ML20198E794

From kanterella
Jump to navigation Jump to search
Proposed Tech Specs Incorporating Addl Sys Leakage Limits & Leak Test Requirements for Systems Outside Containment Which Were Not Previously Contained in TS 4.5.4 Nor Considered in TMI-1 UFSAR DBA Analysis Dose Calculations for 2568 Mwt
ML20198E794
Person / Time
Site: Three Mile Island Constellation icon.png
Issue date: 07/30/1997
From:
GENERAL PUBLIC UTILITIES CORP.
To:
Shared Package
ML20198E790 List:
References
NUDOCS 9708080204
Download: ML20198E794 (46)
v • d • e


Text

_____ ____ _ ___ _ _

TABLE OF CONTENTS

~ SKlio.0 P3gs

-3.16 SHOCK SUPPREdSORS (SNUBBERS) 3-6'

, 3.17- REACTOR BUILDING AIR TEMPERATURE 3-80 3.18 FIRE PROTECTION (DEthTED) 3 86 3.19 -CONTAINMENT SYSTEMS 3 95 3.20 SPECIAL TEST EXCEP]lONS (DELETED) 3-95a 3.21 PADIOACTIVE EFFLUENT INSTRUMENTATION (DELEICQ) 3 96 3.21 1 RADIOACTIVE LIQUID EFFLUENT INSTRUMENTATION (DELETED) 3 96 3.21.2 RADIOACTIVE GASEOUS PROCESS AND EFFLUENT 3-96 MONITORING INSTRUMENTATION (DELETED) 3.22- RADIOACTIVE EFFLUENTS (DELETED)._ 3-96 3.22.1 LIQUID EFFLUENTS (DELETED) 3 96 3.22.2 GASEOUS EFFLUENTS (DELETED) 3-96 3.22.3 SOLID RADIOACTIVE WASTE (DELETED) 3 96 3.22.4 TOTAL DOSE (DELETED) 3 96 3.23 RADIOL-OGICAL ENVIRONMENTAL MONITORING (DELETED) 3 96 3.23.1 MONITORING PROGRAM (DELETED) 3%

- 3.23.2 LAND USE CENSUS (DELETED) 3-96 3.23.3 INTERLABORATORY COMPARISON PROGRAM (DELLTED) 3-%

3.24 REACTOR VESSEL WATER LEVEL 3-128 4 SURVEILLANCE STANDARDS - 4-1 4.1 DPERATIONAL SAFETY RF,yJH 4-1 4.2 REACTOR COOLANT SYSTEM INSERVICE INSPECTIOE 4 11 4.3 TESTING FOLLOWING OPENING OF SYSTEM (DELETED) 4-13 4.4 REACTOR BUILDING 4-29 4.4,1 CONTAINMENT LEAKAGE TESTS 4-29 4.4.2 STRUCTURAL INTEGRITY 4-35 4.4.3 DELETED 4-37 4.4.4 HYDROGEN RECOMBINER SYSTEM 4-38 4.5 EMERGENCY LOADING SEOUENCE AND POWEP. TRANSFER. 4-39 EMERGENCY CORE COOLING SYSTEM AND REACTO3 BUILDING COOLING SYSTEM PERIODIC TESTING 4.5.1 EMERGENCY LOADING SEQUENCE , 4-39 4.5.2 EMERGENCY CORE COOLING SYSTEM 4-41

-4.5.3 REACTOR BUILDING COOLING AND ISOLATION SYSTEM 4 43 4.5.4 ACCIDENT RECIRCULATION SYSTEMS LEAKAGE 4 4.6 EMERGENCY POWER SYSTEM PERIODIC TESTS 4-46 4.7 - REACTOR CONTROL ROD SYSTEM TESTS 4-48 4.7.1 CONTROL ROD DRIVE SYSTEM FUNCTIONAL TESTS 4-48 l'

'4.7.2 'CO!? TROL ROD PROGRAM VERIFICATION 4-50

-iii-

' Amendment 73r81r488cl29r137,116,147rM8r19Ir19h498 9700000204 970730 PDR ADOCK 05000289 P PDR m

]

j t -

- TABLE OF CONTENTE 1

SKliOD lho-3.16  !

S[]OCK SUPPRESSORS (SNUBBERS) 3-63

3.17 ~ BEACTOR BUILDING AIRTEMPERATURE 3 80 11 i

l 3.18 ElRE PROTECTION (DELETED) 3-86 1-3.19 CONTAINMENT SYSTEMS: 3-95 3.20 SPECIAL TEST EXCEPTIONS (DELETED) 3-95a 3.21 RADIOACTIVE EFFl.UENT INSTRUMENTATION (DELETTED) 3-96 I 3.21.1 RADIOACTIVE LIQUID EFFLUENT INSTRUMENTATION (DELETED) 3 96 3.21.2 RADIOACTIVE GASEOUS PROCESS AND EFFLUENT 3-96 4 MONITORING INSTRUMENTATION (DELETED) 3.22 -  : RADIOACTIVE EFFLUENTS (DELETED)~ 3 . 3.22.1 LIQUID EFFLUENTS (DELETED) 3-96

3.22.2 - GASEOUS EFFLUENTS (DELETED) 3 96 o 3.22.3 3 96 SOLID RADIOACTIVE WASTE (DELETED)

! 3.22.4 ' TOTAL DOSE (DELETED) 3-96 1- 3.23 RADIOLOGICAL ENVIRQ@1 ENTAL MONITORING (DELETED) 3%

3.23.1 MONITORING PROGRAM (DELETED) 3 96

! 3.23.2 LAND USE CENSUS (DELETED) 3 96

3.23.3 . INTERLABORATORY COMPARISON PROGRAM (DELETED) 96

[

3.24 REACTOR VESSEL WATER LEVEL 3 128

-4 SURVEILLANCE STANDARDS 41

- 4.1 OPERATIONAL SAFETY REVIEW -

4-1 4.2 REACTOR COOLANT SYSTEM INSERVICE INSPECTION 4 11 4.3 TESTING FOLLOWINQOPENING OF SYSTEM (DELETED) 4 13 ,

[ 4.4 BEACTOR BUILDING 4-29 4.4.1 CONTAINMENT LE AKAGE TESTS 4-29 4 A.2 STRUCTURAL INTE'iRITY -4 35

4.4,3 DELETED 4 ' '

4.4.4 HYDROGEN RECOMBINER SYSTEM 4 38 4.5 EMERGENCY LOADING SEOUENC.S AND POWER TR ANSF18, - 4-39 EMERGENCY CORE COOLING SYSTEM AND REACTOR -

I

! BUILDING COOLING SYSTEM PERIODIC TESTING I 4.5.1 EMERGENCY LOADING SEQUENCE 4-39 l 4.5.2 EMERGENCY CORE COOLING SYSTEM 4-41 4.5.3 REACTOR BUILDING COOLING AND ISOLATION SYSTEM 4 43-4.5.4 ACCIDENT RECIRCULATION SYSTEMS LEAKAGE 4 45 3 _ 4.6 EMERGENCY POWER SYSTEh1 PERIODIC TESTS 4 46 l 4.7 - REACTOR CONTROL ROD SYSTEM TESTS 4-48 1 4.7.1 - CONTROL ROD DRIVE SYSTEM FUNCTIONAL TESTS 4-48

4.7.2, CONTROL ROD PROGRAM VERIFICATION - 4-50 4

-iii-a

- Amendment 73r84r408rl39rl47r146rl47rl58r191rl#7r198 I .

-- . . -- . - - - .- = -.

Ibic The Auxiliary and Fuelllandling Building Air Treatment System (part of the Auxiliary and Fuel liandling Building Ventilation System) consists of four banks of exhaust filters (All-F2A, B, C, and D) and two sets of fans (All-E-14A and C, and All-E14B and D) which take the exhaust air from both the Auxiliary Building and the Fuel liandling Building and discharge it to the Auxiliary and Fuel 11andling Building exhaust stack (References 1 and 2). The air normally passes through all four filter banks (roughing filters, iiEPA filters, and charcoal adsorbers) when either set of fans is in operation. l This system is a Normal Ventilation Exhaust Air Filtration And Adsorption System. Although not Nuclear Safety Related,ifit is available,it can be used to reduce the off site dose for the Waste Gas Decay Tank Rupture (WGTR) - Reference 4, Maximum flypothetical Accident -

(MilA) - Reference 3, or for other events releasing radioactivity through the Auxiliary Building. The dote from these events will be less than the 10 CFR 100 limits with or without system operatian.

l The in-place testing criteria for the filter banks, and laboratory testing for the charcoal l adsorbers meet the guidelines given in Regulatory Guide 1.52, Rev. 2 - 1978, in accordance with ANSI /ASME N510-1980.

Note: The Fuel llandling Building ESF Air Treatment system controls the release resulting from a postulated spent fuel accident in the Fuel liandling Building per Technical Specification 3.15.4.

Rsferences (1) UFSAR Section 9.8.2 " Fuelllandling Building Ventilation System" (2) UFSAR Section 9.8.3 " Auxiliary Building Ventilation System" (3) UFSAR Section 14.2.2.5 " Maximum liypothetical Accident" (4)- UFSAR Section 14.2.2.6 " Waste Gas Tank Rupture" 3-62d Amendment No.45;-122,157,

_ l

l 4.5.4 ACCIDENT RECIRCULATION SYSTEMS LEAKAGE i l

APPlicAhility Applies to those portions of the Decay lleat, Huilding Spray, and Make-Up Systenis which are required to contain post accident sump recirculation fluid.

Ohitclire To maintain a low leakage rate from the accident recirculation systems to prevent significant off-site exposures.

. Speci0catinD 4.5.4,1 The total maximum allowable leakage from the applicable portions of the Decay

' Heat, Huilding Spray and Make-Up System components as measured during refueling tests in Specification 4.5.4.2 shall not exceec 0.3 gpm.,

a 4.5.4.2 During each refueling interval the following tests of the applicable portions of the Decay Heat Removal, Huilding Spray and Make-Up Systems shall be conducted to determine leakage:

l- a. The applicable portion of the Decay lleat Removal System that is outside containment shall be leak tested with the Decay lleat pump operating or i by hydrostatically testing at no less than 350 psig, except as specified in "b"

b. Piping from the Reactor Building Sump to the Building Spray pump and i Decay Heat Removal System pump suction isolation valves shall be pressure tested at no less than 55 psig.

1

c. The applicable portion of the Huilding Spray system that is outside containment shall be leak tested with the Building Spray pumps operating and HS-V-I A/B closed or by hydrostatic testing at no less than 350 psig, except as specified in "b",

i

d. The applicable portion of the Make-Up system on the suction side of the Mr.ke-Up pumps shall be leak tested with a Decay llent pump operating and Dif-V-7A/H open or hydrostatic testing at no less than 200 psig.

, c. The applicable portion of the Make-Up system Imm the Make-Up pumps to the containment boundary shall be leak tested with a Make-Up pump operating or by hydrostatic test at no less than 3050 psig.

f. Visual inspection shall be made for leakage from components of these systems. Leakage shall be measured by collection and weighing or by another equivalent method.

Bucs The leakage rate limit for the accident recirculation portions of the Decay Heat Removal,

. Building Spray, and Make-Up Systems is based on ensuring that potential leakage after a loss-of-coolant accident will not result in oli-site dose consequences in excess of those calculated to comply with the 10 CFR 100 limits (Reference I and 2).

1

. Effetcacss (1) UFSAR, Section 6.4.4 " Design Basis Leakage" andTable 5.4-3 " Leakage Quantities to the Auxiliary Building" (2) UFSAR, Section 14.2.2.5(d) " Effects of Engineered Safeguards Leakage During Maximum Hypothetical Accident"

! Amendment No. 467, 4

1 ENCLOSURE 2 Affected TMI.1 Updated Final Safety Analysis Repon Pages l

l

Thil 1/FSAR 2.5.4 ACCIDENT METEOROLOGY Values for atmospheric diffusion have been determined to provide a basis for assessing possible radiation exposure during the course of the accidents hypothesized in Chapter 14. Table 2.5-15 shows these values, for time periods of 0 to 2,2 to 8,8 to 24,24 to 96, and 96 to 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> at the exclusion distance (2000 ft) and the low population distance (2 miles),

initial values of accident X/Q (Table 2.5-15) were chosen so that, based on a computer search l of 2 years of site weather records, there was less than a 5 percent chance they would be exceeded following an accident if that accident happened at any random time. This was done in the following way. First, both a plume centerline value and a sector average value of X/Q were calculated for each hour of the year for the selectt;d distances. The plume centerline values were

estimated using the following relationships which were found as shown in Reference 6.

i X/Q= 1 l

0 [n or e, + c A]

where:

0 =

Average wind speed (m/sec) oy = Horizontal difrusion coeflicient (m) o, = Vertical diffusion coeflicient (m) c= Building wake coeflicient (=l/2)

A= Approximate cross-sectional area of containment (>2000 m2)

The sector average X/Q values were determined using the general equation described in Subsection 2,5.3. Centerline X/Q values were used for time periods of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or less and sector average values for time periods greater than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

The estimated X/Q values for each hour obtained as described above were then processed sequentially to find, starting with each hour of the year, the average X/Q for the selected time period following that hour for each direction sector. This provided the information needed for estimating what chance there is (at a chosen distance) for a given X/Q to be exceeded when averaged over a selected time period, starting at any random time.

Later evaluations of short term (0-2 Hrs) accident meteorology (Reference 29) using a two year period of onsite meteorological data with better than 90% data recovery established an accident X/Q of 6.8 x 10" sec/cu. m. This value is used in Chapter 14 accident analysis with the exception of Fuel Cask Drop Accident. See 14.2.2.8, Table 14.2-26 and applicable reference.

2.5-5 UPDATE- '

/98

TMI 1/FSAR The values given in Table 2.5-15 were taken from Figures 2.5 4,2.5-5, and Reference 29 Figures 2.5-4 and 2.5-5 are plots of the probability of average X/Q values for different time periods following an accident generated by the WINDOW computer program using the site weather data.

l 2 5-5 A UPDATE-

/98

TMi 1/FS AR 2.5.5 APPL 1 CATION OF SF 6 TRACER GAS TESTS AT THREE hilLE ISLAND TO SITE METEOROLOGICAL RECORDS 2.5.5.1 Introduction During the summer and fall of 1971, atmospheric diffusion experiments using SF6 as a tracer gas were conducted on the Th11 site. The results of these tests were documented in a report titled,

" Atmospheric Diffusion Experiments with SF6 Tracer Gas at Three hiile Island Nuclear Station under Low Wind Speed Inversion Conditiorn," which was filed as Amendment 24 to Docket No.

50-289. NRC Safety Evaluation Report dated 7-11-73 concluded that the tracer tests did not provide sumcient justification for use in the accident diffusion model. Sections 2.5.5 and 2.5.6 are retained for historical purposes, refer to Section 2.5.4 for the applicable accident meteorology.

l The purpose of these tests was to develop a suitable model for use with the site weather data for predicting diffusion conditions during low wind speed nighttime conditions. The following are the results of computer studies which were made using site data and the results of the SF6 test program.

2.5.5.2 hiethods In the SF6 tracer gas test report, it was concluded that difrusion conditio'.s during low wind speed nighttime conditions are best described by models which account for p'ume meander.

To demonstrate the effect of the use of hiodel SW, cumulative probability distributions of hourly '

values of X/0 were prepared using the 2 year period of site weather records from hiay 1967 through hiay 1969 as reported in the FSAR. hiodel SW was used (was fixed at a value corresponding to the Pasquill F ditTusia condition) for nighttime hours which had wind speeds less than 3.5 mph. The X/Q values for all other weather conditions were computed using the range categories to determine Pasquill diffusion groups as in the FSAR. For this study, the speed data are taken at the 100 fl level which is approximately the same level used for correlating results of Phases 2 and 3 of the SF6 tests and corresponds to the mid height of the Reactor Building.

Calm conditions were assumed to have the measured wind direction range value and a wind speed of 0.5 m/sec. The relationships used in this study are as follows:

For all daytime hours and nighttime hours with wind speeds greater than 3.5 mph:

X/Q FSAR = 1 Range 0(nc yo,+cA) hiodel 2.5-7 UPDATE-

/98 1

J

TMI 1/FSAR Referring to Figure 2.5 7, it is seen that the highest average X/Q on land not owned by 4

Metropolitan Edison occurs approximately ESE of the site with a value of 7.5 x 10 sec/m'. The peak value at the river's edge is about 5.0 x 10sec/m' in the WNW direction. Ilowever, it is not expected that any individual would be in the river at this location for significant periods of time.

Ifit is conservatively assumed that an indivijual is fishing at this point in the river 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> per day for 2 months a year, he would be present E s 00 or 480 hours0.00556 days <br />0.133 hours <br />7.936508e-4 weeks <br />1.8264e-4 months <br /> a year (5.5 percent of the time).

Thus, taking an occupancy of 0.055 into r.ccount, the effective X/Q to be used in estimatirs his 4

exposure would be 0.055 x 5.0 x 10 or 2.75 x 10 sec/m'. This is lower than the limiting value 4

of 7.5 x 10 sec/m' at the site boundary in the ESE direction. An occupancy factor of 0.15 would have to be assumed before this point in the river would be limiting. It is highly unlikely that this degree of occupancy would be exceeded on an annual basis, considering the condition of the river l in the winter and the fact that fishing is generally considered to be better downstream than in the impoundment (formed by the York llaven Dam) near the site.

l 2.5.6 ADDITIONAL INFORM ATION RELATED TO TM1 SITE METEOPOLOGY AND DIFFUSION TESTS 2.5.6.1 IntmductiDB Following discussions with the AEC Regulatory Staff and their consultants on July 25.1972, information was submitted to further suppon the applicant's contention that atmospheric difTusion conditions at the Three Mile Island sitejustifv a choice of diffusion conditions for accident analysis equivalent to or better than those specified in AEC Safety Guide No. 4. Since the AEC has judged site difTusion characteristics on vertical temperature difference measurements, a 1 yr period of these data is included in this Subsection, as discussed in item a of Subsection 2.5.6.2.

NRC Safety Evaluation Report dated 7-11-73 concluded that the tracer tests did not provide sufficient justification for use in the accident diffusion model. Sections 2.5.5 and 2.5.6 are retained for historical purposes, refer to Section 2.5.4 for the applicable accident meteorology, item b of Subsection 2.5.6.2 discusses the relationship of the direction chart traces taken during the 1971 tests discussed in Subsection 2.5.5 to the observed diffusion of the tracer gas.

If the meander etTect is to be accounted for in determining ditrusion conditions from a large data ,

base, it should be assured that the direction vane is continuously responding to wind tluctuations.

An analysis of over 3800 hours0.044 days <br />1.056 hours <br />0.00628 weeks <br />0.00145 months <br /> of direction data taken during low wind speeds was made wh'ch indicates that reliable vane response is obtained for all but a small fraction of the 1 yr period of data evaluated. This study is discussed in item c of Subsection 2.5.6.2.

2.5-10 UPDATE-

/98

TMI.1/FSAR

' TABLE 2.5-15 (Sheet 1 of1)

VALUES OF ATMOSP11ERIC DIFFUSION FOR ASSESSING ACCIDENTAL RELEASES 8 i X/O (sec/m )

Time Interval (fir.) 2000 feet 2_tDilu 4

0-2 a6.8 x 10 2.0 x10'8 l 4

2-8 1.5 x 10 2.0 x10'8 8 - 24 4 4.5 x 10'8 4.0 x10 I

4

24 - 96 3.0 x 10'8 2.7 x10 i

.l 4

96 - 720 1.5 x 10'8 1.3 x10

  • This value established by Reference 29. l 4

08 2.5-47 UPDATE-

/98

TMI 1/FSAR

23. - Jane's,- All the Worlds Aircrall,-1967-68 and 1990-91 Editions, 24.- Gilbert / Commonwealth, Inc., Report 2661, Review of Low River Elmy _llydratic Arlalygs, June 1987.
25. Gilbert / Commonwealth, Inc., Report 2744, Reevaluglion of River Elow Hydraulics, April 1988.
26. U.S. Army Corps of Engineers (Baltimore District), Emergency Plan for Raystown Lake, Susquehanna River Basin, Raystown Branch Juniata River, September 1986.

I

27. Pennsylvania Dept. of Environmental Resources, Water Resources Bulletin, No. 6, December 1970.
28. USGS Water Resources Data - Pennsylvania, Water Year 1987, Volume 2.
29. Pickard, Lowe and Garrick, Inc. Letter, dated March 20,1979, From K. Woodard to R. Lengel- Metropolitan Edison,

Subject:

Re-Evaluation of Accident X/Q Values for TMI Using NRC Guide 1.xxx (Draft dated September 1978). Current Guide is 1.145.

2.10-3 '

! UPDATE i

/98 4 . l

TMI l/FSAR Therefore, the results of the NPSH calculations for the assumed accident flows are shown in Table 6.4 1. An additional calculation indicated that sutTicient NPSil would be available for the maximum flows, as limited by procedure, including instrument errors, recirculation flow and taking credit for sump subcooling (containment overpressure).

6.4.3 BASES OF LEAKAGE ESTIMATE While the reactor auxiliary systems involved in the recirculation complex are closed to the Auxiliary Building atmosphere, potentialleakage from component flanges, seals, instrumentation, and valves is accounted for. Potentialleakage through the recirculation system boundary valves, to tanks which arc",ented to the environment,is also accounted for.

The leakage murces considered are:

a. Valves
1) Disc leakage when valve is on recirculation system boundary
2) Stem leakage
3) Bonnet flange leakage
b. Flanges
c. Pump shaft seals j

6.4.4 DESIGN BASIS LEAKAGE The following design basis leakage values have been established and used in Chaptr 14 MilA analysis (Section 14.2.2.5) where appropriate for systems which recirculate post-accident RB sump fluid outside of the containment:

a) For system leakage to the Auxiliary Building - 0.3 gpm b) For system boundary valve leakage to the environment - 3.0 gpm Leakage evaluations, as part of NUREG-0578 requirements, were performed and leak reduction program initiated to further reduce leakage and to comply with 10 CFR 100 guideline limits.

Periodic inspections and tests are performed to ensure that total recirculation system leakage is less than the design basis leak rates above. Chapter 14 presents an analysis of the effects associated with the release of the radioactive fluid. It was concluded from this analysis that the leakage from the engineered safeguards systems to the Auxiliary Building and environment does not result in doses in excess of those allowed by 10 CFR 100 guideline limits.

6.4-3 UPDATE-

/98

TMI l/FS W TABLE 6.4-3 LEAKAGE OUANTITIES TO AUXILIARY BUILQJEQ i

HAS DEEN DELETED

)

6.4-9 UPDATE-

/98 o . J

L TMI l/FSAlt Opening the OAl Damper All D 39 proves the most beneficialin assuring maintenance of the hiain Control floom at 40.10 inches w g. Operating i'rocedure Op-l 104 19 and Abnormal Procedure lip 1203 34, " Control fluilding Ventilation System," require the operator to modulate OAl Darr.per All D 39 to gradual switch position 10.5 when entering emergency mode of operation. The bases for this switch position are provided in 1(eference i1, Appendix 11.

A positive pressure of 0.10 inches w g. is not a criteria for the ent:re CBE. The pressure in the remaining rooms of the CilE varies from a negative pressure of 0.15 inches w g. up to a positive pressure of 0.35 inches w g.

itadiological dose criteria used to establish habitability are; 1

l 1. Total integrated dose to Control lloom operator over a 30 day period following a release shall not exceed 5 Item whole body (gamma) and 30 item skin (beta).

2. Exposure b. e-- on occupancy of the Control floom by an operator for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following release, followed by 60 percent occupancy for the next 3 days, and 40 perc:nt occupancy for the remaining 76 days.

Control lloom operator dose evaluations consider dose from both externai and internal pathways.

The Maximum Ilypothetical Accident (MilA) source terms are considered. l 7.4.5.2.2 Control th111 ding.EnrelopslCBElSnlemlksign 7.4.5.2.2.1 Dc0nition of Contt0Uluildi a gfnYclopelClllD The Control lluilding Envelope (CBE) consists of all areas suved by the Control Building Emergency Ventilation System. The Control Building Envelope is defined in Section 7.4.5.2.1.

These areas are on three levels of the Th11 1 Auxiliary Building (control tower) which communicate directly with each other.

7.4.5.2.2.2 hatilatioRSyllemlkliign The Th11 1 Control lluilding Ventilation System (CBVS) during normal mode of operation serves the Control Buildhp Envelope (CBE) and the Controlled Access Area (Elevation 306'-0" excluding the ilot hiachine iiiup).

7.4-6 UPDATE-

/98

Th11 1/FSAR I

Permanent door seals were installed on all doors which are part of the CBE (as shown on Figure 7.4 2) prior to testing. The improved leak tightness resulted in achieving a positive pressure of 0.10 inches w g. or greater in the Control Room during norma!, emergency and various failure mode testing.

To show that the CBVS sati 6es the radiation dose criteria, the test data was evaluated to determine the amount of outside air used for pressurizing the CBE, to calculate air in61tration into the CBE where 0.10 inches w g. is not achieved.

Reference 11 details the technique employed in calculating the air infiltration rates into the CBE for the areas where 0.10 inches w.g. positive pressure is not achieved, and identi6es the resulting infiltration rates to each area. In61tration rates are determined by assuming a negative pressure of 0.30 inches w.g. in the areas of CBE when positive pressure is less than 0.10 inches w g. This is conservative in that no negative pressure was experienced in the CBE greater than 015 inches w g. The coeiservatism associated with this method assures the calculated operator dose is conservative. These infiltration rates are included in the dose analysis and a source term associated with leakage from the reactor building to the Auxiliary and Fuel llandling Building is applied to deteimine the Control Room operator exposure. This integrated Control Room operator dose was determined by analysis as described in Section 7.4.5.

7.4.5.2.4 Interac110nMillLQlltcLZoncundfinsMie-CoiltaluipsfgulpmeB1 The assumptions utilized in developing the source term applied to in leakage have been reviewed analytically and empirically tn insure they are bounding.

a. Erlease The radionuclides released from the Reactor Building (RB) from MilA source terms are l assumed to be released to the Auxiliary Building (AB) even though only about 4 percent of the RB surface area is adjacent to the AB/ Fuel llandling Building (FilB).

Approximately 40 percent of all RB penetrations are located in the AB/FilB. The Murphy Campe method for calculating releases, which was used for this analysis, is based upon surface area and not the concentration of penetrations Therefore, it is valid to compare the assumption of releasing 100 percent of the radionuclides into the AB to the amount of adjacent Rh surface area to assess a degree of conservatism.

7.4-8 UPDATE-

/98 L

Tht!.1/FSAR

b. ReactotDaiMiagleakage Ril Leakage rate was calculated consistent with the requirements of Regulatory Guide 1.4. This assumes a leak rate of 0.1 percent for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> aller an accident, followed by a leak rate of 0.05 percent for the remaining 29 days. The 0.1 percent leakage rate is based upon RIl design pressure during a LOCA. This method is conservative as the TMI l FSAR analysis determines that the peak Ril pressure lasts for only the initial 100 seconds following a LOCA. Within the first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, building pressure would be less than 5 psig.

c SQutts Term Trarupnt!

The dose model assumes that the Auxiliary and Fuel llandling Building ilVAC System (AFilBilVAC) provides complete mixing of the RB release. In additiori, a driving force is assumed which draws air from the AB & FilB to the patio. These assumptions are conservative.

l l

The AFIIBilVAC design promotes direct removal of RB leakage via its exhaust system through the AB & FilB exhaust stack. In addition, AFilBilVAC Supply Fans will be automatically shutdown upon detecting high radiation in the exhaust, as described in FSAR Sections 9.8.2 and 9.8.3. This mode of operation will increase the direct leakage removal capability of the system and preclude source term transport to the patio. Without supply fans, air supply to the AB & FilB is via infiltration induced by the exhaust fans.

This will draw air away from rather than into the patio.

The dose model assumes a continueus flow through the patio, equivalent to that if the patio llVAC exhaust system were not shutdown. This assumption provides a driving force for introduction of air from the AB & F110 to the patio. This is conservative since, by design, the patio ilVAC System will shutdown automatically as a result of either high radiation in the Control Room IIVAC Supply or an Engineered Safeguard Signal. This will substantially reduce the rate at which contamination can enter the patio and CBE.

Therefore, the TMI l llVAC system designs preciade sneak paths of concentrated source terms to the patio and the existing assumptions used for generating an internal pathway source term are bounding.

7.49 UPDATE-

/98

)

TMI.1/FSAR operator to take manual action to trip the normal fan and stan the emergency fan (All E 18B(A))in the event that All E 17A&B has failed to trip on the high radiation signal. T his action must be taken within 30 minutes. This is ample time to go to the MCC and trip the breaker irrequired.

Internal pathways for radioactive material transport to the CilE were evaluated assuming operation of the AB & FilB ventilation system. In the event of a loss of offsite power (LOOP) this system will not operate. Consequently mixing within the A & FilB generally will be limited to diffusion and local thermal air movement with some local cubicle recirculating systems remaining in operation. Two factors will mitigate the introduction of radioactive materialinto the CllE in this event:

First, the pos'olated entry point for contamination into the pat:0 area is approximately 70 meters of transport distance from the containment penetration area. DilTusion along this path will be slow and will not be assisted by any preferential air movement. Thus, a substantial period will exist before contamination can reach the CBE. Second, the only significant ilVAC system operating at this point will be the Control Room system. Since this system

produces a positive pressure in all or most of the CBE the driving force for i

transport will be away from the CBE and patio, depending on failures assumed.

This will funher mitigate the potential for contamination entering the CBE. These mitigating factors will provide time to permit restoration of offsite power prior to reaching operator dose limits.

7.4.5.2.5 }{adiological Prottc110D Reference 12 provides the details of the analytical model used to determine the radiation dose to the Control Room operators during radioactive releases. This modelis based on:

1. Criteria provided in the Murphy-Campe Report
2. 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> first shift after an incident
3. Beta and gamma dose only
4. Contribution from in leakages through exhaust damper (ail D 37) as well as that from other in-plant sources such as the patio (hallway) area.

A later evaluation (Reference 19) for a conservative power les el of 2772 MWt demonstrates that Control Room doses remain below the dose criteria of 7.4.5.2,1. This evaluation determined doses using a weighting factor that was computed by taking the quantity of each fission product released multiplied by the skin and whole body dose conversion factois. The dose values of Reference 12 wcre multiplied by these factors to obtain the revised Control Room Doses. The principal assumptions and methods remain applicahic to this later evaluation.

7.4 12 t

UPDATE-

/98

TMI l/FSAR Principal assumptions utilized in the dose analysis for external pathways are:

1. 100 percent of the radionuclides are released from the Reactor Building to the environment plus leakage from the ECCS to the Ausiliary Huilding and then to the environment (Reference 19). Intake to the CBVS is via dampers All D 39 and All D 37,
2. Containment leakage rate is based on 0.1 percent for first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, then 0.05 percent for remaining 29 days. 0.1 percent leakage is based on peak Reactor Building pressure during a LOCA. Leak rate is consistent with Regulatory Guide 1.4. Peak pressure lasts only for the initial 100 seconds aller a LOCA. Within the first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, building pressure is less than 5 psig. Credit is taken for the removal ofiodines by the Containment Spray System using sodium hydroxide spray solution.

Principal assumptions utilized in the dose analysis for internal pathways are:

1. 100 percent of the radionuclides are released from the Reactor Building and ECCS system leakage to the Auxiliary Building (Reference 19). Release to the Turbine Building is discounted.
2. Containment leakage rates are the same as for the external pathway analysis.
3. Auxiliary and Fuel llandling Building ilVAC System provides full mixing of Reactor Building release. The ilVAC design provides four air changes per hour. Unique flow patterns which preferentially direct the release to the patio are assumed. l
4. A driving force is assumed which draws air from the Auxiliary and Fuel llandling Building to the patio where air enters CBVS via infiltration to the CBE. This is conservative in that the Patio Ventilation System is shutdown during CBVS emergency mode and therefore no driving force exists.

The analytical modelis validated by in plant testing described in 7.4.5.2.3 and 7.4.5.2.4. A parametric study of total infiltration by source versus beta skin dose and gamma whole body dose was conducted and the results are tabulated in Reference 12.

7.4-13 UPDATE-

/98

m-t&mhkhWL &L -

r./

ll?

Il ! Thtl l/FS AR E!

16. GPUN Topical Report No. 027, " Final Verificat;on and Validation Report on Thil i Safety Parameter Display System", December 30,1985.
17. NRC Letter, "Three hiite Island Unit 1 - Instrumentation to Follow the Course of an i

Accident Required by Regulatory Guide 1.97 (TAC No. h151361)," dated hiarch 31, 1993.

18. Pickard, Lowe and Garrick, Inc. Report PLG 0370, *Probabilistic Risk Assessment of Offsite Releases initiated by a Toxic Chemical Release", dated July 30,1984.

h 19. Pickard, lame and Garrick, luc. Letter dated June 27,1997. Keith Woodard to A. Irani, GPU Nuclear, Inc.. "TMI I Control Room Dose Update (Revision 1)".

L 7.5-2 UPDATE.

/98 i

4 e TMI l/FSAR

8. Local temperature indication of the exhaust air in the inlet and outlet of the filter train.
9. Temperature detection devices in the ESF Filtration enclosure to activate the unit heaters and ventilating exhaust fan for ESF filtration e.. closure.

9.8.2.5 System Evaluation

a. FilBNVS Exhaust air from potentially radioactive areas is passed through roughing, llEPA, and iodine absorber filters for the removal of radioactive iodine.

Shutdown of Fuel llandling 13uilding supply units and isolation of the system from other ventilation systems is automatic in the event of a high radiation signal from the Fuel llandling Building exhaust duct monitor, RM A4 and the Fuel llandling Building area monitor, RM G9. In the ev:nt of a high radiation signal from the vent stack monitor, RM A8, the supply fans All E 10 and All E 11 will stop and the waste gas discharge valve will close.

I b. Fill 1ESFVS Exhaust air from the FilB during a fuel handling accident involving TMI 1 irradiated fuelis passed through the FilBESFVS for removal of a significant portion of the radioactive iodine, gasses and particulates.

The FIIBESFVS is in continuous operation during ali movements c,f TMl 1 irradiated fuel. The system has been demonstrated to maintain a negative pressure in the fuel handling area with respect to the outside environment (demonstrated via the TMI Unit 1 STP 141/19 Revision 0)(Reference 7).

All active components of the filtration trains and their associated fans and dampers are redundant. This fact enables the system to sustain a single active failure without loss of function. Changeover from the FilBESFVS train in operation to the back up train is manually initiated at a remote control panel by a single integrated control system. The train in back up mode is 9.8-16 UPDATE-

/98

TMI l/l S All J. Ileat, smoke, and vapor detectors ir. the Auxiliary lluilding Supply System to stop all supply and exhaust system and close off applicable fire dampers in the outside air supply system.

In addition to the smoke and fire protection devices, the system provides the following emergency control functions:

a. liigh radiation detected in the exhaust duct from the Auxiliary lluilding general areas initiates an alarm and stops the supply air fan to this area. The main exhaust system continues to operate, to exhaust air from the Fuel llandling Building, Auxiliary lluilding, llot Machine Shop, CB Controlled access area, from the decontamination facility cleaning area, and Waste Gas discharge line.
b. liigh radiation in the Auxiliary & Fuel llandling Building exhaust vent downstream of the filter bank initiates an alarm and stops the supply fans All E.10 and All E Il to the Auxiliary Fuel llandling fluilding general area. The exhaust system continues to operate, i

to maintain a negative pressure in both buildings. The waste gas discharge valve will also close on high radiation.

9.8 3.5 SnttutRulatioD Exhaust air from potentially radioactive areas is passed through roughing, llEPA, and iodine adsorber filters for removal of radioactive iodine.

Shutdown of supply units while exhaust units continue to operate is automatic in the event of high radiation signals from monitor RM A8 in the exhaust vent. liigh radiation from RM A6 will stop supply fan All E 11 only.

l Exhaust air from the hot instrument repair shop, located in the Chemical Storage Building on elevation 331 of the Auxiliary Building, is passed through roughing and llEPA filters for removal of potential radioactive particulates. lodine absorber filters are not required because of the absence of any iodine. The exhaust air from this repair shop is continuously sampled for potential airborne activity using a portable air sampler with a built in high radiation alarm Failure of the 11 EPA filters in the llVAC system for the new hot instrument repair shop could result in a release of radioactivity to the environment. Ilowever, the resultant quantity released would be 9.8 20 UPDATE

/98 Table 14.01

TMI.1/FS AR (Sheet 11 of 12)

Equipment And Related Systems Assumed To Function During Accident Analysis OPERATOR ACCIDENT TDV TSV BS BWST RBES SLRDS ACTION Steam Line Break YES YES YES YES YES YES- NO (14.1.2.9)

, Steam Generator YES YES NO NO NO NO YES Tube Failure (14.1.2.10)

Fuelllandling NO NO NO NO NO NO NO Accident' (l4.2.2.1)

Rod Ejection NO NO NO NO NO NO NO Accident (14.2.2.2)

Large Break NO NO YES YES YES NO NO LOCA l (l4.2.2.3)

Small Break NO NO YES YES YES NO YES LOCA, (14.2.2.4)

Maximum NO NO YES YES YES NO NO liypothetical Accident (14,2.2.5) 2 Fuelllandling Accident - Also takes credit for Fuel liandling Building ESF Ventilation System l llEPA and charcoal filters (90% efficiency) and Reactor Building Purge Exhaust System IIEPA and charcoal filters (70% efliciency),

l 14.0-11 UPDATE.

/98 q

TMI.1/FSM Table 14.01 (Sheet 12 of 12)

Equipment And Related Systems Assumed To Function During Accident Analysis ACCIDENT TDV TSV ilS IlWST OPERATOR riles SLRDS ACTION Waste Gas Tank NO Rupture NO NO NO NO NO NO (14.2.2.6)

I Loss of Feedwater Accident W/ stuck open PORY YES YES NO Adequacy of NO NO NO

$00gpm EFW' YES YES NO NO NO NO NO NO ARTS Evaluation YES YES NO NO I (14.2.2.7) NO NO NO Fuel Cask Drop NO NO NO NO Accident' NO NO NO (14.2.2.8) l

' Loss Of Feedwater Adequacy of 500gpm EFW- NNI conditioned sign a s are assumed to reach the Control Room console, the ICS, and the plant computer.

' area Fuel Cask Drop Accident - Also takes credit for key operated interlock of fuel handling crane. system to limit travel 14.0-12 UPDATE-

/98

TMI UFSAR Table 14.0-1 (Sheet 12 of 12)

Equipment And Related Systems Assumed To Function During Accident Analysis 4 OPERATOR l ACCIDENT TDV TSV 11 S IlWST riles SLRDS ACTION Waste Gas Tank NO NO NO NO NO NO NO Rupture (14.2.2.6)

Loss of Feedwater Accident W/ stuck open PORV YES YES NO NO NO NO NO Adequacy of YES YES NO NO NO NO NO

$00gpm EFW' l ARTS Evaluation YES YES NO NO NO NO NO (l4.2.2.7)

Fuel Cask Drop NO NO NO NO NO NO NO i

Accident I (14.2.2.8) 3 Loss Of Feedwater Adequacy of 500 ppm EFW - NN1 conditioned signals are assumed to reach the Control Room console, the ICS, and the plant computer.

  • Fuel Cask Drop Accident - Also takes credit for key operated interlock system to limit travel I area of fuel handling crane.

14.0-12 UPDATE-

/98

TMI.1/FS AR The atmospheric dispersion characteristics of the site are described in Section 2.5. A breathing rate of 3.47 x 10 " m'/S is assumed for the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> exposure. For the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> exposure, a breathing rate of 3.47 x 10 " m'/sec is assumed for the fitst 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, and a rate of 1.74 x 10 " m'/sec is assumed for the remaining 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />. For the 30 day exposure, a breathing rate of 2.32 x 10 4m'/sec is assumed.

The LOCA doses are bounded by the dose results of the MilA accident discussed in Appendix 14C. Thyroid doses and whole body doses are maintained below 10 CFR 100 suidelines.

b. Effects of Reactor fluilding Purging At times during the normal operation of the reactor it may be desirable to purge the Reactor fluilding while the reretor is operating. If a Large Break LOCA were to occur during purging operations, activity would be released to the environment. Assuming the worst

! mpture, essentially all of the reactor coolant will have been blown down. (The activity in l

the Reactor 11uilding is then taken to be the reactor coolant activity afler operation with I percent failed fuel). For this case, the purge valves will be completely closed in 5 seconds, but a small fraction of the Reactor lluilding atmosphere will escape through the purge valves before they close. The analysis assumes unrestricted flow through the purge line for the full 5 seconds closing time. No reduction in flow is assumed as the valve closes. The dose equivalent iodine released is 5.28 Ci. The resulting increase in 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> thyroid dose at the exclusion distance due to purge valve clo ing time is 1,84 rem. The additional thyroid dose that results from this release when added to the MilA dose for a loss of coolant accident

[ without purging is well below the 10 CFR 100 guidelines. Therefore, purging operations can be performed during reactor operation. An evaluation of purging as a means of controlling postaccident hydrogen concentration is in Appendix 14D.

14 2.2.4 SJtnallillenk Loss of CoojanLAccident 14.2.2.4.1 Identification Small break LOCAs are piping mptures whose break areas range from 0.0007 ft (3/8 inches diameter pipe) to as large as 0.5 f12 (10 inch diameter pipe),

The response of the primary system to a small break will greatly depend on break size, its location in the system, operation of the reactor coolant pumps, the number of ECCS trains functioning, and the availability of secondary side cooling. RCS pressuie and pressurizer level histories for various combinations of parameters are presented in order to indicate the wide range of system behavior which can occur for small LOCAs.

14.2 28 UPDATE-

/98

TMI.I/I'S All they were confined to the lleactor fluilding, but was not assumed to occur once they passed to the environment. The iodine activity and the nobic gas activity released l to the Reactor Building are shown in Tables 14.2 8 and 14.2 9. Accident analysis parameters and assumptions are tabulated in Table 14.2 20 (Reference 77) . l

b. Analysis and Results of Environmental Analysis As an upper limit on the consequences of this accident, it can be postulated that the leakage prevention systems are not completely effective in terminating allleakage. For this condition, leakage is assumed to continue at the design leak rate, and reduced afler 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> i

to one halfofits original value. The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> thyroid dose at the exclusion distance (Appendix 14C) and the 30 day dose at the low population zone distance as summarized in Table 14.2-20 are less than the guideline values of 10 CFR 100.

l The direct radiation dose to the whole body following the accident is shown on Figure l 14.2 53. No significant dose exists from this source at the exclusion distance. The dose to the whole body from the passing cloud has been calculated using the same meteorological conditions used for determining the thyroid dose. The whole body dose at the exclusion boundary (Appendix 14C) and low population zone distance as summarized in Table 14.2 20 are less than the guideline values of 10 CFR 100.

A discussion of the use of site meteorological data to estimate probability vs magnitude of dose for given radioactive material releases from the MilA is shown in Appendix 14E.

c. lodine removal sensitivity analysis using various reactor building spray and fan cooler combinations.

In Appendix 14C the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> exclusion boundary thyroid and whole body doses were calculated for the following reactor building cooling system combinations:

1) One spray header pump and one air cooling unit fan operating
2) Two spray header pumps and three air cooling unit fans operating 14.2-43 UPDATE-

/98 i

TMI 1/FSAR The spray remova; coeflicients, decontamination factors, and iodine source fiactions developed in Appendix 140 were used in the scmitivity analysis. It is concluded that the analyzed combinations of reactor building cooling systems have an insignificant affect on the whole body dose and that the use of the first combination of systems will result in a larger thyroid dose.

I

d. Effects of Enginected Safeguards Leakage During the Maximum liypoth:tical Accident ,

An additional source of fission product leakage during the maximum hypothetical accident l

can occur from leakage of the engineered safeguarJs external to the Reactor Huilding during the recirculation phase for long term core cooling. A detailed analysis of the potential leakage from these systems is limited by Technical Specification 4.5.4 and boundary valve leakage tests as described in Section 6.4. It is assumed that the water being recirculated from the Reactor Building sump through the external system piping contains 50 percent of the core saturation iodine inventory. This is the entire amount ofiodine release from the reactor cooling system. The 50 percent escaping from the reactor coolant system is consistent with Reference 53. The assumption that alliodine escaping from the Reactor Building is absorbed by the water in the Reactor Huilding is conservative since much of the iodine released from the fuel will be plated out on the building walls. The iodine is chemically bound to the sodium hydroxide and will not he released to the atmosphere.

Ilowever, it is conservatively assumed that iodine release does occur.

For ES leakage into the Ausillary Hullding. it is assumed that all of the iodine contained in water which flashes is released to the Auxiliary Building atmosphere. The flashing fraction of 1.25% used in the MilA dose consequence analysis is based on a constant sump temperature of 224 F. This value for flashing fraction is conservative based on the sump tempcrature profile during post accident recirculation, lodine released from the remaining water is calculated using a gas / liquid partition cocilicient of 9 x 4

10 . The activity is assumed to be released to the environment with no credit for holdup time or decontamination using the Auxiliary llullding Ventilation charcoal filters.

The analysis assumes 50% of the lodine plates out on the Auxiliary llullding surfaces.

Atmospheric dilution is calculated using the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> dispersion factors developed in Section 2.5. The leakage and resulting thyroid dose at the exclusion distance are shown in Table 14.2 20.

Another source of fission product leakage during the maximum hypothetical accident can occur through system valves to the llWST during the recirculation phase. Iodine releases are the limiting factor in this situation; hence, the two hour offsite doses at the site boundary are evaluated for this type of release. All of the coolant reaching the BWST is assumed to be in the liquid phaset therefore, the same liquid partition factor for iodine as assumed in the engineered safeguards leakage calculation (9 x 103)is assumed in this calculation. Recirculation is assumed to begin when there 14.2 44 UPDATE-

/98

'IMI 1/FSAlt is 300,000 gallons of enipty space in the llWST. The volumetric flow rate of air leaving the llWST is equal to the volumetric flow rate ofleakage into the ilWST. No credit in taken for lodine plateout in the llWST. Atniospheric dilution is calculated using the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> dispersion factors developed in Sectior. 2.6. The leakage and resulting thyroid dose at the exclusion distance are shown in Tahic 14.2 20.

14.2.2.6 htcha1TAniduphtts llupture of a waste gas tank would result in the release ofits radioactive contents to the Auxiliary llullding environment, l

14.2 44a UPDATE-

/98 i

l

  • Ih11 1/l'S AR A tani;is assumed to contain the gaseous activity evolved from degassing all of the reactor coolant following operation with I percent defective fuel. The coolant is then degassed an additional 99 percent according to the liquid / gas partitioning for lodine. The resulting waste gas inventory is I percent of the iodine and all of the noble gas activities associated with one reactor coolant volume. All of this activity is assumed to be released to the Auxiliary Huilding and environment as a puff release. The resulting leakage to the environment is I percent of the lodine and all of the noble gas activilles associated with one reactor coolant volume. No credit for lodine removal by the Auxiliary Hullding ventilation charcoal filters is assumed in this analysis.

The gaseous activity in the tank is listed in Table 14.2 21. The totalintegrated doses at the exclusion distance are:

2 llour Doses at Exclusion Distance Thyroid Dose 6.13 Rem Whole Hody Dose 1.79 Rem These doses are well below the limits of the 10 CFR 100 guideline.

14.2.2.7 Loss Of Fetdntet.Arddcat

a. Identification A loss of feedwater may result from abnormal closure of the feedwater isolation valves, control valve failure, or pump failure. The loss of feedwater flow results in a loss of heat sink, primary system heatup, increased pressurizer level and pressure, and reactor trip on high RCS pressure.

Astepla_nce Criteria For the transient analyses, the acceptance criteria chosen were Prevention of Pressurizer Fill and, Prevention of Saturated Condition in the RC llot Leg, The general criteria are as follows:

1) Core thermal power shall not exceed 112 percent of rated power, 14.2 45 UPDATE-

/98

TMI.1/FSAR TAllLE 14S4 (Sheet 1 of 1)

EISSIDREEDDUCT INVliNIORIES"'

EORllili.CORl!. Tilli AVERAGMSSIBillll AND Tile RIIACIOR COOLANT SYSTEM Atihily(CmicM i Isotope Total Corc* Total Gap

Inventory Activities (HCilg)

Kr 85m 2.40 x 10' $.49 x 10' 2.43 85 6.93 x 10' $.90 x 10' 9.75 87 4.39 x 10' 2.89 x 10' l.28 4

88 6.15 x 10' 8.93 x 10 3.95 Xc 131m 4.87 x 10' 7.10 x 10 4

2.68 133m 3.38 x 10' 9,97 x 10' 4.22 133 1.40 x 10' 9.02 x 10' 392 l 135m 3.63 x 10' l.34 x 10' O.485 l 135 2.40 x 10' 2.17 x 10' 8.37 138 1,29 x 10" 1.56 x 10' O.692 l 1 131 8.17 x 10' l.62 x 10' $.71 132 9.53 x 10' l 74 x 10' 1,92 l

133 1.41 x 10' 3.09 x 10' 6.07 134 1.77 x 10' l.80 x 10' O,757 l

4 135 1.40 x 10' 9.75 x 10 3.08

  • Based on C le 7 Bounding Analysis at 2568 MWt and at 460 EFPD, The source term -

is deliberately increased by applying a conservatism factor of 1.1

  • Zero decay time
  • Based on 1% failed fuel 14.2 54 UPDATE-

, /98

l TMI.1/FSAlt TAllt.li 14.2 18 (Sheet 1 of1) 10DINilACILYlD? IW1JMSILIR011Allla lwlopss AttiyhyEi) 1-131 4.09 x 10' l-132 4.77 x 10' l133 7.05 x 10' l134 8.85 x 10' l 135 7,00 x 10'

lodine activity released into the Reactor Iluilding atmosphere.

Ilasis of source term is given in Table 14.2-4, 14.2 74 UPDATE-

/98

Thti.1/l SAR TAllt.E 14.219 (Sheet 1 of 1)

NOILS DASSElliASILE110hth111ri l

Ileactor fluilding Activity lEntong (Ci)

Kr 83m 1.02 x 10' Kr 85m 2.40 x 10'

( Kr-85 6.93 x 10' Kr 87 4.39 x 10' Kr 88 6.15 x 10' Xe 131m 4.87 x 10' Xe 133m 3.38 x 10' Xe-133 1.40 x 10' Xe-135m 3 63 x 10' Xe-135 2.40 x 10' Xc-138 1.29 x 10' l 14.2 75 UPDATE-N8

_J

6 8 1Miel/l:S AR TABLE 14.2 20 (Sheet I of1)

ENVIRONMINAL DOSILS RESULTING FROMMIL6 Total 2 Ilour Dc4 8 uclusion Distance (Appendis 14C), sem' Tliyroid IW9

?!) ole Miy 6.0 Total 30 Day De

  • at Low I'cpulation Distance, rete Thyroid 13.0 Whole body <l Esclusion Area Th3 rold Dose (rem) From Henctor Building leakage 161.9 I;eactor 11uilding Design Leak Rate et%' day Engineered Safest a h Leakage lodine Concentration in HCS coolant 0.0477 1131 dose equiv Ci/mi leakage to Ausillary Hullding(gph) IN Flashing Fraction 1.25 %

leakage to HWST (gpm) 3 Thyroid Dose at Exclusion Distance. Rem 24.3 (Ausillary Hullding leakage)

Th 3rold Dose at Esclusion Distance Nem 0.12 (leakage to HWST)

Effects of Henctor Buildlag Purge (Th3 roid). Nem I N4 l

The current values of 2 hr and 30 day MilA doses on this table are based on power lesel of 256N Mwt The offsite doses wcre calculated in Reference 77 and consen atisely rounded upw ard to the nest wiiole integer. ,

14.2 76 UPDATE.

/98

TMI l/FSAR 9.0 CAILU.!E1101LDEAlihlOVAL COllEFICIENTS.AND DECONTAhtlNAll0N EACIOR.S 9.1 S1Rall2R0f h1ASDJEDIAN DI Ah1ETER The Thil 1 containment spray system is equipped with SPRACO hiodel 1713 spray nozzles.

Drop size distribution data for these nozzles are not available from the manufacturer. Ilowever, SPRACO was able to supply drop size distribution data for their hiodel 1713 A spray nozzles According to SpRACO, the model 1713A is almost identical to the hiodel 1713 and has the same drop size distribution, which is reported here as Figure 14B 1 The spray drop cumulative volume frequency calculated from these data is given on Figure 14B 2. The mass median diameter Fv (d;) = $0% obtained from Figure 14B 2 is approxima' tely d,',, = 1070 pm.

The calculation of d,,,is most heavily influenced by the largest drops. Figure 14B 1 shows an unusually large drop frequency in the size range 1725 to 1750 pm. Figure 14B 1 was obtained from a finite sample. The expected drop frequency in this size range will probably decrease as the sample size increases. Therefore, the true mass median diameter for the spray is probably a little '

less than 1070 pm. The reasoning serves as furtherjustification for use of equation 11.

To support the argument or Section 2.0 that k should not be evaluated at%, the lodine removal mean diameter was also calculated from the date of Figure 14B-1, according to equation 7. From these data, dm = 143 pm, which is considerably less than 5;.

9.2 REACTOR BUILDlNG SPIMESXSIEhi DESIGN PARAMPTliRS Relevant containment volumes (gas and liquid), rgay flow rates, and other required dimensions are given in Table 14B-1. The containment temperature and pressure were assumed to be 250 F and 75 psia,iespectialy typical conditions following a LOCA. The sodium hydroxide spray solution was assumed to be at the same temperaturc as the containment atmosphere and to have a pil of at least 8.0 l It is known that the spray solution temperature is actually less than 250 F and that once the spray is activated, the containment temperature and pressure will both decrease with time. It will be shown later that choosing temperature and pressure to be as high as realistically possible will result in the most conservative estimates of the removal coefficients.

l 14B 16 UPDATE-

/98

TMI 1/FSAR 9.3 CALCULAllON OF PilYSICAl, PROEERIIES Physical properties and related parameters such as terminal velocity and mass trar sfer coemeients were obtained at 250 F and 75 psia using the methods and data references of Sec lons 2.0 to 8.0.

The results are summarized in Table 14B 2.

9.4 CAI CULATIONS OF SPRAY REMOVAL &QEfMCIENTS AND RECONTAMIN61LON FACTORS Spray removal coemeients for elemental, organic, and particulate iodine are given in Table 148 3 for conditions when only one spray header is operating (minimum safety feature) and when both spray headers are in operation. For one header operating and a spray solution pil of 8.5, and a spray flow rate of 1500 gpm, the elementallodine removal coemeient is about An = 16.5 I

IIR. At was evaluated at the spray drop mass median diametel. For the same conditions At =

38.9 IIR if evaluated at the iodine removal mean diameter (dia = 143 pm) defined by equation 7.

Ilowever, this finding is oflittle practical significance since An is limited to 10 llR when credit is l

taken for instantaneous plateout, as will be done here llowever, if continuous platcout is assumed, no restriction is placed on At and use of At (di a)(option one of Section 2.0) would lead to nonconservative results.

To ensure that the calculated rernoval cocmcicats were conservative in every respect, values of l An, Ao, and 4 were also calculated for other eanditions of temperature and pressure, specifically 1

T = 225'F, P = 45 psia; and T = 200'F, P = lo.7 psia. These were assumed to be typical of intermediate and final conditions that might e>ist upon activation of the spray system. It was j found that values of Ar, Ao, and 4 were always lowest at 250 F and 75 psia. Values of An and Ao at the other T P conditions did not differ appreciably from those given in Table 14B 3 4 is independent of temperature and pressure.

Based on References 26 and 27, removal coellicients were recalculated assuming a spray solution pit of 8.0 and a spray flow rate of 1250 gpm per spray header. The results of this recalculation are also listed in Table 14H-3.

Decontaminatica factors for the three iodine forms are also given in Table 14B 3. These are independent of spray flow rate. According to equation 12, the decontamination factor for elementaliodine is applied before taking credit for instantaneous plateout. The decontamination factor for elementaliodine given in Table 14B 3 has therefore been reduced by a factor of 2.0 (see Section 7.0) since the computer code employed in the dose calculations takes instantaneous plateout credit prior to calculating the equilibrium concentration limit.

140 17 UPDATE-

/98 1

TMI.1/I'SAR TABLE 14B 1 (Sheet 1 of 1) l KliACIOR Bull. DING SPRAY SY.1 TEM. DESIGN PAltabl6IERS Spray Flow Rate, F One header operating 1250 gpm' Two headers operating 2500 gpm Average drop fall height, he 96 fl Containment Building free volume, Vc 2.16 x 10' f18 Containment Building sprayed volume, V 1.45 x 10' ff Containment Building unsprayed volume, Vc V 0.71 x 10' ff-Volume ofliquid in sump for use in equation 9, Vi, 11,245ff (assumed equal to primary system coolant volume)

Volume ofliquid in sump plus overflow from 63,153 fl' containment sump for use in equation 13, V'i.

  • Changed from the original 1500 gpm per header based on Reference 27.

14B 21 UPDATE-

/98

'Ihti l/FSAR TABLE 14B 2 (Sheet 1 of 2) l EllYSICAL PROPERTIES OF CONTAINMENTATAiDSFLIERii l AND SOD _ LUM llYDROX1DE SERAY SOLLLILON t

i 1. Containment Atmosphere Temperature, T 250*F Pressure, P 75 psia l

Gas density. P 4.565 x 10 gm/cm' Gas viscosity,0 2.266 x 10" gm/cm see Gasiodine dif1bsivity, Dg 2.661 x 10 cm'/sec Schmidt number, Sc 1.865 Gas phase mass transfer coemeient, K, 4.523 cm/sec

11. Spray Drop Solution Temperature, T 250F Spray solution pil e J 0* l Liquid density, p, 0.942 gm/cm' Liquid viscosity,11 2.210 x 10 ' gm/cm sec Liquid iodine diffusivity, Dc 6.972 x 10 cm'/sec Elemental lodine partition coeflicient, ils 1600 l Organic lodine partition coeflicient, llo 1.055 Organic iodine reaction rate constant, K 0.003 sec
  • Changed from original 8.5 hased on Reference 26, 14B 22 1

UPDATE-

/98

TMI.1/FS All TABLE 14I13 (Sheet I of 1) l Sl%W REMQYM10llE13GNilANilRECRETAMLNlllONfACIORS l i i Itemoval Coefficients For nil = 8,5 and Snray Flow Itate Of 1500 Gum Per llender l l l

One lleader Two lleaders l l ledintEetm Opsuiting Operating l 1  !

! Elemental, Ar,1110' 16.47' 31.07' f

! Organic, Aallir' O.008 0.015 l Particulate, Ap 1110' 3.64 7.28 l

1  !

i l Hemoval Cocmcients for nil = 8,0 and Spray Flow Itate of 1250 enm ner llender  !

I l

One licader  !

j imline Form Operatine

'l i Elemental, Ar,1111' 7.39' Organic, Ao 1110' 0,0072 d d

Particulate, A p,1110' 3.03 DncittaminatioAFxtets i Decontamination IndictEDDD Eactor (DF) 6 Elemental 100 Organic 1.04 Pasticulate 100

  • Must be reduced to 101110' since credit is taken for instantaneous plateout of elemental iodine.

6 For elemental iodine, DF is reduced by a factor of 2.0 from that given in Section 7.0 for use in the dose computer code, since credit is taken for instantaneous plateout.

Reference 26 d

Reference 27 14B-24 UPDATE-

/98

TMI.1/Fb AR TABLE 14B.3 (Sheet 1 of 1)

El%iY_RiiMQYALIQFHlCENTS ANDDl?CONIAMINATION FACTORS Removal Coemclents For nil = 8,5 and Surav Flow Rate Of 1500 Gum Per llender l One licader Two lleaders l Lostine Form Onciating OnctatLng  !

Elemental, At, llR 16.47' 31.07' Organic, AallR O.008 0.015 Particulate, A,,, llR 3.64 7.28 Removal Cocmcients for nil = 8,0 and Soray Flow Rate of 1250 unm per licader One IIcader lodine Form Oncratine Elemental. Ar, llR 7.39' Organic, Ao IIR O,0072 d ,

d Particulate, Ap,11R 3.03 RecDutaminatiOILDicl00i Decontamination lodine Form Factor (DF) 6 Elemental 100 Organic 1.04 Particulate 100 I Must be reduced to 10 llR since credit is taken for instantaneous platcout of elemental iodine.

6 For elementaliodine, DF is reduced by a factor of 2,0 from that given in Section 7.0 for use in the dose computer code, since credit is taken for instantaneous plateout,

' Reference 26 d

Reference 27 i

14B 24 UPDATE.

/98

TMl l/FSAR APPENDIX 14C ,

I liBLUATION OF ACG[111NUlOSE )

1.0 IlLYRQlD_QQSJLESTlh1ATliS Thyroid doses have been computed for Thil l using the spray removal coeflicients, decontamination factors, and iodine source fractions developed in Appendix 14B for the NaOli spray solution. The calculations were made for the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> dose at the exclusion boundary which is located approximately 610 m from the Containment fluilding 1.I DDSLLCALCULATION P10GiD11RE Equation 1:

The general equation used in the thyroid dose calculation model is:

5 3 l

D n,y = _X_ BR PwT 1 - [ fa [ Qa Q J=l J-i Subscript i refers to the physical form of the iodine, that is, elemental (i = 1 or E), organic (i = 2 or

0) and particulate (i - 3 or P). Subscript J refers to the iodine isotope, that is,1 131 (J = 1),1 132 (J = 2),1-133 (J = 3),1-134 (J=4). and 1-135 (J = 5). l

=

D6y thyroid dose for time period and distance being evaluated (rem)

X/Q = average atmospheric difrusion factor for the time period and distance being evaluated (sec/m')

BR = average breathing rate for the time period (m'/sec)

Po = reactor power level (h1W)

T =

time period over which dose is to be calculated (day)

L =

containment leakage rate (percent of total containment volume per day) fe; =

conversion factor for isotope J (rem / curie)

=

Qa average amount ofiodine ofisotope J and form i released from the containment atmosphere during time period (curies /htW) 14C-1 UPDATE-

/98

TMI.I/FS All The values of the Qa are a function of the lodine form and isotope, the spray removal coeflicient for the iodine form (independent ofisotope), the sprayed and unsprayed containment volumes, and the degree of mixing (circulation) between the sprayed and unsprayed volumes. The initial amount ofindine of each form and isotope available for release from the containment is given by Equation 2:

=

Qas u sQ3 Where:

=

Qoa initial amount ofindine of form i and isotope J available for release from the containment (curies /MW)

=

Q2 total iodine as isotope J in an equilibrium core (curievMW)

=

ui fraction of total iodine in equilibrium core (all isotopes together) available in form i for release from the containment atmosphere.

The a, arejust the source terms presented in Section 6.0 of Appendix 14B for elemental, organic, and particulate iodine, respectively (converted from percent to fraction). Since credit is being taken for instantaneous p!ateout, or (i = E for elemental iodine) is reduced by a factor of 2.0 from that given in Section 6.0 of Appendix 14B. Values of u, are given in Table 14C 1. Values of Qs and fc3 are given in Table 14C-2.

All assumptions for the thyroid dose calculations are summarized in Tables 14C 1 and 14C 2.

The atmospheric diffusion fauor, X/Q, at the exclusion boundary for the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> time period is that provided by Reference 1. The value was derived from two years of site meteorological tower data which has a recovery rate of 90 percent. It is assumed that the plant operates at Po = 2568 MWt priot to undergoing a loss ofcoolant accident.

1.2 10 DINE REMOVAld10 DEL A two compartment model is used to calculate spray iodine removalin the containment since certain areas are not reached by the spray droplets, Mixing between sprayed and unsprayed containment volumes is assumed to be that provided by the analysis of Reference 2. For l l

14C-2 UPDATE-

/98

TMI.1/FSAR 2.0 W110LE BOD _Y.DDJE ESTIMATilS Whoie body dose estimates due to the iodines and noble gases released from the containment were calculated using the methodology of Reg. Guide 1.4. The calculations were made for the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> dose at the exclusion boundary.

2.1 DD03fUiSTJhlAIES The general equation used in the whole body dose calculation model is Equation 5:

16 l

Dwu = (.247)(X/Q) L P T E Ej Cs J=1 (X/Q), L, P, and T are as previously defmed and have the same values as for the thyroid dose estimates.

Dwn = whole body dose for the time period and distance being evaluated Es = average beta-gamma energy for isotope J (Mev/ dis)

C3 = average amount ofisotope J released from the containment atmosphere during time period (curies /MW)

The sixteen isotopes relevant to the whole body dose analysis are given in Table 14C.2. Values l of Es initial amount in core, and halflife are also given. It is assumed, for the noble gases, that the entire quantity of each isotope in the core is available for release from the containment.

2.2 JSOTOPE REMOVAL MQDEL For the five iodine isotopes, the 6 values are obtained from the iodine removal model used for

{

the thyroid dose estimates by summing over the three iodine forms (elemental, organic, and particulate).

Equation 6:

3

_C3 = E Qa i=1 14C-4 UPDATE-

/98 Containment sprays are not capable of removing noble gases from the containment atmosphere.

TMI 1/FSAR

. The equations used to describe the rate of change in the quantity of noble gases in the containment atmosphere as a function of time are of the form Equation 7; dQ = -[Lj + h) Cs dt where Li and k are as previously dermed (k has the same value as for the thyroid dose

- estimates) and C1 = amount ofisotope J in containment at time t (curies /MW) l L

Equation 6 can be solved analytically to yield Equation 8:

Ca = C.a exp (-(Au + At) t}

Then Cor is the initial quantity ofisotope J available for release from the containment (Table 14C-2). Equation 8 can be integrated (averaged) over the time period to yield the Cs.

For the noble gases, whole body dose estimates are independent of the containment spray operating parameters since noble gases are not removed by the spray, However, the contribution to the whole body dose from the iodine isotopes is dependent on spray system parameters.

Therefore, the total whole body dose from all isotopes will differ depending upon the number of spray header pumps and air cooler unit fans in operation.

14C-5 UPDATE-

/98 1

TMI l/FSAR 3.0 DOSJdESULTS Thyroid doses have been computed for the assumed accident conditions. Results obtained for minimum safety features (one spray header pump and one air cooling unit fan operating) show that thyroid doses would be about 189 rem for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> at the exclusion boundary. Thh includes the dose from Reactor Building Leakage, Engineered Safeguards Leakage to the Auxiliary Building and BWST, and the effects of Reactor Building purging. The thyroid dose was calculated in Reference 3 and conservatively rounded upward to the next w hole integer.

This estimate is well below 10 CFR 100 guidelines.

Whole body doses have been computed for assumed conditions aver the course of the accident.

Results obtained from minimum safety features (one spray heater pump and one air cooling unit fan operating) show that whole body doses would be about 6 rem for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> at the exclusion boundary, This is well below 10 CFR 100 guidelines. The w hole body dose was calculated in Reference 3 and consenatively rounded upward to the next whole integer, it is concluded that use of sodium hydroxide spray solution provides adequa.e removal ofiodine to maintain thyroid doses below guidelines with minirnum safety features operable. Whole body doses are also maintained below guidelines.

l l

14C-6 UPDATE-

/98

i TMI 1/FS AR I

4.0 REFERENCES

1. Pickard, Lowe and Garrick, Inc., Letter dated March 20,1979, froin K. Woodard to R. Lengel-Metropolitan Edison,

Subject:

"Re-Evaluation of Accident X/Q Values for TMI Using NRC Guide 1.x11"(Draft dated September 1978). Current Guide is 1.145,

2. GPUN Calculation C-1101-214 E610-016, Rev. O," Reactor fluilding Atmosphere Miting Analysis",7/14/97.
3. GPUN Calculation C-1101-202-E260-329, Rev. 0,"OITsite Dose Analysis of the MilA for 2568 MWi",7/10/97 l

Y 14C-7 UPDATE-

/98

TMI 1/FSAR

, TABLE 14C-1 (Sheet 1 of 1)

ASSUMPTIONS USED FOR DOSILCALCULATIONS Time period, T 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Distance to exclusion boundary 610m Average atmospheric diffusion factor, X/Q 4 3 6.8x 10 sec/m (Ret 1) l 3

Breathing rate, BR 3.47 x 10" m /sec Power level, Po 2568 MWt Containment leakage rate, L 0,1%/ day l Elemental iodine initial release fraction, ai 0.23875*

Organic iodine initial release fraction, ao 0.0100 Particulate iodine initial release fraction, up 0.0125

- Spray removal coefficients, Ai see Table 148-3 Decontamination factors, DFi see Table 14B 3 Total containment free volume, Vc 2.16 x 10' R3 3

Containment Building sprayed volume, Vs 1.45 x 10' R Containment Building unsprayed volume, Vu 0.71 x 10' ff Mixing Flow Between Sprayed and Unsprayed Volumes, 3

FA, With one fan and One spray pump Operrfing 100,000 ft / min (Ref. 2) l l

  • Credit is taken for instantaneous plateout 14C-8 UPDATE-

/98 i

TMI l/FSAR TABLE 14C-2 (Sheet 1 of 1)

ASSliMPTIONS FOR DOSE CALCULATIONS Dose Conversion Average Factor llalfLife Energy (Thyroid)

Isotone (hr) Curles in Core * (Mev/ dis) (rent'Ci) 1-131 193.2 8.17 x 10' O.580 1.48 x 10' l132 2.26 9.53 x 10 7

2.573 5.35 x 10 4

5 1-133 20.3 1.41 x 10" 1.074 4.0 x 10 3

l-134 0.88 1.77 x 10" 2.889 3.73 x 10 i-135 6.68 1.40 x 10" 2.472 1.24 x 10' Kr-83M 1.83 1.02 x 10' O.003 Kr-85M 4.4 2.40 x 10' O.401 4 5 Kr-85 9.43 x 10 6.93 x 10 0.227 Kr-87 1.27 4.39 x 10' 2.105 Kr-88 2.8 6.15 x 10' 2.40 5

Xe-131M 283.2 4.87 x 10 0.164 Xe-133M 54.2 3.38 x 10' - 0.233 Xe-133 126.5 1.40 x 10" 0.196 Xe-135 9.14 2.40 x 10' O.564 Xe-135M 0.26 3.63 x 10' O 431 Xe-138 0.24 1.29 x 10" 1.736

  • Per Table 14.2-4 14C-9 UPDATE-

/98

t l

I l

ENCLOSURE 3 Certificate of Service for TMI-l TSCR No. 266

.

Retrieved from "https://kanterella.com/w/index.php?title=ML20198E794&oldid=3007611"