Text

Proposed Tech Specs 3.6.C Re Coolant Chemistry & TS 3/4.17B Re Control Room Emergency Filtration Sys
	ML20137W220
Person / Time
Site:	Monticello
Issue date:	04/11/1997
From:	; NORTHERN STATES POWER CO.
To:	;
Shared Package
ML20137W196	List: ML20137W192; ML20137W214; ML20137W220;
References
	NUDOCS 9704180058
	Download: ML20137W220 (21)
	v • d • e

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>R s 88 0 0: 3.0 lR U LIMITING CONDITIOilS FOR OPERATION 4.0 SURVEILIANCE REQUIREMENTS -

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y. 4. The reactor vessel head bolting studs 4. When the reactor vessel head studs are 1 "

a' shall not be under tension unless the under tension and the reactor is in the temperature of the vessel head flange Cold Shutdown Condition ~the reactor ES" G and the head are >70*F. vessel shell flange temperature shall be permanently recorded. *

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'6 A C. Coolant Chemistry C. Coolant Chemistry

) 1. The steady state radiciodine concentrati 1. (a) A sample of reactor coolant shall be T l in the reactor coolant shall not exceed taken at least every 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months and

! ? I microcuries of I-131 dose equivalent per j gram of water. i D

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11ases Continued 3.6 and 4.6 C. Coolant chemistry

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-f C ""' -fI'" e p h cle-t per grrr ef e ter 'n I the rereter cect e t s -- *a-t e d thavot-system-can ie ._M1"~ be. reached-iE the-gr-oss-sadi attuity- i n-th e-ga s eous-o f Huent's--ere-nea r ac-puolonged-.shutde" . _ the cleanup-deminera44eer. In the event of I a steam line rupture outside the drywell, the " " staff c culations show the resultant radiological dose y at the -- rt_t r'.tr- boundary d5 ) to 'be less than 30-R ; t th: Syre'd. This dose was calculat'ed on n' the basis of the radiolodine concentration limit of uCL of I-131 dose equivalent per gram of water, ar w h ric iiffu.Len-Seem-e- equ'r:1c. r clevate d r:1 ase-of--40 Pcmquill ,y: "

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4 0. m a m/aat aa;; relcar: i celetlen-valve-elesuee-eine-e f-five-s econd s y cf25,999perde.J 1 The The reactor coolant sample will be used to assure that the limit of Specification 3.6.C.1 is not exceeded.

radiolodine period of 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months .

concentration would not be expected to change rapidly during steady state operation over a In addition, the trend of the radioactive gaseous effluents, which is continuously monitored, is a good indicator of the trend of the radioiodine concentration in the reactor coolant. When a significant increase in radioactive gaseous effluents is indicated, as specified, an additional reactor coolant sample shall be taken and analyzed for radioactive lodine.

' Whenever an isotopic analysis is performed, a reasonable effort will be made to determine a significant percentage of those contributors representing the total radioactivity in the reactor coolant sample.

Ilsually at least 80 percent of the total garcna radioactivity can be identified by the isotopic analysis.

It has been observed that radiciodine concentration can change rapidly in the reactor coolant during transient reactor operations such as reactor shutdown, reactor power changes, and reactor startup if failed fuel is present. As specified, additional reactor coolant samples shall be taken and analyzed for reactor operations in which steady state radiciodine concentrations in the reactor coolant indicate various levels of iodine releases from the fuel. Since the radiolodine concentration In the reactor coolant is not continuously measured, reactor coolant sampling would be ineffective as a means to rapidly detect gross fuel element failures. Iloweve r, some capability to detect gross fuel element failures is inherent in the radiation monitors in the off-gas system and on the snain steam line.

Materials in the primary system are primarily 304 , stainless steel and zircaloy. The reactor water chemistry limits are established to prevent damage to these materials. The liinit placed on chloride concentration is p

to prevent stress corrosion cracking of the stainless steel.

3.6/4.6 .

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In the event of a postulated high energy line break in the RWCU system outside the 148 REV drywell, calculations show the resultant radiological dose at the exclusion area boundary to be less than 10% of the dose guidelines of 10CFR100. This dose was calculated on the basis of the radiciodine concentration limit of 0.25 pCi of I-131 dose equivalent per gram of water.

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3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILIANCE REQUIREMENTS ,

b. When both filter trains'of the control room emergency filtration system are inoperable, restore at least one. train to operable stattis within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in hot shutdown within the next 12

) hours following the 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and I

reduce the reactor coolant water temperature to below 212*F within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

2. Performance Requirements 2. Performance Requirement Test

a. Periodic Requirements a. At least once per 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of system combu,4 operation; or once per operating

. (1)The{resultsofthe in-place DOP cycle, but not to exceed 18 months, tests at 1000 cfm (110%) on llEPA whichever occurs first; or following D filters shall show /<M DOP . painting, fire, or chemical release penetration. C5 0,3% while the system is operating that 7 could contaminate the HEPA filters or p (2) The results of in-place charcoal adsorbers, perform the

> halogenated hydrocarbon tests at following:

LD 1000 cfm (i10%) on charcoal banks D show p penetration. (1) In-place DOP. test the llEPA filter e- l o.5 */o banks.

W (3) The results of laboratory carbon sample analysis shall show-ap499 (2) In-place test the charcoal adsorber.

p mm il.f1 1.Jidm m..~ ul efficim.cy banks with halogenated hydrocarbon

.m .. t m J a t 00 ' , 0 % " . ". . _ tracer.

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(3) Remove one carbon test canister

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--- from the charcoal adsorber.

Subject this sample to a 3O'C p h. 95% /d ^Y. ve laboratory analysis to verify methyl iodide removal efficiency.

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(4) Initiate from the control room 1000 cfm (ilot) flow through both ,

trains of the emergency flitration treatment system. .

3.17/4.17 229w REV

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3.0 LIMITItic C0fiDITIONS FOR 0? ERAT 10N 4.0 SURVEILIANCE REQUIREMENTS

b. The system shall be shown to be b.

operable with; At- least once per operating cycle, but not to exceed 18 months, the following (1) Combined filter pressure drop conditions shall be demonstrated for

<8 inches wat~er. each emergency filtration system i train: ,

6 (2) Inlet heater power output Skw (1) Pressure drop across the combined 10%. ~

filters of each train shall be (3) Automatic initiation upon receipt mensared at 1000 cfm (i10%) flow rate.

of a high radiation signal. '

(2) Operability of inlet heater at nominal rated power shall be verified.

(3) Verify that on a simulated high radiation signal, the train switches to the pressurization mode of operation and the control room is maintained at a positive pressure with respect . to adjacent.

areas at the design flow rate of 1000 cfm (ilot). -

3. Post Maintenance Requirements 3. Post Maintenance Testing C.om W A d

a. After any m ntenance or testing that a. After any maintenance or testing that could affee the HEPA filter or HEPA could affect the leak tight integrity filter mou ing frame leak tight inte- of the HEPA filters, perform in-place grity, the esults of the in-place DOP DOP tests on the HEPA filters.

(1 tests at 1000 cfm (i10%) on HEPA filters yr shall show f DOP penetration. b. After any maintenance or testing that p CdE CD,3 3/.

y b. After any maintenance or testing that could affect the leak tight integrity of the charcoal adsorber banks, could affect the charcoal adsorber t

to p leak tight integrity, the results of I

perform halogenated hydrocarbon tests in-place halogenated hydrocarbon tests on the charcoal adsorbers.

[tr at 1000 cfm (1101) on charcoal l ndsorberbanksshallshowg penetration.

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During normal operation, the CRV system recirculates the air in the-control room envelope as needed. During a high radiation ev,ent, the

Control Room Ventilation System continues to operate and the Control 3.17 Bd:as Room Emergency Filtration Train system will start automatically to A. pressurize the control room protective envelope. The Emergency Control Room Ventilation System C Filtration Train system can also be started manually. ~

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The Control Room Vent 11ati v @on 5ystem provides air conditioning and heating as required to maintain a ..

suitable environment Filtration Train (EFT)inbuildine.

the main The control room and portions of the first and second floors of the Emergency pnee s m o en i m,. n en inny vor Yve..i n 9 m or r^-tr-! r- n ' :

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All analyzed toxic substances for their affectwhich onare thestored controlonroom site or stored / shipped within a 5 mile radius of the plant have been operators. It has been concluded that the operators will have at toxic substance which are transported on highways within 5 before least two minutes to don protective breathing apparatus incapacitation limits are exceeded. For the probability of a release from the plant due to incapacitation of the miles of plant, it has been determined that sufficiently low that this scenario may be excluded. Protection for toxicoperators the paused by a spill is chemicals is provided through operator training.

B. Control Room Emercency Filtration System -

The Control Room Emergency Filtration System assures that the control room operators will be adequately protected against the effects of radioactive leakage which may by-pass secondary containment following a loss

" -- - I nof t c coolant accident or radioactive releases from a steam line break accident. The system is designed to ma slightly re ppressurize the control room on a radiation signal in the ventilation air. Two completely redundant trains d y yn g o . (, oth ch gg r qu[ -fo C, 3 C)f Each train has a filter u it consisting of) prefilter, IIEPA fillers, and charco 1 ad orbers. The IIEPA filters remove particulat s from the Contr 1 Room pressurizing ..r and prevent o ng of the lodine adsorbers.

The in-place The charcoal test resu sadsorbers should indicat are in aalled IIEPAto remove filter '- any

'- radioipdines ie pressurizing air.

charcoal adsorber 1 "- , of less than of less than through DOP testing and a sample results should indicate a radioactive methyl iodide ...mm.ithrough halogenated"- hydrocarbon g - 1 : t n =testing.

= underThe testlaboratory c conditions similar to expected accident conditions. System flows should . r ri - be i - 7.-

verificationoftheseperformanceparameterscombinedwiththequalificatio%eartheirdesignvalues. The Si d testing conducted on new filters and adsorbers provide a high level of assurance that the Emergency titration System will perform as 4

p to Oredicted CFR 50 in reducing doses to plant personnel below those level stated n Criterion 19 of Appendix A q pg.MA-ion 04 It.n OAA Or Dose calculations have been performed for the Control Room Emergency Filtration Syste which show that, G.97[ 9% f assuming ef t standby gas treatment system adsorption and filtration efficiency and control room ]

emergency organ do. s tiltration system adsorption and flitration efficiency and radiolodine plateout, whole body and remain within the NR6 guidelines of 5 rr -^' ?" r^ . respectively.

05A I 3.17 BASES lo cf1LSo A p; r. A, G w pesIa 229y WW REV

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'mntt ol Poom Venitilation System s Control room air temperature is checked each shift to ensure that the continuous duty rating for the instrumenttation and equipment cooled by this system is not exceeded. ,

Demonstrating automatic isolation of the control room using simulated accident signals assures control s

room isolation under' accident conditions.

B. Control Room Emergency Filtration System Air flow through the filters and charcoal adsorbers each month assures operability of the system.

The frequancy of tests and sample analysis is necessary to show that the llEPA filters and charcoal adsorbers can perform as evaluated. The charcoal adsorber tray is installed which can accommodate a sufficient number of representative adsorber sample modules for estimating the amount of penetration the system adsorbs though its life. Sample modules will be installed with the same batch characteristics as the system adsorbent and will be withdrawn for the methyl iodide removal efficiency tests. Each module withdrawn will be replaced or blocked off.4 In-place testing procedures will be established utilizing applicable sections of 9:g :12tery c:id- 1. 53Av4ei; ; 2 and ?"! M51a d""? t .f ard; as procedural guidelines only. 9 If test results are unacceptable, all a # rbent in the train is replaced. Any HEPA ()

filters found defective are replaced. gr_ N Slo- 17 87 y Pressure drop across the combined liEPA filters and charcoal adsorbers of less tha tn inches of water at T the system design flow rate will indicate that the filters and adsorbers are not,c ogged by excessive d amounts of foreign matter.

h gg Demonstrating automatic control room pressurization using simulated accident signals assures control room pressurization with respect to adjacent areas under accident conditions.

?

The laboratory methyl iodide test of the carbon adsorber is to be performed in accordance with ASTM D 3803-89, " Standard Test Method for Nuclear-Grade Activated Carbon. "

w The individual test results obtained from in-place penetration testing for the HEPA filter upstream of the charcoal adsorber 4.17 BASES and of the HEPA filter downstream of the charcoal adsorber unit 229z are to be multiplied together to determine the penetration of REV the combination of the two filters in series as a unit to satisfy the criteria of the specifications.

Exhibit C Monticello Nuclear Generating Plant

'l Revision One to License Amendment Reauest dated July 26.1996 '

i Revised Technical Specification Pages l Exhibit C consists of the Technical Specification pages with the proposed changes ,

incorporated Existing pages affece

. t d b ythis change are listed below: 1 f.aQR l l

123 148 ;

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3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS

4. The reactor vessel head bolting studs 4. When the reactor vessel head studs are '

shall not be under tension unless the under tension and the reactor is in the temperature of the vessel head flange Cold Shutdown Condition,-the reactor and the head are 270*F. vessel shell flange temperature shall be permanently recorded.

C. Coolant Chemistry C. Coolant chemistry

1. The steady state radiolodine. 1. (a) A sample of reactor coolant !

concentration in the reactor coolant shall be taken at least shall not exceed 0.25 microcuries of I- every 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months and l 131 dose equivalent per gram of water.

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Bases Continued 3.6 and 4.6 C. Coolant Chemistry In the event of a steam line rupture outside the drywell, calculations show the resultant radiological dose at the exclusion area boundary to be less than 101 of the dose guidelines of 10CFR100. This dose was calculated on the basis of the radiciodine concentration limit of 2 pCi of I-131 dose equivalent per gram of water. In. the event of a postulated high energy line break in the RWCU system outside the drywell, calculations show the. resultant radiological dose at the exclusion area boundary to be less than 10% of the dose guidelines of 10CER100. This dose was calculated on the basis of the. radiciodine

, concentration limit of 0.25 pCi of I-131 dose equivalent per gram of water.

i The reactor coolant sample will be used to assure that the limit of Specification 3.6.C.1 is not exceeded. -The radiciodine concentration would not be expected to change rapidly during steady state operation over a period of 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months . In addition, the trend of the radioactive gaseous effluents, which is continuously monitored, is a good indicator of the trend of the radiciodine concentration in the reactor coolant. When a significant increase in radioactive gaseous effluents is indicated, as specified, an additional reactor coolant sample shall be taken and analyzed for radioactive iodine, i

Whenever an isotopic analysis is perfermed, a reasonable effort will be made to determine a significant percentage of those contributors representing the total radioactiivity in the reactor coolant sample. Usually at least 80 percent of the total gamma radioactivit'y can be identified by the isotopic analysis.

It has been observed that radioiodine concentration can change rapidly in the reactor coolant during transient reactor operations such as reactor shutdown, reactor power changes, and reactor startup if failed fuel is present. As specified, additional reactor coolant samples shall be taken and analyzed for reactor operations in which steady state radiciodine concentrations in the reactor coolant indicate various levels of iodine releases from the fuel. Since the radiciodine concentration in the reactor coolant is not continuously measured, reactor coolant sampling would be ineffective as a means to rapidly detect gross fuel element failures. However, some capability to detect gross fuel element failures is inherent in the radiation monitors in the off-gas system and on the main steam line.

Materials in the primary system are primarily 304 stainless steel and zircaloy. The reactor water chemistry limits are established to prevent damage to these materials. The limit placed on chloride concentration is to prevent stress corrosion cracking of the stainless steel.

4 3.6/4.6 148 REV

3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILIANCE REQUIREMENTS

b. ' When both filter trains of the control ,

room emergency filtration system are :

. inoperable, restore at least one train to i operable status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in '

hot shutdown within the'next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

., following the 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and reduce the reactor coolant water temperature to below 212'F within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

2. Performance Requirements 2. Performance Requirement Test

a. Periodic Requirements a. At least once per 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of system operation; or once per operating cycle,

l (1) The combined results of the in- but not to exceed 18 months, whichever place DOP tests at 1000 cfm (i101) occurs first; or following painting, ;

on HEPA filters shall show 50.3% fire, or chemical release while the j DOP penetration. system is operating that could contaminate the HEPA filters or charcoal '

(2) The results of in-place adsorbers, perform the following: )

halogenated hydrocarbon tests at* '

1000 cfm (i10%) on charcoal banks (1) In-place DOP test the HEPA filter l show 50.3% penetration, banks. !

(3) The results of laboratory carbon (2) In-place test the charcoal adsorber sample analysis shall show the banks with halogenated hydrocarbon l methyl iodide penetration $0.4% tracer. !

when test at 30*C and 95% relative ;

humidity. (3) Remove one carbon test canister fron ;

the charcoal adsorber. Subject this sample to a laboratory analysis to verify methyl iodide removal i efficiency. j (4) Initiate from the control room 1000 cfm (110%) flow through both trains of the ;

emergency filtration treatment system. ;

r 229w [

3.17/4.17 REV l l

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. i 3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILIANCE REQUIREMENTS -

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-b. The system shall be shown to be b. At least once per operating cycle, but operable with: not to exceed 18 months, the following conditions shall be demonstrated for i

' (1) Combined filter pressure drop $8 each emergency filtration system train:

inches water.

(1) Pressure drop.across the combined

-(2) Inlet heater power output Skw i filters of each train shall be 101. measured at 1000 cfm (1101) flow rate.

(3) Automatic initiation upon receipt of a high radiation signal. (2) Operability of inlet heater at

'

nominal rated power shall be j verified. i (3) Verify that one simulated high radiation signal, the train switches to the pressurization mode of operation and the control room is maintained at a positive pressure with respect to adjacent areas at the design flow rate of 1000 cfm (110%).

3. Post Maintenance Requirements 3. Post Maintenance Testing

a. After any maintenance or testing that a. After any maintenance or testing that '

could affect the HEPA filter or HEPA could affect the leak tight integrity of i filter mounting frame leak tight the HEPA filters, perform in-place DOP l integrity, the combined results of the tests on the HEPA filters.

in-place DOP tests at 1000 cfm (1101) on i l HEPA filters shall show 50.3% DOP b. After any maintenance or testing that penetration. could affect the leak tight integrity of the charcoal adsorber banks, perform

, b. After any maintenance or testing that halogensted hydrocarbon tests on the could affect the charcoal adsorber leak charcoal adsorbers. !

tight integrity, the resulta of in-place }

halogenated hydrocarbon tests at 1000 ;

cfm (1101) on charcoal adsorber banks !

l shall show $0.31 penetration.

4 3.17/4.17 229x REV i i

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i 3.17 Bases A. Control Room Ventilation System .

The Control Room Ventilation (CRV) System provides air conditioning and heating as required to maintain a suitable environment in the main control room and portions of the first and second floors of the E;nergency Filtration Train (EFT) building. During normal operation, the CRV system recirculates the air in the control room envelope as needed. During a high radiation event, the Control Room Ventilation System continues to operate and the Control Room Emergency Filtration Train system will start automatically to pressurize the control room protective envelope. The Emergency Filtration Train system can also be started manually.

All toxic substances which are stored on site or stored / shipped within a 5 mile radius of the plant have been analyzed for their affect on the control room operators. It has been concluded that the operators will have at i least two minutes to don protective breathing apparatus before incapacitation limits are exceeded. For toxic substance which are transported on highways within 5 miles of the plant, it has been determined that the ,

probability of a release from t plant due to incapacitation of the cperators caused by a spill is sufficiently low that this scenario may be excluded. Protection for toxic chemicals is provided through operator training.

B. Control Room Emergency Filtration System ,

The Control Room Emergency Filtration System assures that the control room operators will be adequately protected against the effects of radioactive leakage which may by-pass secondary containment following a loss of coolant accident or radioactive releases from a steam line break accident. The system is designed to slightly pressurize the control room on a radiation signal in the ventilation air. Two completely redundant trains are provided.

Each train has a filter unit consisting of a prefilter, HEPA filters, and charcoal adsorbers. The HEPA filters remove particulates from the Control Room pressurizing air and prevent clogging of the iodine adsorbers. The charcoal adsorbers are installed to remove any radioiodines from the pressurizing air. The in-place test results should indicate a HEPA filter penetration of less than or equal to 0.3% through DOP testing and a charcoal adsorber penetration of less than or equal to 0.3% through halogenated hydrocarbon testing. The laboratory carbon sample results should indicate a radioactive methyl iodide penetration of less than or equal to 0.4% under test conditions similar to expected accident conditions. System flows should be near their design values. The ;

verification of these performance parameters combined with the qualification testing conducted on new filters and adsorbers provide a high level of assurance that the Emergency Filtration System will perform as predicted in reducing doses to plant personnel below those level stated in criterion 19 of Appendix A to 10 CFR 50.

Dose calculations have been performed for the Control Room Emergency Filtration System which show that, assuming 85% standby gas treatment system adsorption and filtration efficiency and 98% control room emergency filtration system adsorption and filtration efficiency and radioiodine plateout, whole body and organ doses remain within the guidelines of 10CFR50, Appendix A, General Design Criterion 19.

3.17 BASES 229y REV l

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4.17 Bases A. Control Room Ventilation System Control room air temperature is checked each shift to ensure that the continuous duty rating for the instrumentation and equipment cooled by this system is not exceeded.

Demonstrating automatic isolation of the control room using simulated accident signals assures control room isolation under accident conditions.

B. Control Room Emeroency Filtration System Air flow through the filters and charcoal adsorbers each month assures operability of the system.

The frequency of tests and sample analysis is necessary to show that the HEPA filters and charcoal adsorbers can perform as evaluated. The charcoal adsorber tray is installed which can accommodate a sufficient number of representative adsorber sample modules for estimating the amount of penetration the system adsorbs though its life. Sample modules will be installed with the same batch characteristics as the system adsorbent and will be withdrawn for the methyl iodide removal efficiency tests. Each module withdrawn will be replaced or blocked off. The laboratory methyl iodide test of the carbon adsorber is to be performed in accordance with ASTM D 3803-89, " Standard Test Method for Nuclear-Grade Activated Carbon. " In-place testing procedures will be established utilizing applicable sections of ASME N510-1989 as procedural guidelines only. The individual i test results obtained from in-place penetration testing for the HEPA filter upstream of the charcoal adsorber and of the HEPA filter downstream of the charcoal adsorber unit are to be multiplied together to determine the penetration of the combination of the two filters in series as a unit to satisfy the criteria of the specifications. If test results are unacceptable, all adsorbent in the train is replaced. Any HEPA filters found defective are replaced.

Pressure drop across the combined HEPA filters and charcoal adsorbers of less than or equal to 8 inches of l .

water at the system design flow rate will indicate that the filters and adsorbers are not clogged by excessive amounts of foreign matter. '

Demonstrating atitomatic control room pressurization using simu2 9tted accident signals assures control room ,

pressurization with .espect to adjacent areas under accident conditions.

l 4.17 BASES Page 229z REV

I Exhibit D Monticello Nuclear Generating Plant Revision One to License Amendment Reouest dated July 26.1996 MNGP MSLBA Evaluation Summary

, The radiological evaluation of the Main Steam Line Break Accident (MSLBA)is described in i USAR Section 14.7.3.

l Assumptions 1

\ 2 The postulated accident involves a guillotine break of one of the four main steam lines outside i of the containment, resulting in mass loss from both ends of the break. There is no fuel

damage as a consequence of this event, therefore the only activity released to the environment l

is that associated with the steam and liquid discharged from the break. Initially only steam will issue from the broken end of the steam line. Subsequently, rapid depressurization due to the break causes the reactor pressure vessel water level to rise, resulting in a steam-water mixture flowing from the break until the main steam isolation valves (MSIVs) are closed. For the MSIV

! closure time, an analysis input of 10.5 seconds after the MSLBA is used. Activity associated j with the discharged coolant is airborno in the turbine building instantaneously and released to the environment without delay.

l The analysis assumes that the accident occurs at hot standby conditions. At these conditions, steam generation from the decay heat in the core is very low and cannot make up the steam loss through the break. The results are a high rate of vessel depressurization and rapid rising of water level to the main steam line inlet. In addition to hot standby conditions, the 10CFR50 Appendix K break flow model was assumed in order to maximize the two-phase break flow rate, l Both of these assumptions yielded the maximum coolant mass releases through the break.

Two cases were analyzed. The first case assumes reactor pressure is at the safety relief valve opening setpoint of 1158 psia. The second case assumes the initial reactor pressure is at the l pressure regulator setpoint of 965 psia. The results show that the mass leaving the raactor l pressure vessel through the break is 71,574 lbm of liquid and 4,030 lbm of steam for the first case. In the cecond case the mass leaving the reactor pressure vessel through the break is 66,223 lbm of liquid and 4,243 lbm of steam.

l Accident parameters relevant to the radiological analysis are summarized in Table D-1. The l atrnospheric dispersion factors to the site boundary and to the control room intake, as well as ;

l control room parameters, are included in Table D-1.

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t Noble' Gas Concentration The assumed noble gas activity is the Monticello Technical Specification limit which corresponds to an off-gas release rate of 0.26 Cl/sec (rounded to 0.3 Ci/sec in the USAR) at 30 i

minutes delay. This activity is assumed to consist of a standard isotopic fraction.

lodine Concentration j i

The analysis used an input of 2 pCl/gm dose-equivalent of lodine-131 for the activity in the reactor coolant. A portion of the released coolant exists as steam prior to the accident.

Therefore, it is necessary to separate the initial steam mass from the total mass released and i assign a certain percentage of the fission product activity contained in this portion of the steam by an equivalent rnass of reactor coolant. A 2% carryover ratio was assumed for the analysis.

Offsite Dose and Control Room Dose Evaluations l 1 Activities released to the environment due to the MSLBA are calculated for both hot standby l l conditions. The case for reactor pressure at the safety relief valve opening setpoint and the I case for reactor pressure at the pressure regulator setpoint. In addition, the analysis was :

i performed for coolant concentrations based on both the TID-14844 and Regulatory Guide l l 1.109 thyroid dose conversion factors. I

! l Offsite dose consequences are presented in Table D-3.

Control room dose consequences are presented in Table D-4.

l l

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i D-2 1

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i TABLE D-1 '

Assumptions for MNGP MSLBA Anal'[ sis PARAMETER

_ VALUES Power Level Hot sta ldby at 4% power (66.8 MWt)

RPV Pressure (psia)

Case 1 1158 Case 2 965 Time Elapse for MSIV Full Closure (seconds) 10.5 i Fuel Rod Damage O Mass of Steam-Water Mixture Leaving Break (15m) See Table D-2 Reactor Coolant Dose Equivalent 1-131 (pCl/gm) 2 lodine Carryover Factor (%) 2 lodine Releases ( Cl/cm ) For Total Release See Table D-2 TID-14844 l-131 0.77 l-132 3.38 l-133 2.85 l-134 7.27 l l-135 2.56 Reg. Guide 1.109 ,

1-131 1.08 ;

l-132 4.72 l l-133 3.98 l l-134 10.2 l l-135 3.58 Thyroid Dose ConversHn Factors (rem /Ci) 1-131 1.08E+06 t 1-132 6.44E+03 1-133 1.80E+05 l-134 1.07E+03 1-135 3.13E+04 D-3

i i i

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i TABLE D-1 (continued) ,

Assumptions for MNGP MSLBA Analysis PARAMETER VALUES Data for Control Room Volume of Control Room (ft ) 27,000 Filter intake (cfm) 900 Efficiency of Charcoal adsorber (%) 98 Unfiltered inleakage (T < 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ) (cfm) 250 Unfiltered inleakage (T > 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ) (cfm) 10 Occupancy Factor 0 - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 1.0 1 - 4 days 0.6 4-30 days 0.4 Control Room Intake Atmospheric Dispersion Factors (sec/m*)

Ground Level Release 0 - 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 1.67E-03 8 -24 hours 1.41 E-03 1 - 4 days 9.65E-04 4 - 30 days 5.62E-04 J Offsite Atmospheric Dispersion Factor (sec/ m 3)

Ground Level Release 0 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months (EAB/LPZ) 9.20E-04/7.93E-05 i 2 - 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months (LPZ) 7.93E-05 8 -24 hours 5.35E-05 1 - 4 days 2.28E-05 4 - 30 days 6.68E-06 l

l l

I l l D-4 i

l l ,

i TABLE D-2 Mass Release From MSLBA (Ibm) CASE 1 CASE 2 Totalliquid released through break 71,574 66,223 l Liquid released flashing to steam 10,548 8,151 Initial steam prior to steam line covered 4,030 4,243 l

lodine Release From MSLBA (Cl)

TID-14844 - l l-131 28.8 26.2 1 1-132 126 114 l-133 106 96.3 l-134 271 246 l

l-135 95.6 86.6 Reg. Guide 1.109 l-131 40.2 36.4 l-132 176 160 1-133 148 135 l-134 379 343 1-135 134 121 l

l

4 D-5

I r ,i e I

i TABLE D-3 l

MNGP MSLBA Offsite Dose (REM) l 2 - Hour Exclusion Area Boundary 30 Day Low Population Zone TID-14844 DCF Thyroid Whole Body Thyroid Whole Body Case 1 17.3 0.28 1.49 0.02 Case 2 15.7 0.26 1.35 0.02 Reg. Guide DCF Case 1 24.2 0.40 2.08 0.03 Case 2 21.9 0.36 1.89 0.03 10CFR100 300 25 300 25 TABLE D-4 MNGP MSLBA Control Roo.n Dose (REM)

Thyroid Whole Body TID DCF Case 1 7.26 0.003 Case 2 6.58 0.002 Reg. Guide DCF

! Case 1 10.1 0.004 Case 2 9.18 0.003 GDC 19 30 5 i

l D-6

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Exhibit E MT3 cello Nuclear Generating Plant Revision One to License Amendment Reauest dated July 26.1996 MNGP RWCU Evaluation Summary l

l TABLE E-1 inputs for MNGP RWCU Evaluation PARAMETER VALUES Time Elapse for Operator Action from Break Initiation (Min.) 10 Time Elapse for RWCU Valve Closure (seconds) 29 l Fuel Rod Damage 0 l- Mass of Steam-Water Mixture Leaving Break (Ibm) 443,460 1

Data for Control Room Volume of Control Room (ft ) 27,000 Filter intake (cfm) 900 Efficiency of Charcoal adsorber (%) '98 Unfiltered inleakage (cfm) 250 Control Room Intake Atmospheric Dispersion Factors Ground Level Release (sec/m ) 1.67E-03 Offsite Atmospheric Dispersion Factor Ground Level Release (sec/ m )

Exclusion Area Boundary 9.20E-04 l

Low Population Zone) 7.93E-05 Reactor Coolant Dose Equivalent 1-131 ( Ci/gm) 0.25 lodine Releases 8

Isotope Concentration (pCi/cm ) Total Release (Ci) Thyroid Dose Conversion Factors (rem /Ci) 1-131 0.135 27.2 1.10E+06 l132 0.590 119 6.30E+03 1-133 0.497 100 1.80E+05 l-134 1.27 255 1.10E+03 i

1-135 0.447 89.9 3.10E+04 o

I i

e o a .%r e e a TABLE E-2 l MNGP RWCU Line Break Offsite Dose (REM) 2 - Hour Exclusion Area Boundary 30 Day Low Population Zone l Thyroid Whole Body Thyroid Whole Body 1 Dose (REM) 16.5 1.66 X 10" 1.42 1.43 X 10 I 10CFR100 300 25 300 25 Guideline TABLE E-3 MNGP RWCU Line Break Control Room Dose (REM) ,

Thyroid Whole Body l Dose (REM) 6.98 9.04 X 10-3 GDC 19 Guideline 30 5 l

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4 E-2

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