Text

Proposed Tech Specs Re Post Accident Access to Vital Areas (Plar 3-98-6)
	ML20249A281
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Site:	Millstone
Issue date:	06/10/1998
From:	; NORTHEAST NUCLEAR ENERGY CO.
To:	;
Shared Package
ML20249A274	List: B17308, Application for Amend to License NPF-49,by Incorporating Attached Proposed Rev Into Chapter 12 of Millstone Unit 3 FSAR; ML20249A281;
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~ 4 Docket No. 50-423 B17308 --

Millstone Nuclear Power Station, Unit No. 3 Proposed License Amendment Request Post Accident Access to Vital Areas (PLAR 3-98-6)

Marked Up Pages June 1998 l-9906160265 990610 i

4 PDR ADOCK 05000423 i P

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U.S. Nuclear Reguletory Commission B17308%ttachment 2\\Page 1 MARKUP OF PROPOSED REVISION Refer to the attached markup of the proposed revision to the Final Safety Analysis Report (FSAR). The attached markup reflects the currently issued version of the FSAR.

The following FSAR changes are included in the attached markup.

Chapter 12 - List of Tables - Updated Section 12.3.1.3.2, - Deleted and replaced with Attachment 2A Table 12.3 Deleted and replaced with Attachment 2B Table 12.3 Deleted and replaced with Attachment 2C Figure 12.3-10 (5 cheets)- Deleted and replaced with Attachment 2D l

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H MNPS-3 FSAR CHAPTER 12 LIST OF TABLES i

TABLE NUMBER TITLE 12.2-1 Parameters Used in Calculation of Design Radiation Source inventories 12.2-2:

Radioactive Sources in Containment Building 12.2-3 isadioactive Sources in the Auxiliary Building 12.2-4 Radioactive Sources in the Waste Disposal Building

'12.2-5 Radioactive Sources in the Fuel Building 12.2-5A Other Radioactive Sources 12.2-6 Inventory of an Average Fuel Assembly after 650 Days of Operation at 3,636

~

Mwt at Shutdown and 100 Hours After Shutdown (gCi) 12.2-7 Source Intensity in the Most Radioactive Fuel Assembly

After 650 days of -

Operation at 3,636 Mwt 12.2-8 Radionuclides Concentrations in the Spent Fuel Pool From Refueling 100 Hours After Shutdown

12.2-9 Radiation Sources

Reactor Coolant Nitrogen - 16 Activity 12.2-10 Assumption Used in the Calculation of Airborne Concentrations 12.2-11 Airborne Concentrations inside MEjor Buildings ( Ci/cc) 12.3-1 Radiation Zones 12.3-2 Radiation Monitoring System - Area Radiation Detector Location

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12.5-2 Health Physics and Chemistry Laboratory Equipment py

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$ A SG S L<.S,% o h G.& PL hC C' w.r7H ATi?)c H H Ed'i'c)b-MNPS-3 FSAR 12.3.1.3.2 Post-Accident Access to Vital Areas 1

q,,p REG 0737 requires that following a design basis accident there exists adequate access yqy to 'tal plant areas (consistent with the dose guidelines of GDC 19). At Millstone 3, this requir ont is met by providing sufficient material thickness to shield components containi post-accident radioactive inventories, consistent with anticipated access 97 3t l requireme s and stay times. Established procedures are also available to provide g de-lines for im menting effective controls. Post-accident access to vital plant area, routes to these area distances travelled along these routes, assumed walking speed, d

occupancy tim in the vital areas are defined on Figure 12.3-10 and in Table 2.3-3 and 12.3-4. Dose ra.s along these routes and at manned locations within the v' al areas are f

calculated in orde to assure that anticipated integrated doses to personn from post-accident radiation s urces are below specified limits.

The following location are identified as requiring access following accident:

1.

Auxiliary B ' ding - rod drive and motor control c ter areas (electrical areas) on elevation 4 feet-6 inches and elevation 43 >et-6 inches; 2.

Main Steam Val Building - access to manu I override atmospheric steam reliefs; (Ohi\\>

3.

Hydrogen Recombin Building - access control area, sampling room, recombiner cubicles, a d hydrogen an zer panels; 4.

Fuel Building - access re ired to r lign fuel pool cooling; and 5.

Sample Analysis Area - Mill o 3 chemistry laboratory. If uninhabitable Qo%

(high airborne activity), Millst e 1 and 2 chemistry laboratories have been identified as alternate samp -

atysis areas for Millstone 3 accidents. If necessary, the EOF also sa MMA spectrometer system.

6.

Control Building swit gear area - lant equipment operators are directed q 7.g here to restore po r to MCC 32-3 7.

Turbine Buildin 38 ft. level to compte actions of restoring MCC 32-3T.

l 8.

Diesel Ge rator enclosures to reset MC Breakers for EDG keep warm 1

systems Dose rates and do es within these areas, including contribu n enroute to these areas, are determined and ielding requirements specified. Does rates d doses for areas 6-8 were not determine as no piping sources are in the immediate vicini of these areas. The only 91Wt direct source f radiation would conceivably be the MP-3 contain ent building. As stated I

j

( previously, oes rates from containment shine are negligible and t 'refore were not i

/

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1253 MP3 12.3-6 January 1998 l

MNPS-3 FSAR

~

{corn ted for these locations. It was therefore determined that no specific doese calc a-n.g l

tions ere required for areas 6-8. In addition to the areas defined above, the follo g

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addition I areas require post-accident access and continued occupancy:

l(teki) 1.

Control Room - post-accident control room habitability is disc sed in Section

.4, and control room post-LOCA doses are presented in S ction 15.6.5; and j

l 2.

Te hnical Support Center - post-LOCA dose evaluatio is presented in W5 Sec 'on 15.6.5.

D lt has been determin that post-accident access to areas ad essed in NUREG-0737 which have not Leen i entified above is not required for Mill onc 3.

4 Following an accident, it assumed that accident mitiga on personnel would exit the p.9 control building and travel long predetermined routes t the previously identified vital

{

areas at an assumed walki speed of 3 feet per sec

d. These predetermined routes (primary routes and, where a plicable, alternate rout.s) are described below.

1.

Auxiliary Building. The primary rou is from the control building through the service building co ' dor to the au iary bay and then across the road to the auxiliary building (Pn ary Route o.1 on Figure 12.3-10). The alternate route is from the cont i buildi through the service building (past the radioactive chemistry la orat y), followed by a right turn just beyond the count room, then out alo e roadway to the auxiliary building (Alternate Route No.1 on Figure 12.' 10).

2.

Main Steam Valve Buil ing.

e primary route is the same as Alternate Route No.1 as far a he entra e from the roadway to the auxiliary build-ing, beyond which int the prim y route to the main steam valve building follows the roadw y to the end of iis building, around the corner to the ladder, and then p this ladder to th entrance of the steam relief area (Primary Rout

o. 2 on Figure 12.3-1

). The alternate route (Alternate l

Route No. 2 n Figure 12.3-10)is from e control building through the service buil ing corridor to the auxiliary b y, across the auxiliary bay to the railroad lo ding area and out through the r ling steel door, along the railroad tracks t the auxiliary bay roadway, and the along this roadway to the main steam alve building ladder leading up to the tranceway into the steam relief rea.

3.

Hy rogen Recombiner Building. The primary route Primary Route No. 3 on Fi ure 12.3-10)is from the control building through he service building (past

.t e radioactive chemistry laboratory), followed by a r ht turn at the corridor l

/ eading to the roadway, then along the roadway beyon, the main steam valve building and over to the entrance to the hydrogen recombiner building.

The alternate route (Alternate Route No. 3 on Figure 12.3-10)is from the

, control building through the service building corridor and through the auxilia-ry bay to the railroad loading area, then out of the loading area and across to the entranceway to the hydrogen recombiner building.

l l

12S3.MP3 12.3-7 January 1998 \\

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o MNPS-3 FSAR Primary and alternate Routes 3a are essentially the same as primary d alter-kbi nate Routes 3, respectively, except that Routes 3a are for access t the H,

)

analyzer rather than the H recombiner.

1 2

Dok)l 4.

Chemistry Laboratories. The primary route (Primary Route

'o. 4 on Figure 12.3-10)is from the control building via Primary Route No 3 to the hydrogen recombiner building and from the hydrogen recombiner ilding to the Unit 3 hemistry laboratory, backtracking along primary Rout 3. Primary Route

. 4 is followed by an emergency worker who, hav' g drawn samples in th hydrogen recombiner building, then transports ese samples to the Mill one 3 chemistry laboratory for analysis. Alt rnate Route 4 is essentially the s e as alternate Route 3.

In case the nit 3 chemistry laboratory is uninh itable because of high airborne radioactivity, he Unit 1/ Unit 2 chemistry labor tory serves as an alternate.

holo)

Alternate route a is one of the routes to this aboratory. It is the same as primary Route 4 r alternate Route 4 untille ving the hydrogen recombiner building. The ope tor would then head w st toward the Unit 3 turbine building and then south to t Unit 1/ Unit 2 che 'stry laboratory. This route encounters stairs and/or an elev or and therefore ould probably eliminate the use of the lead pig. Alternate 4b 's another rout to the alternate laboratory. It would involve the use of a tru and forklif to move the. shielded sample, therefore, would result in lower dos s than al rnate Route 4a.

5.

Fuel Building. The prima r te (Primary Route No. 5 on Figure 12.3-10)is from the control building a g the north side of the plant around the boron test tanks to the fuel buil n. The alternate route (Alternate Route No. 5 on Figure 12.3-10)is throu i th ervice building to the auxiliary bay, through the auxiliary building b-to the ailroad loading area, then east along the roadway adjacent to e transfor er area, and finally around past the hydrogen recombin building and gineered safety features building to the entrance to the fu building.

e post-accident doses witi n the vital areas previou ly identified and the post-accident d)ses enroute to and from ese vital areas have been termined based upon liquid and gaseous source terms def' ed in NUREG-0737 as follows.

1.

Liquid-co taining systems - 100 percent of the ore equilibrium noble gas invento,50 percent of the equilibrium iodine i entory, and 1 percent of all rema' ing core fission products are assumed to b mixed in liquid systems whi are pressurized. Depressurized liquid source (such as containment su p water) are assumed to contain 50 percent of t e core equilibrium io ine inventory and 1 percent of the remaining core sion products exclud-i, g noble gases.

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2.

Containment atmosphere - 100 percent of the core equili ium noble gas f

inventory,50 percent of the core equilibrium iodine activit and 1 percent of all remaining core fission products are assumed to be releas d to the contain-ment. Of this release, the 1 percent of all remaining core fision products is washed down into the sump. Fifty percent of the iodines relea ed into the l 12sa. sara

-19.3 8

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JMuary 1998

MNPS-3 FSAR

- containment (i.e.,25 percent of the core iodines) plate out on con inment internal surfaces, leaving the remaining 50 percent of the release iodines (or 25 percent of the core iodines) to be mixed in the containment mosphere.

3.

Gas-containing systems - Based upon (2) above, gas-contai 'ng systems are assumed to contain 100 percent of the core equilibrium n le gas inventory and 25 percent of the core equilibrium iodine activity.

To determine ost-accident doses to personnel going to and worki).g in the previously entified vital reas, the following sources of radiation were co idered in the Millstone 3 ost-accident a ess analysis.

1.

Auxi ry Building R diation from containment atmosp ere shining through electrical pe trations.

Radios ine buildup in the SLC filter.

Sump wa r in the safety in' ction system piping located below the elevation 2 feet-6 inch fl or.

Containment a osph e shine through the personnel hatch scattered into the MCC an ro control area.

2.

Main Steam Valve Buildi

- Radiation from containment atmosphere shining through the main stea line netrations.

3.

Hydrogen Recombi r Building Radiation om containment mosphere contained within the active hydroge recombiner reaction amber and blower.

Radi non from containment atmos here and condensate contained in pipi carrying containment atmosp re to the active hydrogen re mbiner building.

adiation scattered out of the active reco biner cubicle entrancewa) to the recombiner control area.

4.

Fuel Building p

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Direct shine frorn containment.-

l Shine from the personnel hatch in the auxiliary buildin,

j l

=*

Shine from the RHR heat exchanger in the ESF building.

l; l

Shine from the~ fuel pool cooling pumps.

12ss.un 12.3-9 January 1998 l-

MNPS-3 FSAR

~

s l

5.

Along routes from control building to the vital areas.

a Skyshine from containment.

Direct shine from containment.

Systems co aining sources of radiation which are identified in NUREG-0737 b which have not been 'dentified in buildings discussed above are considered to be e' er irrelevant following an ac ' dent or negligible ccatributors of radiation following an a ident. For example, the RH nd RSS systems do not exist in any of the areas into hich post-accident access is quired and are, therefore, considered irrelevant w' h respect to radiation dose contri tion. Also, the GWS system is a negligible co ributor of radiation following an accident cause when the accident occurs, the letdo n is automatically isolated.

To determine the feasibility f gaining post-accident access to e vital areas previously identified, dnses and dose rat within the vital areas and dos s enroute to and from these vital areas have been calculate The doses along the route previously identified were calculated based upon the follows assumptions.

1.

All doses to the out-of-or portions of t routes are due to skyshine from the containment and dire shine from t e containment dome.

2.

The integrated dose along ea rout is based upon the out-of-doors time spent along the route and the c rre ponding maximum dose rate at the time following the accident that the to te is being traveled.

3.

The walking speed is assum,e o be feet per second. In addition, doses irl the vital areas were calculat d based u on the following assumptions.

a. Post-accident starting ti e from control bu ing:

to auxiliary buildi g -immediately following e accident;

to main ste valve building - 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months following e accident;

to hydr g,eh ecombiner building to start hydrogen re ombiner - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months follo i g the accident; oh rogen recombiner building for operation of the hydr en analyzer imrnediately following accident;

/ to hydrogen recombiner building to draw samples -immediately ollowing

)/

the accident and again at 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following the accident;

to fuel building - 15 hours1.736111e-4 days 0.00417 hours 2.480159e-5 weeks 5.7075e-6 months following the accident.

1

b. Indoor occupancy time in vital area:

6 T.xiliary building - 10 minutes;

' l 1rsa.urs 12.3-10 January 1998

MNPS-3 FSAR in main steam valve building - 10 minutes; in hydrogen recombiner building for recombiner operation - 20 minutes in yrinth area to operate valves and 40 minutes at recombiner ps in the (ydrogen recombiner building to turn the hydroge alyzer on -

e 20 minut s in the sample cubicle labyrinth to opera valves and 5 minutes t the analyzer panel to turn the an zer on; in hydrogen re mbiner building to dr samples - 12 minutes at SSP remote panel,6 inutes near control panel, 5 minutes in hydrogen cubicle labyrinth,1 inut in sample cubicle and sample cubicle laby-rinth for valve lineup a operation, and 5 minutes for sample retrieval; in fuel building realign fue col cooling - 45 minutes.

The assumptions discuse above are summarized i Table 12.3-3, and the calculated doses and dose rate re tabulated in Table 12.3-4. in dition, maximum dose rates in the i

vital areas are sh n on Figure 12.3-10 route maps. The sults of the post-accident l

access and st ' ding study for Millstone 3 show that no eme ency worker would be I

exposed to ost-accident doses greater than the GDC 19 dose g eline of 5 rem as a (lohD result o raveling to, working in, and returning from any of the Mills ne 3 post-ac.cident f

vital 4 eas.

12.3.2 Facility Design Features

]

The Millstone 3 design is consistent with the guidance presented in Regulatory Guide 8.8, Revision 4, C2, which discusses specific features in the facility and equipment design that limit radiation exposure to levels that are ALARA. The following features have been incorporated.

12.3.2.1 Location and Design of Equipment to Minimize Service Time in the auxiliary building, nonradioactive equipment, such as the reactor plant component cooling systern and components used to process the waste evaporator distillate, are I

located outside high radiation cubicles in areas designated as Radiation Zones 11 or ill (defined in Table 12.3-1). In the containment structure, nonradioactive equipment requiring servicing is typically located in Radiation Zone IV areas. Exceptions include those [~M components attached to the reactor coolant system, such as the reactor coolant pump motor cooling equipment and the equipment support snubbers.

Major radioactive components which may require servicing are typically located in individu-l WH ally shielded cubicles. These cubicles are designed such that radiation contributions from adjacent cubicles is small compared to sources within the cubicle. The resultant dose rate in any cubicle in which equipment is being serviced is due to sources within the cubicle to g,g radiation penetrating through shield walls from adjacent cubicles, and to radiation stream-ing through shield wall penetrations. The design basis for shield walls enclosing cubicles containing process equipment is discussed in Section 12.3.1. Shield wall arrangement and (k n,A dimensions are shown for the Containment Building, Figure 3.8-60; Auxiliary Building, QMs)

Figure 3.8-62; Fuel Building, Figure 3.8-63; and Waste Disposal Building, Figure 3.8-74.

uss.ws 12.3-11 January 1998

i

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MNPS-3 FSAR TABLE 12.3-3 r

POST-ACCIDENT U VITAL AREAS -

ROUTES, START I

WALKING SPEED, OCCUPANC b ES c-D apLe w;A AdiMM g,,g I

Indoor Assumed Ocdupancy Outdoor Post-Assumed Time Round-Trip Accident Walking in Vital '

Route,

Distance Starting Speed Area n

Vital Area Number (ft)

Time thr)

(ft/s (min)

Auxiliary Buildino RCC and MCC Pn ary areas on 24'6" 1

112 0

3 10 el. and 43'6" Alterna el.

1 122 3

10 Main Steam Valve Buildina Access to man-

' Primary ually override 2

310 5

3 10 atmospheric Alternate steam reliefs 2

5 5

3 10 t

' Hvdroaen flecombiner Buildina Access to con-Primary trol area for 3

740 24 3

recombiner Alt nate operation 3

700 24 3

'88 Access to H, Primary l

analyzer for Sa*

740 0

3 W

analyzer oper-Alternate ation 3am 700 0

3 ccess to sample Primary 1,/ g 7 I

.. rea in HRB then 4

740 0'2' 3

'5' t chemist:y

. Primary i

la ratories for 4

740 24 3

'5' sam analysis Alternate 4

700 0

3

'5'

- uss-swa 1 of 4 December 1994 I a

.+

MNPS-3 FSAR g

POST-ACCIDENT ACCESS TO VITAL AREAS -

ROUTES, STARTING TIMES, WALKING SPEED, AND OCCUPANCY TIMES Assumed Indoor Assumed Occupancy.,

Outdoor Post-sumed Time Round-Trip Accident Walking in Vital Route,

Distance Starting Speed Area n

Vital Area Number (ft)

Tire.e (hr>

Jft/s)

(min)

Iternate 4

700 4

3 Al rnate a

1025 0

3 Alter te 4a 1025 24 3

Fuel Buildina Acc6ss required Primary to realign fuel 5

330 15 3

45 pool cooling Alternate 5

30 15 3

45 NOTES:

1.

See Figure 12.3-10 for c plete route desc *ptions to vital areas.

2.

The route from the c trol building to hydroge recombiner building to Unit 3 chemistry N

laboratory must be aveled at two separate tim

,t

= 0 and t = 24 hr post-accident.

3.

Indoor occupan time in the HRB for access to co trol area for recombiner operation consists of:

1.

a.

20t mutes in labyrinth area to operate valves fo recombiner operation, and b.

minutes at recombiner panel.

4.

In or occupancy time for H, analyzer operation consists of:

20 minutes in the labyrinth area to operate valves for the nalyzer operation.

a.

i letty b.

5 minutes at the panel to turn the analyzer on.

L 5.

Indoor occupancy time in the HRB for access to sample area consists of:

~~ N l 12 saws 2 of 4 December 1994 d

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MNPS-3 FSAR MBLE 12.3-5 POST-ACCIDENT ACCESS TO VITAL AREAS -

ROUTES, STARTING TIMES, WALKING SPEED, AND OCCUPANCY TIMES a.

15 minutes - s tem warm-up and valve lineup check.

1) 6 minutes at P remote panel.

2) 2 minutes to dress 'n protective clothing.

3) 2 minutes in sample c bicle to check valve lineup.

4) 1 minute to remove prot tive clothing.

5) 4 minutes to operate valves 'n sample cubicle labyr' th area.

t b.

10 minutes at SSP control station for ' itial valve line p.

c.

30 minutes at SSP control station for ini i purge. sample collection, pure water flust, isolation, and additional purging and flushi d.

5 minutes in hydrogen recombiner cubicle la. rinth to isolate reactor coolant influent valves.

se/r7 e.

10 minutes to retrieve samples.

1) 2 minutes to dress in protecti clothing.

2) 5 minutes in sample cubici to collect samples.

3) 3 minutes to remove pr tective clothing and prepa e samples for delivery to chemistry laboratory.

The samples can now b taken back to the lab. Remainin steps are optional: only necessary to facilitat uture sampling, f.

13 minutes at SS remote panel and SSP control station for a ditional purging of sampling area.

g.

10 minutes r valve operation.

1) 2 inutes to dress in protect ve clothing.

2) minutes to restore valve lineup to normal.

3 minutes to remove protective clothing and completion of tasks.

l 12ss-a.ws 3 of 4 -

December 1994l

f l

t MNPS-3 FSAR TABG 12.3-3 POST-ACCIDENT ACCESS TO VITAL AR S-ROUTES, STARTING TIMES, WALKIN

PEED, AND OCCUPANCY TIM NOTE:

The above ta are necess for obtaining a liquid sample. Containment air can be done ad ional personnel at the same time; it would involve approximately the exposure to the additional person (s). If done without the liqui, it wou e lower since they would not have to waitfor some of the ore time-consu tasks that are necessary for the liquid samples.

l'If7 6.

Alternate Routes and 4b include travel from the contro uilding to the HRB to the Unit 1/ Unit 2 chem' try laboratory. Travel back to the control bu ing is not accounted for.

The workme are assumed to leave the site from the chemistry oratory.

7.

Routes rimary and alternate 3a are the same as primary and alternat 3, respectively, inside the HRB The hydrogen analyzers are located in the upper vation equip-exce m t room.

. l irs >sup 4 of 4 December 1994

.hL MNPS-3 FSAR

- f TABLE 12.3 4 POST-ACCIDENT ACCESS TO VITAL AREAS -

L[k

_ ROUTES. DOSE RATES, AND TOTAL DOSES Q&U M

Indo r Outdoor Dose-Dose Indoor Outdoor Total Route,,,

Rate' Rate'8' Dose Dose Dose Number (Rem /hr (Rem /hr) 1R_em]

[R ml fBgsl e

f Primary s

1 28.5 11.0 4.75 0 2 4.87

. Alternate 1

28.8 11.0 4.75 0.11 4.86

' Primary 1.2m o,o4cm 2

15.6 1.5 2.6 0.044 2.68 Alternate 2m 0.040m 2

15.6 1.

2.6.

0.082 2.72 Primary 3

O.10 0.049 0.007 0.056 Alternate f

3

'* 8

.10 0.049 0.006 0.056 Primary

( 3a 14.0 0.732 0,959 1.691 Alternate 3a

'S 14.0

0. 32 0.907 1.639 Primary 4 (t = 0 hr)

'5' 14.0 0.08 0.960'"

1.04 I

L Primary 4 (t = 24 hr)

O.10 0.004 0.007'S 0.011 Alternate 4 (t = 0 ir) 85) 14.0 0.083 0.907'"

0.99 Alte ate

. 4) = 24 hr) 0.10 0.004

.00 6'"

O.010 i

Alternate 4a (t = 0 hr)

'5' 14.0-0.083 1.33'"

1.41 L

f 12m m 1 of 3 Decemogr 1994 l

t..

MNPS-3 FSAR TABLE 12.3-4 POST-ACCIDENT ACCESS TO VITAL AREAS -

ROUTES. DOSE RATES, AND TOTAL DOSES Indoor Outdoor Dose Dose Indoor utdoor Total t2 y

Route,\\,

Rate :

Rate' 8 Dose Dose Dose Number \\

(Rem /hr)

IRgml IBaml IBaml x

\\

p7 Alternate 4a (t = 24

)

O.10 0.04 0.009'"

O.013 Primary 5

0.040 0.25 0.030 0.030 0.060 Alternate 5

0.040 0.25 0.030 0.036 0.066 NOTES:

1.

See Figure 12.3-10 f complete rout descrit;tions to vital areas.

2.

Maximum dose rate to i ividual w ile inside building in vital area.

3.

Maximum dose rate receive o path while traveling to vital area.

Ul81 l 4.

Indoor dose rates in HRB at 24 hr for recombiner operation.

"IU l 143 mrem /hr in r combine valve operating labyrinth a.

b.

2.2 mrem /hr t recombiner pa el 5.

Indoor dose rat in HRB at t = 0 hr fo taking samples:

a.

2200 Rern/hr in sample valve opera ng labyrinth

' b.

2. 5 mrem /hr at SSP control station ofq 0.0 mrem /hr in sample cubicle while prepar g to obtain samples c.

143 mrem /hr in recombiner valve operating lab inth, j

3.1 mrem /hr at containment air SSP remote panel e.

f.

2.2 mrem /hr at liquid SSP remote panel.

6.

Indoor dose rates in HRB at t = 24 hr for taking samples 91.5 mrem /hr in sample valve operating labyrinth a.

l12sutaP3 2 of 3 December 1994

MNPS-3 FSAR TABLE 12.3-4 POST-ACCIDENT ACCESS TO VITAL AREAS -

ROUTES. DOSE RATES. AND TOTAL DOSES b.

O.1 mrem /hr at SSP control station c.

0.0 mR

/hr in sample cubicle while preparing to obtai samples d.

6 mrem /hr i recombiner valve operating labyrinth le/7 e.

O.13 mrem /hr at ontainment air SSP remote nel f.

O.1 mrem /hr at liquid SP remote panel 7.

The outdoor dose rate of 1.2 Re r and out or dose of 0.040 Rem account for climbing the MSV Building ladder for access o the vi I area. (A climbing time of 1 minute is assumed.)

8.

Indoor dose rate in the HRB at t = 0 f

analyzer operation.

a.

2.2 Rem /hr in the sample I yrinth for Ive operation b.

- 20 mrem /hr at the ana zer panel 9.

This includes dose from c ing the sample. Dose r e at t = 0 hr is 7.7 Rem /hr and at t = 24 hr is 55 mrem Integrated doses are:

a.

250 mrem

!II7 b.

1.8 mRe f

c.

450 Rem d.

3 mrem e.

480 mrem f.

3.4 mrem I 12*xwa.

3 of 3 December 1994 l

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12.3.1.3.2 Post-Accident Access to Vital Areas A radiation and shielding design review was performed in accordance with NUREG-0737, Action item ll.B.2 (USNRC,1980), in order to ensure personnel accessibility after a design-basis accident (DBA). The DBA considered for this evaluation was the loss-of-coolant accident (LOCA). The projected dose to complete each activity necessary to mitigate a DBA LOCA, en route to and in vital areas, is less than the 5 rem design limit of NUREG-0737. At Millstone 3, this requirement is met by providing sufficient shielding of components containing post-accident l

radioactive inventories, consistent with anticipated access routes and stay times.

l One activity evaluated as part of the review was the post-LOCA hydrogen purge of containment. Because this activity requires multiple failures resulting in a loss of both trains of the Hydrogen Recombiner, this is considered beyond the design basis of a DBA LOCA. As such, the design guidance for performance of an activity within 5 rem is no longer applicable.

The projected dose for this activity is less than 25 rem, which falls within the authorized limit of 25 rem as described in the Emergency Plan for severe accident mitigation.

Areas requiring accessibility (vital areas) are those areas where post-LOCA actions can be taken over the short-term to ensure the capability of operators to control and mitigate the consequences of an accident. A description of the post-accident activities is summarized below and in Table 12.3-3.

1.

Locally trip the reactor trio breakers and bvoass breakers This action is performed at the 43' 6" elevation in the auxiliary building MCC rod control area. This is done in the event that the reactor failed to trip. This action must take place as soon as possible. Thus, the 0 to 30 minute time frame is assumed. While this step is done only in the event of an ATWS (beyond the design basis scenario), it is conservatively included as a required operator action.

2.

Samplina PASS (Post Accident Sampling System) samp!es are required to, actopic and gas composition analyses of containment air and isotopic, pH, boron and chloride analyses of reactor coolant / sump samples. NUREG-0737 II.B.3 states that the combined time for sampling and analysis should be 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months or less from the time a decision is made to take a sample. In a design basis LOCA where fuel tailure would occur almost immediately, the O to 30 minute time is assumed.

3.

Local actions needed to realian Spent Fuel Pool Coolina. RBCCW and Service Water for spent fuel pool coolina FSAR 9.1,3.3 states that spent fuel pool cooling will be initiated approximately 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after the LOCA. This requires operator action in the spent fuel pool building. The 2 to 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months time frame is assumed.

4.

Powerina the Plant Process Computer The Plant Process computer is normally not powered from an Emergency Bus. Itis powered from an uninterruptible power supply that may last for only 30 minutes. Thus, Page1of 7 1

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the O to 30 minute category is assumed. The plant process computer is used for SPDS and OFlS. In order to restore power to the plant process computer, MCC 32-3T is energized on the 38' level in the turbine building.

5.

Powerina the Si accumulator valves For post-LOCA cooldown and depressurization, the Si accumulator isolation valves are closed to prevent injection of nitrogen that might interrupt natural,:irculation. It is necessary to re-power the valves from the 24' 6" level in the auxiliary building. Since this would be done only after the plant is stabilized in preparation for a cooldown, the 30 minute to 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months time frame is assumed.

l 6.

Initiate hydroaen monitor FSAR Section 6.2.5.2 states that this system will be available to provide continuous monitoring within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and 30 minutes of an accident. For dose consequence evaluation, availability within 30 minutes was assumed for conservatism. Thus, the O to 30 minute category is assumed. Access to the hydrogen recombiner building is needed in order to initiate hydrogen monitoring.

7.

Initiate hydroaen recombiner FSAR Section 6.2.5.3 assumes that the hydrogen recombiner will be placed in service within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at which time the hydrogen concentration is 1.6% However, it also states that the hydrogen recombiner is started well before this concentration is attained.

To meet the latter requirement, emergency procedures specify initiation of the hydrogen recombiner when hydrogen concentration exceeds 0.2%. Thus, the 30 minute to 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months time frame is assumed. Access to the hydrogen recombiner building is needed in order to initiate the hydrogen recombiner.

8.

Initiate hydroaen ourne FSAR Section 6.2.5.3 states that if hydrogen purge is initiated at 4 days, the 50 scfm purge flow rate is sufficient to maintain hydrogen concentration in the containment below 4%. Thus, the 4 day to 30 day category is assumed. Access to the 4'6",24' 6" and 43'

)

6" levels of the auxiliary building are needed.

j 9.

Locally open the breakers for RWST/Charaina pump suction valve Valves 3CHS*LCV112D and 3CHS*LCV112E are closed and associated breakers are opened as a backup to the check valve to prevent back leakage to the RWST when post-LOCA recirculation is established.

As stated in FSAR Section 6.3.2.8, the minimum time frame for post-LOCA recirculation is 33 minutes. Thus, the 30 minute to 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months category is assumed. This action is performed on the 24' 6" level of the auxiliary i

building.

10.

Install and Operate air compressors for RSS sumo pumps Safety grade sump pumps were installed because of the potential for ground water intrusion flooding the RSS sump pumps. Installation of air compressors is necessary to Page 2.of 7

N'C operate the sump pumps. It is assumed that operation of the sump pumps would take place roughly 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months post accident. Thus, the 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months category is assumed.

The air compressors will be operated periodically, once every twelve hours. Thus, this also should be evaluated for the 1 to 4 day and the 4 day to 30 day category. The air compressors are installed outside of the ESF building.

11.

Open the breakers for the non-safety arade sumo pumps The operation of the non-safety grade sump pumps may mask the presence of a leak.

Thus, the need to secure sump pumps in ECCS pump cubicles and common areas in the Auxiliary and ESF buildings. The 1 day to 4 day time frame is assumed. This action requires access to the 21' elevation of the ESF building and the 24'6" elevation of the auxiliary building.

Stoppina non-saf t; arade HVAC fans to maintain the SLCRS boundarv.

12.

f Prior to placing ine Control Room on emergency ventilation, the operators will verify that the non-safety grade fans that could affect the SLCRS boundary are tripped. If not, local actions in the auxiliary building (43' 6" level) and service building are necessary to trip fans 3HW-FN1 A, 3HVV-FN1B, 3HVQ-FN2, 3HVR-FNS and 3HVR-FN7. This will be done after Control Building Isolation has been actuated for one hour. Thus, the 30 minutes to 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months category is assumed.

13.

Alian Service Water to AFW to provide lona term SG decay heat removal For LOCAs where the break is not big enough to remove decay heat, the design basis long term cooling method is service water as back up to auxiliary feedwater in conjunction with post-LOCA recirculation. As discussed in the basis for Technical Specification 3/4.7.1.3, the Demineralized Water Storage Tank supply will last a minimum of 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months . Thus, the 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months category is assumed. This action is performed on the 21' elevation of the ESF building.

14.

Reset MCC breakers for Diesel Generator keep warm systems This action is taken when offsite power is available and the running dies'el generator is stopped. The keep warm system assures that the diesel generator would be maintained in the optimum condition for a subsequent start if a loss of offsite power occurs later in the transient. This action is performed in the emergency diesel generator building.

In addition to the areas and activities defined above, the fo! lowing areas require post-accident access and continued occupancy:

1. Control Room - post accident control room habitability is discussed in Section 6.4 and control room post-LOCA doses are presented in Section 15.6.5.

2. Technical Support Center - post-LOCA dose evaluation is presented in Section 15.6.5.

I It has been determined that post-accident access to areas addressed in NUREG-0737, which have not been identified above, is not required for Millstone 3.

Page 3 of 7

% mo S-to3 Table 12.3-4 provides an estimate of the anticipated times after a LOCA that vital area access is required, with consideration given to the typical 30 minute minimum time frame assumed for operator action outside control room and the X/O intervals assumed in the FSAR Chapter 15 accident analysis. Outside travel routes are shown on Figure 12.3-10 and are listed on Table 12.3-3. Figure 12.3-10 also provides details conceming the location of specific areas where radiation levels are significantly greater than in adjacent areas. A general description of the ingress travel routes, primary and alternates, are described below (the egress path is the same as the ingress path except for alternate routes to the backup Chemistry Laboratory in travel-route 4).-

Travel route 1:The primary route is from the control building through the service building to the auxiliary building (no outdoor travel). The alternate route is from the control building to the service building to the exit between the service and auxiliary building to the north entrance to the auxiliary building.

Travel route 2: The primary route is from the control building through the service building to the exit between the service building and the auxiliary building, along the north side of the waste disposal building then south to the ESF trailer to the mobile air compressors stored outside of the ESF Building, inside the Radioactive Materials Area fence). The alternate route is from the control building to the service building to the turbine building to the RR loading area, east along the roadway past the RWST, through the ESF. trailer to the outside of the ESF building.

Travel route 3:The primary route is from the control building through the service building corridor leading to the roadway beside the MSV building to the RCA gate south of the hydrogen recombiner building (HRB) and into the HRB. The alternate route is from the control building to the service building to the turbine building to the RR loading area to the RCA gate adjacent to the HRB.

Travel route 4:The primary route is from the control building through the service building corridor leading to the roadway beside the MSV building to the RCA gate south of the hydrogen recombiner building (HRB) and into the HRB. The alternate route is from the control building to the service building to the turbine building to the RR loading area to the RCA gate adjacent to the HRB. The sample analysis is performed in the MP3 chemistry lab which is on the egress path. Figure 12.3-10, sh. 4, provides two additional routes for sample analysis in the MP1/ MP2 service building.

Travel route 5: The primary route is from the control building through the service building to the exit between the service building and the auxiliary building,, following the roadway north and east of the waste disposal building, then entering the fuel building. The alternate route is from the control building through the service building corridor to the turbine building to the RR loading area, then east along the roadway past the RWST, north to the fuel budding.

Travel route 6:The primary route is from the control building to turbine building 38' level. No alternate is given since doses would only increase.

Page 4 of 7

4mp 3-u 3 s

e Travel route 7: The primary route is from the control building to the emergency diesel generator building. No alternate is given since doses would only increase with any other route.

Travel route 8:The primary route is from the control building through the service building to the exit between the service and auxiliary buildings, along the north side of the waste disposal building, south to the ESF trailer to the ESF building or the. turbine driven Aux. Feedwater Pump Room. The alternate route is from the control building through the service building corridor to the turbine building to the RR loading area, then east along the roadway past the RWST, through the ESF trailer and into the ESF building or the turbine driven Aux. Feedwater Pump Room.

Travel route 9:The primary path is from the control building through the service building corridor to the turbine building auxiliary bay, lower level then across the road to the auxiliary buidling. The alternate path is from the control building to the service building, past the Chemistry Laboratory, exit the service building to the auxiliary building.

The following general assumptions and criteria are used as a basis for review of all vital areas and access routes as applicable:

1. The starting point for all activities is the Unit 3 Control Buiiding.

2. In order for an access / egress pathway to be considered; acceptable, the total dose for activities required for mitigation of the design basis accident (which includes the dose to perform the activity and the associated transit dose) must be no greater than the 10 CFR Part 50 Appendix A GDC-19 dose criteria. The determination of total dose is based on the earliest time post-LOCA when access to the designated vital area is required as identified in Table 12.3-4.

3. All calculated outside pathway doses are assumed to be comprised of contributions from (a) containment radiation (both direct shine and skyshine contributions) and (b) direct radiation from the overhead plume.

Gaseous and liquid LOCA source terms used in the review are not less than that stated in NUREG-0737, Section ll.B.2, which provides the minimum source terms to be used for evaluation of the adequacy of radiation protection to the operators.

To determine post-accident doses to personnel for performance of and transit to the previously identified activities, the following sources of radiation were considered in the post-accident analysis.

1. Auxiliary Building Radiation from containment atmosphere shining through electrical penetrations.

. Radiciodine buildup in the SLCRS filter.

Sump water in the safety injection system piping located below the elevation 24' 6" floor.

Page 5 of 7 I

l S8 me?> - W5 f

i Containment atmosphere shine through the personnel hatch and surrounding walls and l

floors Sump water in safety injection and charging system piping and associated shine through walls and floors

2. Hydrogen Recombiner Building l

Radiation from con!9inment simosphere contained within the active hydrogen recombiner reaction chamber and blower.

Radiation from containment atmosphere and condensate contained in piping carrying containment atmosphere to the active hydrogen recornbiner building.

Direct radiation shine from the recombiner(s) through the concrete walls into the recombiner control room and sample room.

Radiation scattered out of the active recombiner cubicle entranceway to the recombiner control area.

Radiation from containment atmosphere contained and/or RCS/ SUMP contained within the PASS sample lines and sample chamber (sample cubicle area).

Radiation scattered front the PASS sample lines (containing sample) in the recombiner and sample rooms to the sample labyrinth and recombiner control room.

Radiation from PASS samples.

3. Fuel Building Direct shine from containment Plume shine Shine from the RHR heat exchanger in the ESF building.

Shine from the fuel pool cooling pumps.

4.

ESF Building Shine from RSS and SlH piping Shine through the wall from the Recirculation Coolers Shine from RWST piping Shine from Auxiliary Steam piping

5. Along routes from control building to the vital areas.

Skyshine from containment.

Page 6 of 7

hp3 to3 Direct shine from containment.

Plume shine.

Systems containing sources of radiation which are identified in NUREG-0737 but which have not been identified in buildings discussed above are considered to be either irrelevant following an accident or of negligible contributors to personnel exposure following an accident. For example, the GWS system is a negligible contributor of radiation following an accident because when an accident occurs, the only use of this system is for post-1.OCA hydrogen purge. Use of this system occui3 no sooner than 4 days post-LOCA. Upon completion of the initiation of hydrogen purge, the only other activity required for mitigation is operation of air compressors for RSS sump pumps (Activity 10 in Table 12.3-3). That activity is performed outside of the ESF Building and there is no impact from GWS piping in that area.

The results of the dose calculations indicate that the plant shielding and design provide adequate protection to operators following a design basis LOCA to ensure compliance with the NUREG-0737 design dose requirements.

1 l

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Page Z<of 7

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Docket No. 50-423 B17308 l 8 Millstone Nuclear Power Station, Unit No. 3 Proposed License Amendment Request Post Accident Access to Vital Areas

-(PLAR 3-98-6)

FSAR Table 12.3-3 Rewrite June 1998

otB mr3 t6%

W4 PS-3 PS AQ-

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Ch' p

Table 12.3-3 Operator Activity Locations and Time Durations Activity Location Approx.

Travel

{

Duration Route *

(minutes) 1 Locally trip the reactor trip breakers and 43' 6" Aux. Bldg.

<5 9

bypass breakers (MCC Rod Control) 2 PASS Samples Hydrogen Sample Collection Recombiner Building

<100 (HRB) 4 Sample Analysis Laboratory

<130 3 Local actions needed to realign Spent Fuel Pool Cooling, RBCCW and Service Water Spent Fuel Bldg.

<15 5

for spent fuel pool cooling 4 Powering the Plant Process Computer 38' Turbine Bldg.

<10 6

5 Powering the Si accumulator valves 24' 6" Aux. Bldg.

<5 9

6 Initiate hydrogen monitor HRB

<30 3

7 initiate hydrogen recombiner HRB

<60 3

8 initiate hydrogen purge 4' 6" Aux. Bldg.

<60 1

24' 6" Aux. Bldg.

43' 6" Aux. Bldg.

9 Locally open the breakers for RWST/CHG 24' 6" Aux. Bldg.

<5 9

pump suction valves 10 Install and operate air compressors for RSS sump pumps Outside of ESF Bldg.

a. install air compressors

<120 2

b. operate air compressors

<20 11 Open the breakers for the non-safety grade 21' ESF Bldg.

<15 2

sump pumps in the ESF and Auxiliary 24' 6" Aux. Bldg.

<15 1

buildings 12 Stopping non-safety grade HVAC fans to 43' 6" Aux. Bldg.

<5 9

maintain the SLCRS boundary 4'6" Service Bldg.

<5 13 Align Service Water to AFW to provide long 21' ESF Bldg.

<60 8

term SG decay heat removal 14 Reset MCC breakers for Diesel Generator Emergency Diesel N/A**

7 keep warm systems Generator Building

Figure 12.3-10 graphically depicts each route by route number

" There are no appreciable dose rates in the Emergency Diesel Generator Building

l 3;

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Docket No. 50-423.

B17.308 l

l l

l l C Millstone Nuclear Power Station, Unit No. 3 -

l Proposed License Amendment Request Post Accident Access to Vital Areas (PLAR 3-98-6)

FSAR Table 12.3-4 Rewrite June 1998 e

qrd3 -103 Table 12.3-4 Activity initiation Time Time Frame

Activity 0 to 30 rninutes 1 Locally trip the reactor trip breakers and bypass breakers 2 Sampling 4 Powering the process computer 6 Initiate hydrogen monitor 14 Reset MCC breakers for Diesel Generator keep warm systems 30 minutes to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 5 Powering the Si accumulator valves 7 Initiate hydrogen recombiner 9 Locally open the breakers for RWST/CHG pump suction valves 12 Stopping non-safety grade HVAC fans to maintain the SLCRS boundary 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 3 Local actions needed to realign Spent Fuel Pool Cooling, RBCCW and Service Water for spent fuel pool cooling 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 10 install and operate air compressors for RSS sump pumps 13 Align Service Water to AFW to provide long term SG decay heat removal 1 day to 4 days 10 Operate air compressors for RSS sump pumps 11 Open the breakers for the non-safety grade sump pumps in the ESF and Auxiliary Building 4 days to 30 days 8 Initiate hydrogen purge 10 Operate air compressors for RSS sump pumps

the starting time of the time frame listed is used for source term decay correction

Docket No. 50-423 B17308 D Millstone Nuclear Power Station, Unit No. 3 Proposed License Amendment Request Post Accident Access to Vital Areas (PLAR 3-98-6)

FSAR Figure 12.3-10 Revisions June 1998

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.g. Docket No. 50-423 B17308 Millstone Nuclear Power Station, Unit No. 3 Proposed License Amendment Request Post Accident Access to Vital Areas (PLAR 3-98-6) Description of the Change, Background and Safety Summary i. t 1 June 1998 (.. (..- g

E l U.S. Nuclear Regulitory Commission B17308\\ Attachment 3\\Page 1 l Backaround The changes result from NRC Inspection Report 50-423/97-208 for NUREG 0737 Item II.B.2, Plant Shielding. It was determined that the FSAR description of post-accident vital area routing was out of date because the RCA [ boundary fence created an access problem on the designated routes to the Hydrogen Recombiner and Fue! Building. This FSAR change revises the routes l to accommodate the fence location and allow for the time to unlock gates. As a result of this change, new source terms were identified and added to the dose calculation to account for DBA LOCA plume shine and RWST back leakage dose. A time and motion study was used to refine the calculation of post accident sample doses to the emergency workers. Subsequently, an engineering evaluation identified additional activities or revisions to activities previously evaluated. The dose consequences of all activities identified were re-evaluated in a new dose calculation. The activities are listed below. The current FSAR descriptions are out of date and as such do not include all required post-accident actions. This FSAR change adds actions recognized in Plant Operating Procedures to those currently listed in the FSAR. Activities not previously evaluated with respect to doses are denoted as "New". 1. Locally trip the reactor trip breakers and bypass breakers 2. Sampling 3. Local actions needed to realign Spent Fuel Pool Cooling, RBCCW and Service Water for spent fuel pool cooling l 4. Powering the Plant Process Computer (new activity) j 5. Powering the Si accumulator valves (new activity) i 6. Initiate hydrogen monitor { i 7. Initiate hydrogen recombiner 8. Initiate hydrogen purge (new activity) 9. Locally open the breakers for RWST/ Charging pump suction valve (new activity) 10. Install and Operate air compressors for RSS sump pumps (new activity) 11. Open the breakers for the non-safety grade sump pumps (new activity) 12. Stopping non-safety grade HVAC fans to maintain the SLCRS boundary. (new activity) 13. Align Service Water to AFW to provide long term SG decay heat removal (new activity) 14. Reset MCC breakers for Diesel Generator keep warm systems (new activity) One route was removed from the FSAR since it was determined that the actions cited were not required to mitigate a DBA LOCA. The route removed was:

U.S. Nuclear Regulatory Commission B17308\\ Attachment 3\\Page 2 ' A - Locally dump ste' am For. post LOCA cooldown and depressurization, the operator is directed to locally dump steam from the steam generators if it cannot be done from the control room. For the DBA LOCA, cooldown and depressurization for RHR entry is not needed. The long-term cooling requirement is met by post-LOCA recirculation. Thus, this action is not a required operator action to mitigate a design basis LOCA. It should be noted that a new activity (item 8 above), " Initiate hydrogen purge", is not a design basis activity. This activity requires use of the Containment Vacuum system which is identified in the FSAR and SER as a backup to the Hydrogen Recombiner. (The Containrnent Vacuum system is independent of the Containment Purge system which is credited for fuel handling accidents.) This system is not safety related and is non-seismic except for the containment isolation valves and associated piping. Therefore, use of it for post-LOCA .l activities is conceded beyond the design basis of the plant. As such, the design i dose limit criterion of GDC 19 (5 rem) will not be applied. Instead, the Emergency Plan provides a dose limit of 25 rem that is applicable for severe accident mitigation. This dose limit will be used as the design limit. Description of the Chanae Unit 3 FSAR Section 12.3.1.3.2, Figure 12.3-10, Tables 12.3-3 and 12.3-4 are being extensively rewritten to update the FSAR with revised post accident mitigation activities, vital area access travel routes, and times. . SAFETY

SUMMARY

The proposed changes to the FSAR (descriptions of post-LOCA vital area access travel routes, activities, exposure times and compliance with dose criteria) constitute an Unreviewed Safety Question because:

1. There are newly identified (for dose) activities which add to the dose consequences

2. The revised activities differ significantly from the current FSAR and can not be compared to the consequences.

These changes (except as noted below) are safe because they meet the design dose criterion of.NUREG 0737 as applied to a DBA LOCA. In the case of-initiation of hydrogen purge, which is considered as a "beyond design basis" activity, the projected dose is less than the.25 rem dose limit specified in the Emergency Plan for severe accident mitigation and is thus considered safe. .n

r" U.S. Nucibar' Regulatory Commission [. . B17308\\ Attachment 3\\Page 3 Evaluated post-LOCA activities, that are within the design bases, are within the dose criterion.of GDC 19 as specified in NUREG 0737. These activities and doses differ from what was originally presented in the FSAR and SER. As a result of these' differences, a direct comparison with the dose consequences listed in Table 12.3-4 of the FSAR can not be made. The FSAR is very specific- .about locations that were evaluated per NUREG 0737. Additional areas and -activities were added and as a result there. are new dose consequences associated with the activities in those locations. In review of existing locations, new source terms were identified, such as plume shine which result in changing the dose consequences. As a result of the changes, it:can not be readily determined whether there is an increase in consequences. Due to this uncertainty and the inclusion of newly identified (for dose) activities which add to the dose consequences, this FSAR change is being treated as an Unreviewed Safety Question. Initiation of hydrogen purge is an evaluated activity that is beyond the design basis of the plant. This activity is beyond the design basis because-the ' Containment Air Vacuum system is not a safety related system (except for the . containment isolation valves) but is credited in NNECO's engineering evaluation for a hydrogen purge of containment, starting 4 days post-LOCA. This system acts as a backup to the Hydrogen Recombiners when both trains of the, recombiner are inoperable (Single failure is not credited-here. SRP 6.2.5, Combustible Gas Control in Containment discusses that following a LOCA "there be suitable redundancy in components and features,.., assuming a sing,'e failure."). Being beyond the design basis, the Emergency Plan authorizes one-time only exposures of up to 25 rem for severe accident mitigation actions. Based upon the 25 rem limit, this post-LOCA activity.:an be performed safely. Because of the increase in consequences, this activity is considered an Unreviewed Safety Question.

.,-.g. Docket No. 50-423 B17308 Millstone Nuclear Power Station, Unit No. 3 Proposed License Amendment Request Post Accident Access to Vital Areas (PLAR 3-98-6) Significant Hazards Consideration and Environmental Considerations ~ i June 1998 l ?,

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U.S. Nuclear Regulatory Commission 1

B17308\\ Attachment 4\\Page 1 - Significant Hazards Consideration NNECO has reviewed the proposed revision in accordance with 10CFR50.92 and has concluded that the revision does not involve a significant hazards consideration (SHC). The basis for this conclusion is that the three criteria of 10CFR50.92(c) are not satisfied. The proposed revision does not involve an SHC because the revision would not: 1. Involve a significant increase in the probability or consequence of an accident previously evaluated. Final Safety Analysis Report (FSAR) Section 12.3.1.3.2, Post-accident access to vital areas, and its associated Figures and Tables are being updated..The current FSAR descriptions are out of date and as such do not include all required post-accident actions. Therefore, this FSAR change adds actions to those listed in the FSAR as well as incorporating the recalculation of the doses associated with the required post-accident actions. The dose calculations utilize the appropriate post-accident source terms, area access requirements and stay times, including the appropriate routes to the areas. The calculations show that for all design basis required post-accident actions the calculated dose to the Operators / Emergency workers performing those actions remains below the 5 rem criterion of General Design Criteria (GDC) 19. The revision to the FSAR provides the required post-accident required operator actions. Changing the FSAR to include the current post-accident vital access requirements and associated information for the supporting dose calculations can not cause an accident. In addition, the calculated dose to the Operators / Emergency workers for all design basis required actions is below the GDC 19 limit of 5 rem. Therefore, the proposed revision does not involve a significant increase in the probability or consequence of an accident previously evaluated. 2. Create the possibility of a new or different kind of accident from any accident previously evaluated. The change is to the calculated post-accident vital access dose analyses and . the FSAR description of that analyses. No new procedural Operator / Emergency. worker actions are associated with the change. However, since the information in the FSAR was outdated, there are Operator / Emergency actions being added to the FSAR. Dose calculations associated with those actions have been performed ~ utilizing the appropriate assumptions with respect to source terms, vital area access travel routes and. stay times, and times when the post-accident actions would be performed. The analyses confirmed that the calculated doses associated with all required post-accident actions are less than the 5 rem limit of GDC 19. _ There are no changes to the Emergency Operating Procedures [l associated with this change.

,I, e"* U.S. Nucl:ar Regulatory Commission B17308\\ Attachment 4\\Page 2 Therefore, the proposed revision does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Involve a significant reduction in a margin of safety. I The dose calculations confirm that the calculated dose associated with all design basis post-accident Operator / Emergency worker actions is below the limit of 5 rem of GDC 19. There is one action, initiation of hydrogen purge, for which the calculated dose to the Operator / Emergency worker exceeds 5 rom. This action is a backup means of limiting the hydrogen concentration inside containment post-accident. This action would only be performed for multiple failures which would disable both trains of the safety-grade hydrogen recombiner system. As such this action is not a required design basis action and does not need to meet the 5 rem limit. The calculated dose for this action is below the 25 rem limit that is specified in the Station Emergency Plan for severe accident mitigation actions. Therefore, the proposed revision does not involve a significant reduction in a margin of safety. In conclusion, based on the information provided, it is determined that the proposed revision does not involve an SHC. Environmental Considerations NNECO has reviewed the proposed license amendment against the criteria of 10CFR51.22 for environmental considerations. The proposed revision does not involve an SHC, does not significantly increase the type and amounts of effluents that may be released offsite, nor significantly increase individual or cumulative occupational radiation exposures. Based on the foregoing, NNECO concludes that the proposed revision meets the criteria delineated in 10CFR51.22(c)(9) for categorical exclusion from the requirements for environmental review.}}

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