Rate (SCFM) 80

-6 4-30-day X/Q (sec/m )

2.5 x 10 e

t l

e TABLE 15'-4 FUEL HANDLING ACCIDENT CALCULATION INPUT PARAMETERS Shutdown Time 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Total Number of Fuel Rods in the Core 54120 Number of Fuel Rods Involved in the Refueling Accident 264 Power Peaking Factor 1.65 Iodine Fractions Released from Pool Elemental 75%

Organic 25%

Effective Filter Efficiency Elemental 95%

Organic 95%

X/Q Values, Sec/m 0-2 hours @ 1260 meters 3.0 x 10 '

~

0-2 hours-@ 3200 meters 2.8 x 10-5 O

e e

w

--g-,.--

-n

TABLE 15-5 Gas Decay Tank Rupture The assumptions used to calculate the offsite doses from a gas decay tank rupture were:

1.

Gas Decay Tank contains one complete primary coolant loop inventory of noble gases resulting from operation with 1% failed fuel (1,290,000 curies of noble gases).

2.

X/Q values based on on-site meteorological data. (0-2 hours

@ 1950 meters, 3.0 x 10 '. 0-8 hours @ 6440 meters, 2.8 x 10-5),

~

t 9

9 e-,-

n

-,,e--

.,--,s

+

s

-- -