ML20154A370

Proposed Tech Specs Sections 3.3.2.1,3.4.6.2,3.4.8,3.6.2.1, 3.6.5.1,3.7.6.1 & 3.9.15,revising Info Re Revised MSLB Analyses & Revised Determinations of Radiological Consequences of MSLB & LOCA

ML20154A370
Person / Time
Site:	Millstone
Issue date:	09/28/1998
From:	NORTHEAST NUCLEAR ENERGY CO.
To:
Shared Package
ML20154A362	List: B17413, Application for Amend to License DPR-65,revising TS Sections 3.3.2.1,3.4.6.2,3.4.8,3.6.2.1,3.6.5.1,3.7.6.1 & 3.9.15 as Result of Revised MSLB Analyses & Revised Determinations of Radiological Consequences of MSLB & LOCA ML20154A370
References
NUDOCS 9810020287
Download: ML20154A370 (150)

Text

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Docket No. 50-336 817413 Millstone Nuclear Power Station, Unit No. 2 Proposed Revision to Technitsi Specifications

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Control Room Ventilation System Marked Up Pages i

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INSTRUMENTATION l

SURVEILLANCE REQUIREMENTS (Continued)

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4.3.2.1.3 The ENGINEERED SAFETY FEATURES RESPONSE TIME of*each ESF function 'shall be denonstrated to be within the limit at least.once per 18 months.

Each test shall include at least one channel per function ~

such that all channels are tested at.least once every N times 18. months l

where N is the total number of redundant channels in a specific ESF l

function as shown in the " Total.No. of Chanpels". Column of Table 3.3-3.

The trip value shall be'such that the containment purge l

4.3.2.1.4 effluent shall not result in calculated concentrations of radioactivity offsite in excess of 10 CFR Part 20, Appendix B. Table II.6 Forghe purposes of calculating this trip value, a x/Q = 5.8 x 10~ sec/m shall and a X/Q = 7.5 x 10-gm is a}igned to purge through the building vent be used.when the syst sec/m shall be used when the system is aligned l

to purge through the Unit 1 stack, the gaseous and-cprticulate (Half Lives greater than 8 days) radioactivity shall be-ered toEbe Xe-133 l

and Cs-137,5respectively. However, the setpoints shall be no greater than 5 x 10 cpm.

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MILLSTONE - UNIT 2 henJwd Alo. f t

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t REACTOR COOLANT SYSTEM 3

REACTOR COOLANT SYSTEN LEAKAGE LIMITING CONDITION FOR OPERATION

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l 3.4.6.2 Reactor Coolant System leakage shall be limitu to:

I No PRESSURE BOUNDARY LEAKAGE, a.

036 b.

1 GPM UNIDENTIFIED LEAKAGE, r--

-l-GPM-tete + primary-to-secondary leakage through bett, sic c.

-ienerater: : d 0.10 Om tl.re;;h any one steam generator, and l

d.

10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System.

APPLICABILITY: MODES 1, 2, 3 and 4.

l ACTION:

With any PRESSURE B0UNDARY LEAKAGE, be in COLD a.

SHUTDOWN within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

b.

With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE,

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l reduce the leakage rate to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in COLD SHUTDOWN within the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

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~ SURVEILLANCE REQUIRENENTS

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i 4.4.6.2 Reactor Coolant System-lc;k;ges shall be demonstrated to be within

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-- limits by performance of a Reactor Coolant System water inventory balance at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> during steady state operation except when operating in the shutdown cooling mode, 4

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MILLSTONE - UNIT 2 3/4 4-9 Amendment No. 75, 77, 77, D.

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INSERT A - Paae 3/4 4-9 I

l 4.4.6.2.2 Primary to secondary leakage shall be demonstrated to be within the above limits by performance of a primary to secondary leak rate determination at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The provisions of Specification 4.0.4 are not applicable for entry into MODE 4.

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REACTOR C0OLANT SYSTEM d6F' vari U,100 7 SPECIFIC ACTIVITY

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LINITING CONDITION FOR OPERATION 3.4.8 The specific activity of the primary coolant shall be limited to:

a.11.0 pC1/ gram DOSE EQUIVALENT I-131, and

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b. s 100/E pC1/gra%

d ss APPLICABILITY: MODES 1, 2, 3, 4, and 5.

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, ACTION:

MODES 1, 2 and 3*:

With the specific activity of the primary coolant > 1.0 pCi/ gram a.

DOSE EQUIVALENT I-131 but within the allowable limit (below and to the left of the line) shown on Figure 3.4-1, operation may continue for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

Specification 3.0.4 is not applicable.

l b.

With the specific activity of the primary coolant > 1.0 pti/ gram DOSE EQUIVALENT I-131 for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> during one continuous time interval or exceeding the limit line shown on Figure 3.41, be in HOT STANDBY with T,,, < 515'F within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

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With the specific activity of the primary coolant > 100/E pCi/g:ag, c.

be in HOT STANDBY with T < 515'F within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

MODES 1, 2, 3, 4 and 5:

of yost.

spu,rreodu,'/y d.

'With the specific activity of the primar coolant > 1.0 pCi/ gram DOSE EQUIVALENT I-131 or > 100/E pCi/gra, perform the sampling and I

analysis requirements of item 4 a) of Table 4.4-2 until the specific activity of the primary coolant is restored to within its limits.

• With T,,,2. 515'F.

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{Igt. STONE-UNIT 2 3/4 4-13 Amendment No. 7. JJJ. JJJ JJJ,

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February 3. 1987 p o CHANGE REACTOR COOLANT

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-1 QN L Y SURVEILLANCE REQUIREMENTS 4.4.8 The specific activity of the primary coolant shall. be detemined to be within the limits by performance of the sampling and analysis program of Table 4.4-2.

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TABLE 4.4-2 PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM ppff gp TYPE OF MEASUREMENT 19Ai,4 L YSIS AND ANALYSIS "IN IP.J".

. FREQUENCY 1.

Gross Activity Determination 3 times per 7 days with a maximum time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> between samples 2.

Isotopic Analysis for' DOSE 1 per 14 days EQUIVALENT I-131 Conce ration A<dlsis y

-y 30 Radiochemical for T Determination 1 per 6 months 4.

Isotopic Analysis for Iodine a)

Once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, Including I-131, I-133, and 1-135 whenever thD 90SE T.)

EQUIVALO'T I-13'r

-eveceds 1.0 wCi/gr =,

Spee4 a %)y.enerJs c

/C/>di/pm,bost b)

One sample between 2 69u.rv,4Lj:,or.2-/3/; o*

and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> following a THERMAL POWER change

/c ff (;/F** ?g exceeding 15 percent o

N of the RATED THERMAL 9/055 5,rccMi ae k,)y,o,,d POWER within a one j

hour period.

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90 days of 90 west 0.PEb1;-rn % elepsed smch fio ek was /<<s+ -ruberdu s t 6 yg hws or /enju.

TItc frc Nsn~t of Spee</suh.9 9,0 9 a,.7 i,of aplicalh.

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1 20 30 40 50 60 70 80 90 100 PERCENT OF RATED THERMAL POWER FIGURE 3.4-1 DOSE EQUIVALENT l-131 Primary Coolant Specific Activity Limit Versus Percent of RATED THERMAL POWER with the Primary Coolant Specific Activity > 1.0 Ci/ gram Dose Equivalent 1-131 MILLSTONE - UNIT 2 3/4 4-16 i

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CONTAINNENT SYSTEMS a., o, -

3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS CONTAINMENT SPRAY.VID COOLING SYSTEMS l

LIMITING CONDITION FOR OPERATION i

3.6.2.1 Two containment spray trains and two containment cooling trains, with each cooling train consisting of two containment air recirculation and cooling units, shall be OPERABLE.

APPLICABILITY: MODES 1, 2 and 3*.

s ACTION:

Inoperable Equipment Required Action

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a.

One containment a.1 Restore the inoperable containment spray spray train train to OPERABLE status within-;Ldays or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

b.

One containment b.1 Restore the inoperable containment cooling cooling train train to OPERABLE status within 7 days or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

c.

One containment c.1 Restore the inoperable containment spray spray train train or the inoperable containment cocling

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AND train to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or One containment be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

cooling train d.

Two containment d.1 Restore at least one inoperable containment cooling trains cooling train to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

e.

All other e.1 hter LC0 3.0.3 immediately.

combinations SURVEILLANCE REQUIREMENTS 4

4.6.2.1.1 Each containment spray train shall be demonstrated OPERABLE:

l At least once per 31 days on a STAGGERED TEST BASIS by:

a.

1.

Starting each spray pump from the control room, 2.

Verifying, that on recirculation flow, each spray pump develops a discharge pressure of 2 254 psig,

• The Containment Spray System is not required to be OPERABLE in MODE 3 if pressurizer pressure is < 1750 psia.

MILLSTONE - UNIT 2 3/4 6-12

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CONTAINMENT SYSTEMS

/Vo CNgwGE E h SURVEILLANCE REQUIRENENTS (Continued) 3.

Verifying that each spray pump operates for at least 15

minutes, I-4.

Cycling each testable, automatically operated valve-in each spray train flow path through at least one complete l

cycle, 5.

Verifying that upon a sump recirculation actuation signal the containment sump isolation valves open and that a recirculation mode flow path via an OPERABLE shutdown cooling heat exchanger is established, and i

6.

Verifying that all accessible manual valves not locked, sealed or otherwise secured in position and all remote or i

automatically operated valves in each spray train flow j

path _are positioned to take suction from the RWST on a Containment Pressure--High-High signal.

b.

At least once per 18 months, during shutdown, by cycling each l

L power operated valve in the spray train flow path not testable l

during plant operation through at least one complete cycle of full travel.

g'.n At least once per 18 months by verifying a total leak rate c.

less than or equal to 12 gallons per hour in conjunction with the high pressure safety injection system (reference Specifica-tion 4.5.2.c.5) at:

1)

Discharge. pressure of greater than or equal to 254 psig on recirculation flow for those parts of the system between the pump discharge and the header isolation valve, including the pump seals.

2)

Greater than or equal to 22 psig at.the pump suction for the piping from the containment sump check valve to the pump suction.

d.

At least once per 5 years by performing an air or smoke flow test through each spray header and verifying each spray nozzle is unobstructed.

4.6.2.1.2 Each containment air recirculation and cooling unit shall be l

demonstrated OPERABLE at least once per 31 days on a STAGGERED TEST BASIS by:

Starting, in low speed, each unit from the control room, a.

- b.

Verifying that each unit operates for at least 15 minutes, and

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v c.

Verifying a cooling water flow rate of 2 500 gpm to each cooling unit.

MILLSTONE - UNIT 2 3/4 6-13 Amendment No. 215 j.

Septembsr 30,1997 AJO C H o tt-CONTAINMENT SYSTEMS FoR Jwhtmtthe wcy 3/4.6.5 SECONDARY CONTAINMENT ENCLOSURE BUILDING FILTRATION SYSTg LINITING CONDITION FOR OPERATION 3.6.5.1 shall be OPERABLE.Two separate and independent Enclosure Building Filtration Trains l

APPLICABILITY: MODES 1, 2, 3 and 4.

ACIION:

With one Enclosure Building Filtration Train inoperable, restore the inoperable train to OPERABLE status within 7 days or be in COLD SHUTDOWN within the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

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SURVEILLANCE REQUIREMENTS 4.6.5.1 OPERABLE:_Each Enclosure Building Filtration Train shall be demonstrated l

1 I

At least once per 31 days on a STAGGERED TEST BASIS by initiat-a.

ing, from the control room, flow through the HEPA filter and charcoal absorber train and verifying that the train operates for at least 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> with the heaters on.

b.

At least once per 18 months or (1) after any structur,a1 maintenance on the HEPA filter or charcoal absorber housings, or 2) following painting, fire 'or chemical release in any ventilation (zone comuni-cating with the train by:

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$tS70NE - UNIT 2 3/4 6-25 Amandman2 006-208

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-Septembe 20,1997 CONTAINMENT SYSTDtS SURVEILLANCE REQUIREMENTS (Continued) 1.

Verifying that the cleanup train satisfies the in-place testing acceptance criteria and uses the test procedures of Regulatory l Positions C.5.a.

C.5.c and C.5.d of Regulatory Guide Revision 2. March 1978, and the train flow rate is 9000 cfm 1.52, I i 10%.

2.

Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52,* Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulato'ry Guide 1.52, Revision 2, March 1978.*

3.

Verifying a train flow rate of 9000 cfm i 10% during train l

operation when tested in accordance with ANSI H510-1975.

After every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation by verifying c.

withih 31 days after r6moval that a laboratory analysis of a representa-tive carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide'1.52, Revision 2, March 1978.*

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d.

At least once per 18 months by:

cL6

-1.

Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is 1 -+ nches Water Gauge while operating the train at a flow rate of 9000 cfm 10%.

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2.

' Verifying that the train starts on an Enclosure Building Filtra-l tion Actuation Signal (EBFAS).

l After each complete or partial replacement of a HEPA filter bank by I

e.

verifying that the HEPA filter banks remove greater than or equal to 99% of the DOP when they are tested in-place in accordance with ANSI N510-1975 while operating the train at a flow rate of 9000 cfm l

i 10%.

i i

ASTM D3803-89 shall be used in place of ANSI H509-1976 as referenced in table 2 of Regulatory Guide 1.52. The laboratory test of charcoal should be i

conducted at a temperature of 30*C and a relative humidity.of 95% within the g.m mm tolerances specified by ASTM D3803-89.

e; shall have a removal efficiency of ;t 95%. Additionally, the charcoal sample MILLSTONE - UNIT 2 3/4 3-26 Amendment No. 77, 77, 777, fh 0292 k

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September 30,1997 Ajo cH4mcr f10K INfoRanizen omty CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) f.

After each complete or partial replacement of a charcoal absorber bank by verifying that the charcoal absorbers remove greater than or equal to 99% of a halogenated hydrocarbon refrigerant test gas when they are tested in-place in accordance with ANSI N510-1975 while operating the train at a flow rate of 9000 cfm 10%.

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i MILLSTONE - UNIT 2 3/4 6-27 Amendment No. 208 0292

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3/4.7.6 CONTROL ROOM EMERGENCY VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION 3.7.6.1 Two independen /ontrol /oom /mer cy y/entilation tyttm shall be OPERABLE.

g gy APPLICABILITY: ALL MODES ACTION:

Modes 1, 2, 3, and 4: fras.v N' /* h'+ Na'+

With one Control Room Emergency Air Clean-up Sy: tem inoperable, restore the inoperable -sy;t:5 to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

MODES 5 and 6*:

tram l

f With one Control Room Emergency Air Clean Up Syster inoperable, a.

restore the inoperable yster to OPERABLE status within 7 days or W,,h /d" initiate and maintain operation of the remaining OPERABLE Control Room Emergency Air Clean-Up Systis in the recirculation mode.

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b.

With both control Room tmergency-Air Clean Up Sy t0'": inoperable, or i

with the OPERABLE Control Room Emergency Air Cle;a-Up System i required to be in the recirculation mode by ACTION (a. not capable of being powered by an OPERABLE normal and emergency po)wer source; I

suspend all operations involving CORE ALTERATIONS or positive reactivity changes.

o In Modes 5 and 6, when a Control Room Emergency Air Ch:n up :y:te$ is determined to be inoperable solely because its emergency power source is inoperable, or solely because its normal power source is inoperable, i.t may be considered OPERABLE for the purpose of satisfying the requirements of 3.7.6.1 Limiting Condition for Operation, provided:

(1) its.

corresponding normal.qt emergency power source is OPERABLE; and (2) all of its redundant system (s), subsystem (s), train (s), component (s) and device (s) are OPERABLE, or likewise satisfy the requirements of the specification. Unless both conditions (1) and (2) are satisfied within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, then Limiting condition for Operation (LCO) 3.7.6.1.a or 3.7.6.1.b shall be invoked as applicable.

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MILLSTONE - UNIT 2 3/4 7-16 AmendmentNo.72,JZS,h/

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4 22, M 94-PLANT SYSTEMS s:D)

SURVEILLANCE REQUIREMENTS s.g hh Each p'ontrol poom ptnergency pentilation -:f 4.7.6.1 y:t= shall be demonstrated OPERABLE:

At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying that the control room air a.

temperature is f 100'F.

b.

At least once per 31 days on a STAGGERED TEST BASIS by initiating l

from the control room, flow through the HEPA filters and charcoal absorber train and verifying that the +ystem operates for at least l

15 minutes.

At -least once per.18 months or (1) after any structural maintenance c.

on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical release in any ventilation zone communi-cating with the +ystem,by:

1.

Verifying that the cleanup 4ystem satisfies the in-place testing acceptance criteria and uses the test procedures of Regulatory Positions C.5.a C.S.c and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978, and the -:y:t : flow rate is 2500 cfm i 10%.

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2.

Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accor-dance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revi-sion 2, March 1978.*

The carbon sample shall have a removal efficiency of 2 95 pe nt.

3.

Verifying a cy:t:: flow rate of 2500 cfm 10% during -syster operation when tested in accordance with ANSI N510-1975.

d.

After every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal 'adsorber operation by verifying within 31 days after removal that a laboratory analysis of a repre-sentative carbon sample obtained in accordance with Regulatory Position. C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, l

meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978.*

i ASTM D3803-89 shall be used in place of ANSI H509-1976 as referenced in table 2 of Regulatory Guide 1.52. The laboratory test of charcoal should be conducted at a temperature of 30*C and a relative humidity of 95%

n..

within the tolerances specified by ASTM D3803-89.

[

MILLSTONE - UNIT 2 3/4 7-17 Amendment No. JJ 77, J JJF.

m life 199,

.. __-,.. mma,--

.m.s_mm.m m.c.. -,._,..,m.,.y.m,,_s-.,._*_.._

m PLANT SYSTEM _s E4 23' 1994 SURVEILLANCE REQUIRENENTS (Continued) 1 At least once per 18 months by:

'I e.

1.

Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than+1nches Water y

Gauge while operating the -eysta-at a flow rate of 2500 cfm i 10%.

2.

Verifying that on a recirculation signal, the gstem automaJ-cally switches into a recirculation mode of operation with f1r through the.HEPA filters and charcoal adsorber banks.

tooth & Gsk /

l%~ Emr/yescy Venblah.~ m in open hng w n nu/ma / snode am/

ill r 5 mobt puff (

Mode, j

p Q:.

4 N

MILLSTONE - UNIT 2 3/47-17a Amendment No. JJ, 77 Jpj3 JJ7, em 111,199. HP

.~

e

• tober 29, 1990' PLANT SYSTEMS

$;p:

SURVEI! LANCE REQUIREMENT (Continued)

(nho( @u Ewp<y UcnfkJi,a

_ a-4 3.

Verifying that control room air in-leakage is less than 4

/30 4 406 SCFM with the #^tt the recirculation /flitration mode.1 Air C0ndioi:ning System operating in i

i f.

verifying that the HEPA filter banks remove greater i

4 99% of the DOP when they are tested in-place in accordance with ANSI N510-1975 while operating the <ytta-at a flow rate of 2500 cfm i 10%.

After each complete or partial replacement of a charcoal adsorber g.

i bank by verifying that the charcoal adsorbers remove greater than or equal to 99% of a halogenated hydrocarbon refrigerant test gas when they are tested in-place in accordance with ANSI H510-1975 while operating the 4y:t:a at a flow rate of 2500 cfm i 10%.

l i

1 I

I 4

i i

s 1

MILLSTONE - UNIT 2 3/4 7-18 AmendmentNo.77,166,((

~.. _ _ _

Saptember 30,1997 Ajo CHaeGE REFUELING OPERATIONS N*

STORAGE POOL AREA VENTILATION SYSTEM - FUEL STORAGE Cwty LIMITING CONDITION FOR OPERATION 3.9.15 At least one OPERABLE and capable o. Enclosure Building Filtration Train shall be tary exhaust mode and exhausting through HEPA filters and charcoalf.

l adsorbers on a storage pool area high radiation signal.

APPLICABILITY: WHENEVER IRRADIATED FUEL IS IN THE STORAGE POOL.

ACTION:

With the requirements of the above specification not satisfied, suspend all operations involving movement of fuel within the storage pool or crane Building Filtration Train is restored to OPERABLE status.cperation SURVEILLANCE REQUIREMENTS 4.9.15 be demonstrated OPERABLE:The above required Enclosure Building Filtration Train s l

At least once per 31 days on a STAGGERED TEST BASIS by initiating, a.

from the control room, flow through the HEPA filters and charcoal i

adsorbers and verifying that the train operates for at.least 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> with the heaters on.

l b.

At least onc.e per 18 months or (I) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical release in any ventilation zone communi-cating with the train by:

l MILLSTONE - UNIT 2 l

0213 3/4 9-16 Amendment No. 77, 208

. __ __y_

~

mmn,r-n n m na m m m Cspuu.bei CCA DDP REFUELING OPERATIONS SURVEILLANCE REQUIREMENTS (Continued) 1.

Verifying that the cleanup train satisfies the in-place testing acceptance criteria and uses the test procedures of Regulatory l

Positions C.S.a C.S.c and C.S.d of Regulatory Guide 1.52, Revi-sion 2, March' 1978, and the train flow rate is 9000 cfm i 10%.

l 2.

Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.h of Regulatory Guide l.' 52, Revision 2, 1

March ~1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978.'*

3.

Verifying a train flow rate of 9000 cfm when tested in accordance with ANSI N510-1975.10%duringtrainoperationl After every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation by verifying c.

within 31 days after removal that a laboratory analysis of a repre-sentative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory

~

Guide 1.52, Revision 2, March 1978.*

d.

At least once per 18 months by:

M 1.

Verifying 'that the pressure drop across the combined HEPA filters and charcoal adsorber banks is sVr inches Water Gauge while operating the train at a flow rat,e of 9000 cfm

.10%.

l 2.

Verifying that on a Spent Fuel Storage Pool Area high radiation signal, the train automatically starts (unless already operating and directs its exhaust. flow through the HEPA filters and charcoa)l l

adsorber banks.

After each complete or partial replacement of a HEPA filter bank by e.

verifying that the HEPA filter banks remove greater than or equal to 99% of the DOP when they are tested in-place in accordance with ANSI H510-1975 while operating the train at a flow rate.of 9000 cfm 10%.

l ASTM D3803-89 shall be used in place of ANSI H509-1976 as referenced in table 2 of Regulatory Guide 1.52.

be conducted at a tem The laboratory test of charcoal should 4

within the tolerances,perature of 30*C and a r61ative humidity of 95%

specified by ASTM D3803-89.

charcoal sample shall have a removal efficiency of ;t 95%. Additionally, the MILLSTONE - UNIT 2 0293 3/4 9-17 Amendment No. 77 M. ///

.,,,.n._.- _;=- --

September 30,1997

/v<> c h%NGE REhJELING OPERATIONS SA TA4%BP A rzw (h

outy L

SURVEILLANCE REQUIREMENTS (Continued) f.

After each complete or partial replacement of a charcoal adsorber nank by verifying that the charcoal adsorbers remove greater than when they are tested in place in accordance with ANSI N while operating th'e train at a flow rate of 9000 cfm i 10%.

l f.

MILLSTONE - UNIT 2 0293 3/4 9-18 Amendment No. ?nn

_ _ m.m.,m A.. m y :. 100; REACTOR COOLANT SYSTEM

^

BASES

}

3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.6.1 LEAKAGE DETECTION SYSTEMS The RCS leakage detection systems required by this specification are provided to monitor and detect leakage from the Reactor Coolant Pressure Boundary. These detection systems are consistent with the recomendations of Regulator Systems."y Guide 1.45,

• Reactor Coolant Pressure Boundary Leakage Detection 3/4.4.6.2 REACTOR COOLANT SYSTEM LEAKAGE Industry experience has shown that while a limited amount of leakage is i

expected from the RCS, the unidentified portion of this leakage can be reduced to a threshold value of less than 1 GPH.

low to ensure early detection of additional leakage.This threshold value is sufficie j

The 10 GPM IDENTIFIED LEAXAGE limitation provides allowance for a limited amount of leakage from known sources whose presence will not interfere with i

the detection of UNIDENTIFIED LEAXAGE by the leakage detection systems.

g

]-

The -tetil steam generator tube leakage limit of 4(GPM fcr cil steam 47 4 generator #' ensures that the dosage contribution from the tube leakage will be

/gimit;d tc : : :11 fracti:r, cf P;rt 100 limiti; in the event of either a steam c

enerator tube rupture or steam line break.

The +7GPM limit is consistent

/

fA )with the assumptions used in the analysis of these ace'ident A

/r55

/Mh [be PRESSURE BOUNDARY LEAXAGE of any magnitude is unacceptable since it may

-hw indicative of an impending gross failure of the pressure boundary.

of 6 red Therefore, the presence of any PRESSURE BOUNDARY LEAXAGE requires the unit to Des y be promptly placed in COLD SHUTDOWN.

I The

.10 GPM Prima to Secondary takage limitati assures str ural I integr y.

A tube w a through-w circumferenti crack which

/oW D D.10 under no 1 operating nditions retal the structu enks at l margins (Spd,, g ) oc mended in R latory Guide J.121.

In addit n, the total eakage under

,,./

c'cident condi ns would remaWbelow the 1 GPM limit.

MILLSTONE - UNIT 2 B3/44-3 Amendment Nos. 12!,

~~

_u, o oc inno _

CnNTAINNENT SYSTEMS

" ~

BASES 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS 3/4.6.2.1 CONTAINMENT SPRAY AND COOLING SYSTEMS

[

The OPERABILITY of the containment spray system ensures that contain-ment depressurization and cooling capability will be available in the event of a LOCA. The pressure reduction and resultant lower containment leakage rate are consistent with the assumptions used in the accident analyses. The leak rate surveillance requirements assure that the leakage assumed for the system outside containment during the recircula-tion phase will not be exceeded.

l The OPERABILITY of the containment cooling system ensures that

1) the containment air temperature will be maintained within limits during normal operation, and 2) adequate heat removal capacity is available when operated in conjunction with the containment spray system during post-l LOCA conditions.

To be OPERABLE, the two trains of the containment spray system shall be capable of taking a suction from the refueling water storage tank on a containment spray actuation signal and automatically transferring suction to O

the containment sump on a sump recirculation actuation signal. Each j

containment spray train flow path from the containment sump shall be via an OPERABLE shutdown cooling heat exchanger.

The containment cooling system consists of two containment cooling trains.

Each containment cooling train has two containment air recirculation and cooling units.

For the purpose of applying the appropriate action statement, the loss of a single containment air recirculation and cooling unit will make the respective containment cooling train inoperable.

Either the containment spray system or the containment cooling system has sufficient heat removal capability to handle any design basis accident.

However, the containment spray system is more effective in dealing with the superheated steam from a main steam break inside containment. Therefu.c, et Mast-GGG tr&in GE C56thiinu6nt 5ps ay

's always scyuircu lu istr-OPERAntE, wnen

~

-pressur-4:cr pressure ;2 2 USGW.

j MERT 3/4.6.3 CONTAINMENT ISOLATION VALVES The Technical Requirements Manual contains the list of containment isolation valves (except the containment air lock and equipment hatch). Any changes to this list will be reviewed under 10CFR50.59 and approved by the Plant Operations Review Committee (PORC).

The OPERABILITY of the containment isolation valves ensures that the

..s

")'

containment atmosphere will be isolated from the outside environment in the event of a release of radioactive material to the containment atmos-phere or pressurization of the containment.

Containment isolation within MILLSTONE - UNIT 2 B 3/4 6-3 Amendment No.

77,

& pf, EJp, l

INSERT B - Paoe B 3/4 6-3 In addition, the containment spray system provides a mechanism for removing iodine from the containment atmosphere. Therefore, at least one train of containment spray is required to be OPERABLE when pressurizer pressure is > 1750 psia, and the allowed outage time for one train of containment spray reflects the dual function of containment spray for heat removal and iodine removal.

~.___.

~

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

f

-Aus":t 1, 1975 p_lANT SYSTEMS BASES 7-z

_3/4.7.4 SERVICE WATER SYSTEM i

.The OPERABILITY of the service water system ensures that suffici cooling capacity is available for continued operation of vital compone and Engineered Safety Feature equipment during normal and accident c 1

ditions.

The redundant cooling capacity of this system single failure, is consistent with the assumptions used in the accident

~

, assuming a analyses.

I

_3/4.7.5 FLOOD LEVEL to an elevation of 22 feet.The service water pump motors are no If the water level is exceeding plant grade level or if a severe storm is approaching the plant site, one service wate motor will be protected against flooding to a mini t

the reactor.

f action to provide pump motor protection will be initiat

\\

reaches plant grade level.

3/4.7.6 CONTROL ROOM EMERGENCY VENTILATION SYSTEM The OPERABIL.'TY of the /o 01 /

p!me ency tilation stem ensures that 1) the ambient air temperature does not exceed the allowaM temperature for continuous duty rating for the equipment and instru cooled by this system and 2) the control room will remain habitable for mentation operations personnel during and following all credible accident conditions The OPERABILITY of this system in conjunction with control room provisions is based on limiting the radiation exposure to personnel This limitation is consistent with the requirements of G Criteria of Appendix "A",10 CFR 50.

gn 19 g x TMUT C MILLSTONE - UNIT 2 B 3/4 7-4

4 INSERT C - Paae B 3/4 7-4 i

The control room radiological dose calculations use the conservative minimum acceptable flow of 2250 cfm based on the flow rate surveillance requirement of 2500 cfm i 10%.

Currently there are some situations where the CREV System may not automatically start on an accident signal, without operator action.

Under most situations, the emergency filtration fans will start and the CREV System will be in the accident lineup. However, a failure of a supply fan (F21A or B) or an exhaust fan (F31 A or B), operator action will be required to return to a full train line up. Also, if a j

single emergency bus does not power up for one train of the CREV System, the opposite train filter fan will automatically start, but the required supply and exhaust fans 7

will not automatically start. Therefore, operator action is required to establish the whole train line up. This action is specified in the Emergency Operating Procedures.

The radiological dose calculations do not take credit for CREV System cleanup action until 10 minutes into the accident to allow for operator action.

{

When the CREV System is checked to shift to the recirculation mode of operation, this will be performed from the normal mode of operation, and from the j

smoke purge mode of operation.

l

Docket No. 50-7%

B17413 1

Millstone Nuclear Power Station, Unit No. 2 Proposed Revision to Technical Specifications Control Room Ventilation System Retyped Pages September 1998

INSTRUMENTATION SURVEILLANCE REQUIRENENTS (Continued) 4.3.2.1.3 The ENGINEERED SAFETY FEATURES RESPONSE TIME of each ESF function shall be demonstrated to be within the limit at least once per 18 months.

Each test shall include at least one channel per function such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific ESF function as shown in the " Total No. of Channels" Column of Table 3.3-3.

4.3.2.1.4 The trip value shall be such that the containment purge effluent shall not result in calculated ' concentrations of radioactivity offsite in excess of 10 CFR Part 20, Appendix B, Table II.6For thg hall purposes of calculating this trip value, a x/Q = 5.8 x 10- sec/m s beusedwhenthesystgmisa}ignedtopurgethroughthebuildingvent and a X/Q = 7.5 x 10-o sec/m shall be used when the system is aligned to purge through the Unit 1 stack, the gaseous and particulate (Half Lives greater than 8 days) radioactivity shall be assumed to be Xe-133 and Cs-137 respectively. However, the setpoints shall be no greater than 5 x10g cpm.

i MILLSTONE - UNIT 2 3/4 3-11 Amendment No. JP 0409

REACTOR COOLANT SYSTEN.

REACTOR COOLANT SYSTEN LEAKAGE LINITING CONDITION FOR OPERATION l

3.4.6.2 Reactor Coolant System. leakage shall be limited to:

a.

No PRESSURE BOUNDARY LEAKAGE, b.

1 GPM UNIDENTIFIED LEAKAGE, c.

0.035 GPM primary-to-secondary leakage through any one steam generator, and d.

10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System.

APPLICABILITY: MODES 1, 2, 3 and 4.

f ACTION:

a.

With any PRESSURE B0UNDARY LEAKAGE, be in COLD SHUTDOWN within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

b.

With' any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE, reduce the-leakage rate to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in COLD SHUTDOWN within the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

SURVEILLANCE REQUIRENENTS 4.4.6.2.1 Reactor Coolant System IDENTIFIED LEAKAGE and UNIDENTIFIED LEAKAGE shall be demonstrated to be within limits by performance of a Reactor Coolant System water inventory balance at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> during steady state operation except when operating in the shutdown cooling mode.

4.4.6.2.2 Primary to secondary leakage shall be demonstrated to be within the above limits by performance of a primary to secondary leak rate determination at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The provisions of Specification 4.0.4 are not applicable for entry into MODE 4.

MILLSTONE - UNIT 2 3/4 4-g Amendment No. 17,77,71,p),

uw 191,111,1H,119,

REACTOR COOLANT SYSTEM l

SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION 3.4.8 The specific activity of the primary coolant shall b'e limited to:

a.

1 1.0 pCi/ gram DOSE EQUIVALENT I-131, and b.

s 100/E Ci/ gram of gross specific activity.

l APPLICABILITY: MODES 1, 2, 3, 4, and 5.

ACTION:

MODES 1, 2, and 3*:

With the specific activity of the primary coolant > 1.0 pCi/ gram a.

DOSE EQUIVALENT I-131 but within the allowable limit (below and to the left of the line) shown on Figure 3.4-1, operation may continue for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

Specification 3.0.4 is not applicable.

b..

With the specific activity of the primary coolant > 1.0 pCi/ gram DOSE EQUIVALENT I-131 for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> during one continuous time interval or exceeding the limit line shown on Figure 3.4-1, be 1

in HOT STANDBY with T,,, < 515'F within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

I With the specific activity of the primary coolant > 100/E pci/ gram c.

of gross specific activity, be in H0T STANDBY with T,,, < 515*F within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

MODES 1, 2, 3, 4 and 5:

d.

With the specific activity of the primary coolant > 1.0 Ci/ gram DOSE EQUIVALENT I-131 or > 100/E C1/ gram of gross specific activity, perform the sampling and analysis requirements of item 4 a) of Table 4.4-2 until the specific activity of the primary coolant is restored to within its limits.

• With T,,,2 515'F.

MILLSTONE - UNIT 2 3/4 4-13 Amendment No. 7. JJJ JJJ, JJJ, o<11

Ilf,

I TABLE 4.4-2 PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGEM TYPE OF MEASUREMENT SAMPLE AND ANALYSIS AND ANALYSIS FREQUENCY 1.

Gross Activity Determination 3 times per 7 days with a i

maximum time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> between samples 2.

Isotopic Analysis for DOSE 1 per 14 days EQUIVALENT I-131 Concentration 1

3.

Radiochemical. Analysis for 1 per 6 months

• l E Determination 4.

Isotopic Analysis for Iodine a) Once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, Including I-131, I-133, and I-135.

whenever the specific

' activity exceeds 1.0

1. Ci/ gram, DOSE EQUIVALENT I-131, or 100/E pci/ gram of gross specific activity, and b) One sample between 2 and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> following a THERMAL POWER change exceeding 15 percent of the RATED THERMAL POWER within a one hour period.

-

• Sample to be taken after a minimum of 2 EFPD and 20 days of POWER OPERATION have elapsed since reactor was last subcritical for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or longer. The provisions of Specification 4.0.4 are nnt applicable.

MILLSTONE - UNIT 2 3/4 4-15 Amendment No.

0412

CONTAINMENT SYSTEMS 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTENS (ONTAINMENT SPRAY AND COOLING SYSTEMS LIMITING CONDITION FOR OPERATION 3.6.2.1 Two containment spray trains and two containment cooling trains, with each cooling trair consisting of two containment air recirculation and cooling units, shall be OPERABLE.

APPLICABILUl: MODES 1, 2 and 3*.

ACTION:

Inoperable Equipment Required Action a.

One containment a.1 Restore the inoperable containment spray spray train train to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or l

1 be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

b.

One containment b.1 Restore the inoperable containment cooling cooling train train to OPERABLE status within 7 days or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

c.

One containment c.1 Restore the inoperable containment spray spray train train or the inoperable containment cooling AND train to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or One containment be in H0T SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

cooling train d.

Two containment d.1 Restore at least one inoperable containment cooling trains cooling train to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

e.

All other e.1 Enter LC0 3.0.3 immediately.

combinations SURVEILLANCE REQUIREMENTS 4.6.2.1.1 Each containment spray train shall be demonstrated OPERABLE:

a.

At least once per 31 days on a STAGGERED TEST BASIS by:

1.

Starting each spray pump from the control room, 2.

Verifying, that on recirculation flow, each spray pump develops a discharge pressure of 1 254 psig,

• The Containment Spray System is not required to be OPERABLE in MODE 3 if pressurizer pressure is < 1750 psia.

MILLSTONE - UNIT 2 3/4 6-12 Ainendment No. 7#

0413

CONTAINMENT SYSTEMS l

SURVEILLANCE REQUIREMENTS (Continued) l l

1.

Verifying that the cleanup train satisfies the in-place testing i

acceptance criteria and uses the test procedures of Regulatory l

Positions C.S.a, C.S.c and C.S.d of Regulatory Guide 1.52, Revision 2, March 1978, and the train flow rate is 9000 cfm i 1M.

2.

Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with.

Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, mr '.s the laboratory testing criteria of Regulatory Position r ". Of Regulatory Guide 1.52, Revision 2, March 1978.*

3.

Verifying a train flow rate of 9000 cfm 10% during train operation when tested in accordance with ANSI N510-1975.

After every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation by verifying c.

within 31 days after removal that a laboratory analysis of a representa-tive carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory i

Guioe 1.52, Revision 2, March 1978

• d.

At least once per 18 months by; 1.

Verifying that the pressure drop across the combind HEPA filters and charcoal adsorber banks is s 2.6 inches Water Gauge while l

1 operating the train at a flow rate of 9000 cfm 10%.

J 2.

Verifying that the train starts on an Enclosure Building Filtra-l tion Actuation Signal (EBFAS).

e.

After each complete or partial replacement of a HEPA filter bank by l

verifying that the HEPA filter banks remove greater than or equal to l

99% of the D0P when they are tested in-place in accordance with ANSI N510-1975 while operating the train at a flow rate of 9000 cfm i 10%.

l l

ASTM D3803-89 shall be used in place of ANSI N509-1976 as referenced in table 2 of Regulatory Guide 1.52.

The laboratory test of charcoal should be conducted at a temperature of 30*C and a relative humidity of 95% within the tolerances specified by ASTM D3803-89.

Additionally, the charcoal sample j

shall have a removal efficiency of ;t 95%.

MILLSTONE - UNIT 2 3/4 6-26 Amendment No. 75, 77, J7), 197, 0414

PLANT SYSTEMS 3/4.7.6 CONTROL ROOM EMERGENCY VENTILATION SYSTEM LINITING CONDITION FOR OPERATION 3.7.6.1 Two inder.endent Control Room Emergency Ventilation Trains shall be l

OPERABLE.

APPLICABILITY: ALL MODES ACTION:

Modes 1, 2, 3, and 4:

With one Control Room Emergency Ventilation Train inoperable, restore the inoperable train to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

MODES 5 and 6*

1 a.

With one Control Room Emergency Ventilation Tra.in inoperable, restore the inoperable train to OPERABLE status within 7 days or initiate and maintain operation of the remaining OPERABLE Control Room Emergency Ventilation Train in the recirculation mode.

[

b.

With both Control Room Emergency Ventilation Trains inoperable, or with the OPERABLE Control Room Emergency Ventilation Train required to be in the recirculation mode by ACTION (a.) not capable of being powered by an OPERABLE normal and emergency power source, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.

In Modes 5 and 6, when a Control Room Emergency Ventilation Train is I

determined to be inoperable solely because its emergency power source is inoperable, or solely because its normal power source is inoperable, it may be considered OPERABLE for the purpose of satisfying the requirements of 3.7.6.1 Limiting Condition for Operation, provided: (1) its corresponding normal.o_t emergency power source is OPERABLE; and (2) all of its redundan.t system (s), subsystem (s), train (s), component (s) and device (s) are OPERABLE, or likewise satisfy the requirements of the specification. Unless both conditions (1) and (2) are satisfied within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, then Limiting Condition for Operation (LCO) 3.7.6.1,a or 3.7.6.1.b shall be invoked as applicable.

MILLSTONE - UNIT 2 3/4 7-16 Amendment No. 7E, M, M,

0415

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS 4.7.6.1 Each Control Room Emergency Ventilation Train shall be demonstrated l OPERABLE:

a.

At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying that the control room air temperature is 1 100*F.

b.

At least once per 31 days on a STAGGERED TEST BASIS by initiating from the control room, flow through the HEPA filters and charcoal absorber train and verifying that the train operates for at leastl 15 minutes.

c.

At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical release in any ventilation zone communi-cating with the train by:

l 1.

Verifying that the cleanup train satisfies the in-place l testing acceptance criteria and uses the test procedures of Regulatory Positions C.5.a, C.S.c and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978, and the train flow rate is 2500l cfm 10%.

2.

Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accor-dance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52 Revi-3 sion 2, March 1978.*

The carbon sample shall have a removal efficiency of 2 95 percent.

3.

Verifying a train flow rate of 2500 cfm 10% during train l operation when tested in accordance with ANSI N510-1975.

d.

After every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation by verifying within 31 days after removal that a laboratory analysis of a repre-sentative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978.*

• ASTM D3803-89 shall be used in place of ANSI N509-1976 as referenced in table 2 of Regulatory Guide 1.52.

The laboratory test of charcoal should be conducted at a temperature of 30*C and a relative humidity of 95% within the tolerances specified by ASTM D3803-89.

MILLSTONE - UNIT 2 3/4 7-17 Amendment No. 77, 77, Jpp, JJ7, 0415 175. If7 175.

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS e.

At least once per 18 months by:

l 1.

Verifying that the pressure drop across the combined HEPA l

filters and charcoal 'adsorber banks is less than 3.4 inches Water l

Gauge while operating the train at a flow rate of 2500 cfm l

10%.

2.

Verifying that on a recirculation signal, with the Control Room Emergency Ventilation Train operating in the normal mode and the l

smoke purge mode, the train automatically switches into a l

recirculation mode of operation with flow through the HEPA l

filters and charcoal adsorber banks.

l l

1 l

l l

l l

MILLSTONE - UNIT 2 3/4 7-17a Amendment ho. 75,17,199,JJ7, on's 175,Iff,179, l

l

i PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 3.

Verifying that control room air in-leakage is less than 130 SCFM with the Control Room Emergency Ventilation System operating in the recirculation / filtration mode.

f.

After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter banks remove greater than or equal to 99% of the D0P when they are tested in-place in accordance with ANSI l N510-1975 while operating the train at a flow rate of 2500 cfm 10%.

g.

After each complete or partial replacement of a charcoal absorber bank by verifying that the charcoal absorbers remove greater than or equal to 99% of a halogenated hydrocarbon refrigerant test gas when they are tested in-place in accordance with ANSI N510-1975 while l operating the train at a flow rate of 2500 cfm 10%.

MILLSTONE - UNIT 2 3/4 7-18 Amendment No. 77, 199, J #,

REFUELING OPERATIONS SURVEILLANCE REQUIREMENTS (Continued) 1.

Verifying that the cleanup train satisfies the in-place testing acceptance criteria and uses the test procedures of Regulatory Positions C.S.a, C.5.c and C.5.d of Regulatory Guide 1.52, Revi-sion 2, March 1978, and the train flow rate is 9000 cfm i 10%.

2.

Veriiying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.G.a of Regulatory Guide 1.52, Revision 2, March 1978.*

1 3.

Verifying a train flow rate of 9000 cfm 10% during train operation when tested in accordance with ANSI N510-1975.

c.

After every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation by verifying within 31 days after removal that a laboratory analysis of a repre-sentative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978.*

d.

At least once per 18 months by:

1.

Verifying that the pressure drop across the combined HEPA fi'ters and charcoal adsorber banks is s 2.6 inches Water Gauge l

while operating the train at a flow rate of 9000 cfm i 10%.

2.

Verifying that on a Spent Fuel Storage Pool Area high radiation signal, the train automatically starts (unless already operating) 1 and directs its exhaust flow through the HEPA filters and charcoal adsorber banks, e.

After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter banks remove greater than or equal to 99% of the D0P when they are tested in-place in accordance with ANSI N510-1975 while operating the train at a flow rate of 9000 cfm 10%.

ASTM D3803-89 'shall be used in place of ANSI N509-1976 as referenced in table 2 of Regulatory Guide 1.52.

The laboratory test of charcoal should be conducted at a temperature of 30*C and a relative humidity of 95% within the tolerances specified by ASTM D3803-89.

Additionally, the charcoal sample shall have a removal efficiency of 2 95%.

MILLSTONE - UNIT 2 3/4 9-17 Amendment No. 77, J75, 797.

0416

REACTOR COOLANT SYSTEM BASES 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.6.1 LEAKAGE DETECTION SYSTEMS j

The RCS leakage detection systems required by this specification are provided to monitor and detect leakage from the Reactor Coolant Pressure Boundary.

These detection systems are consistent with the recommendations of Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems."

3/4.4.6.2 REACTOR COOLANT SYSTEM LEAKAGE Industry experience has shown that while a limited amount of leakage is i

expected from the RCS, the unidentified portion of this leakage can be reduced to a threshold value of less than 1 GPM. This threshold value is sufficiently low to ensure early detection of additional leakage.

The 10 GPM IDENTIFIED LEAKAGE limitation provides allewance for a limited i

amount of leakage from known sources whose presence will

)t interfere with the detection of UNIDENTIFIED LEAKAGE by the leakage de'. tion systems.

The steam generator tube leakage limit of 0.035 c'M per steam generator ensures that the dosage contribution from the tube leakage will be less than the limits of General Design Criteria 19 of 10CFR50 Appendix A in the event of either a steam generator tube rupture or steam line break.

The 0.035 GPM limit is consistent with the assumptions used in the analysis of these accidents.

PRESSURE B0UNDARY LEAKAGE of any magnitude is unacceptable since it may be indicative of an impending gross failure of the pressure boundary.

Therefore, the presence of any PRESSUrd B0UNDARY LEAKAGE requires the unit to be promptly placed in COLD SHUTDOWN.

MILLSTONE - UNIT 2 B 3/4 4-3 Amendment Nos. J77, J77,

CONTAINMENT SYSTEMS BASES 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS 3/4.6.2.1 CONTAINMENT SPRAY AND COOLING SYSTEMS The OPERABILITY of the containment spray system ensures that contain-ment depressurization and cooling capability will be available in the event of a LOCA.

The pressure reduction and resultant lower containment leakage rate are consistent with the assumptions used in the accident analyses. The leak rate surveillance requirements assure that the leakage assumed for the system outside containment during the recircula-tion phase will not be exceeded.

The OPERABILITY of the containment cooling system ensures that 1) the containment air temperature will.be maintained within limits during normal operation, and 2) adequate heat removal capacity is available when j

operated in conjunction with the containment sproy system during post-LOCA conditions.

To be OPERABLE, the two trains of the containment spray system shall be capable of taking a suction from the refueling water storage tank on a containment spray actuation signal and automatically transferring suction to the containment sump on a sump recirculation actuation si.gnal.

Each-containment spray train flow path from the containment sump shall be via an OPERABLE shutdown cooling heat exchanger.

The containrrant cooling system consists of two containment cooling trains.

Each containment cooling train has two containment air recirculation and cooling units.

For the purpose of applying the appropriate action statement, the loss of a single containment air recirculation and cooling unit will make the respective containment cooling train inoperable.

Either the containment spray system or the containment cooling system has sufficient heat removal capability to handle any design basis accident.

However, the containment spray system is more effective in dealing with the superheated steam from a main steam break inside containment.

In addition, the containment spray system provides a mechanism for removing iodine from the containment atmosphere.

Therefore, at least one train of containment spray is required to be OPERABLE when pressurizer pressure is 2 1750 psia, and the allowed outage time for one train of containme'nt spray reflects the dual function of containment spray for heat removal and iodine removal.

3/4.6.3 CONTAINMENT ISOLATION VALVES The Technical Requirements Manual contains the list of containment isolation valves (except the containment air lock and equipment hatch). Any changes to this list will be reviewed under 10CFR50.59 and approved by the Plant Operations Review Committee (PORC).

The OPERABILITY of the containment isolation valves ensures that the containment atmosphere will be isolated from the outside environment in the event of a release of radioactive material to the containment atmos-phere or pressurization of the containment. Containment isolation within MILLSTONE - UNIT 2 B 3/4 6-3 Amendment No. 7), JJ, //p, /JJ

_ _ _ _ _. _ _. ~. -. _. _ _. _ _ _ _. _ _ _ _ _. _.... _ _ =.

PLANT SYSTEMS RARFC 3/4.7.4 SERVICE WATER SYSTEM The OPERABILITY of the service water system ensures that sufficient cooling capacity is available for continued operation of vital components and Engineered Safety Feature equipment during normal and accident con-ditions. The redundant cooling capacity of this system, assuming a single failure, is consistent with the assumptions used in the accident analyses.

3/4.7.5 FLOOD LEVEL The service water pump motors are normally protected against water damage to an elevation of 22 feet.

If the water level is exceeding plant grade level or if a severe storm is approaching the plant site, one service water pump motor will be protected against flooding to a minimum elevation of 28 feet to ensure that this pump will continue to be capable of removing decay heat from the reactor.

In order to ensure operator accessibility to the intake structure action to provide pump motor protection will be initiated when the water level reaches plant grade level.

3/4.7.6 CONTROL ROOM EMERGENCY VENTILATION SYSTEM The OPERABILITY of the Control Room Emergency Ventilation System l

ensures that 1) the ambient air temperature does not exceed the allowable temperature for continuous duty rating for the equipment and instrumentation cooled by this system and 2) the control room will remain habitable for operations personnel during and following all credible accident conditions.

The OPERABILITY of this system in conjunction with control room design provisions is based on limiting the radiation exposure to personnel occupying the control room to 5 rem or less whole body, or its equivalent.

This limitation is consistent with the requirements of General Design Criteria 19 of Appendix "A",10 CFR 50.

l

~

The control room radiological dose calculations use the conservative minimum acceptable flow of 2250 cfm based on the flowrate surveillance requirement of 2500 cfm 10%.

J Currently there are some situations where the CREV System may not automatically start on an accident signal, without operator action. Under most situations, the emergency filtration fans will. start and the CREV System will be in the accident lineup. However, a failure of a supply fan (F21A or B) or an exhaust fan (F31A or B), operator action will be required to return to a full train lineup. Also, if a single emergency bus does not power up for one train of the CREV System, the opposite train filter fan will automatically start, but the required supply and exhaust fans will not automatically start.

Therefore, operator action is required to establish the whole train lineup.

This action is specified in the Emergency Operating Procedures. The radiological dose calculations do not take credit for CREV System cleanup action until 10 minutes into the accident to allow for operator action.

When the CREV System is checked to shift to the recirculation mode of operation, this will be performed from the normal mode of operation, and from the smoke purge mode of operation.

MILLSTONE - UNIT 2 B 3/4 7-4 Amendment No.

0419

.. =. - _ -

-. -..- _~.-....

Docket No. 50-336 B17413 Millstone Nuclear Power Station, Unit No. 2 Proposed Revision to Technical Specifications FSAR Change

. Main Steam Line Break and Radiological Consequences i

September 1998

i MNPS 2 FSAR 6.4.4 Availability and Reliability 6.4.4.1 Special Features The components of the containment spray system are designed to general requirements including seismic response as described in Section 6.1. All components are protected from missile damage and pipe whip by physically separating duplicate equipment, as described in Section 6.1.

To assure the availability of water to the pumps, separate suction headers from the refueling water storage tank are provided for the spray pump located in the two separate and shielded pump rooms, which house the pumps of the engineered safety features systems. Each of the two pump rooms contains one spray pump, one low-pressure safety injection pump and one high-pressure safety injection pump. Two separate headers, one to each of these pump rooms, are also provided from the containment sump.

The containment spray pumps are located in the lowest elevation of the auxiliary building at Elevation (-) 45-6 to assure a flooded suction. This assures pump priming and protects the mechanical seals in the spray pumps. In this location, the available NPSH is always greater than the required NPSH (see Table 6.41).

To assure adequate design margins, the available NPSH for the contsinment spray pumps is conservatively calculated at 27 feet during the recirculation mode in accordance with Safety Guide 1. This is based on a containment sump water level at Elevation (-) 22-6, neglecting containment pressure and assuming that all the safety injection and containment spray pumps are operating at their respective runout conditions. The calculated minimum containment sump water levelis at Elevation (-) 15-6. The peak calculated containment pressure is 51.2 psig with 250 F sump water and 14.7 psig with 92 F sump water lg To increase system reliability, the containment spray pump motors have the capacity to start with the motor operated valves on the discharge header fully opened.

The refueling water storage tank (RWST) and containment sump assure sources of water for the containment spray system. These components are described in Section 6.2.

j The liberation of combustible gases, resulting from metal corrosion in the containment postaccident environment,is described in Subsection 14.8. The contribution to metal corrosion from a continuous borated water spray within the containment is less than 200 mils /yr. Therefore, with the brief exposure to the containment sprays during postinci-dent conditions, this corrosion is negligible.

No :::d9 E -

r 'ce th: re&cticr :f f::: cn p;cd :t ::n::rt::t!:n in tM ;cnt:Inment -

ct..,..au h u... L; k ated :;:cy.

A failure mode analysis is given in Table 6.4-2.

Inadvertent initiation of the spray system does not offect the safety of the unit, since within the containment all the instruments are dripproof or weatherproof, all the motors are dripproof or totally enclosed and signal cable runs are enclosed in waterproof jackets. All piping or equipment insulation which may come in contact with sprays are of the metal enun 6.4 4 March 1996

MNPS 2 FSAR high radiation levels. Operation of the CRFS is monitored by filter bank differential pressure and temperature indication. Fan operation is monitored by motor trip alarms, in t.a\\

st /97 )

in the event of a LOCA or a fuel handling accident, the control room air conditioning system is automatically switched to the isolation / recirculation mode. Tests show that the p.n) unfiltered in-leakage is less than 100 cfw/

3 g cf A fresh air make-up system will not be used to maintain a positive pressure differential with respect to the external environment or the adjacent intarnal spaces at any time during the normal or emergency modes of operation.

The control room air conditioning system mode of operation includes an automatic isolation of the system to the complete recirculation mode and automatic initiation of the bypass filtering operation. This automatic switchover to the complete recirculation mode and L

filtering mode is initiated by the EBFAS or the AEAS.

The post-accident mode of operation is a closed cycle with air intakes and outlets isolated.

The control room atmosphere is exhausted from the space, filtered, and cooled as required {gs.uJ ms and returned to the space. Outside air is not introduced into the system unless required for personnel safety.

9.9.10.4 Availability and Reliability 9.9.10.4.1 Special Features The components of the control room air conditioning system are designed to engineered safety feature requirements including seismic response as described in Section 6.1. All components are protected from missile damage and pipe whip by physical separation of duplicate equipment, as described in Section 6.1.

Each air conditioning subsystem is capable of maintaining a suitable environment within the control room. Each system is designed for the normal control room cooling load which is greater than the cooling requirements under post-accident operation. Each system is

[fr.-q completely independent, including the control and filtration systems with the exception of

\\'/'7 /

some common ductwork and dampers. Common components such as dampers are isolated during post-accident operation. Control inputs to these devices are overridden.

Each subsystem is powered by a separate emergency source (Section 8.3). A failure mode analysis for the control room air conditioning system is given in Table 9.9-17. Although there are common plenums, all ductwork is considered a passive component not subject to a single failure mode.

j.

The charcoal filter elements within the CRFS are analyzed to ensure adequate residual heat (41-sf sale?j removal capabilities following any single failure. The analysis concludes that the maximum temperature calculated, based on a radioactive filter inventory which was conservatively assumed to be ten (10) times greater than the maximum inventory calculated resulting from a design basis accident at the site, was less than 212*F(100'C). This is substantially below the charcoal ignition temperature, thus filter bed isolation should not constitute a fire hazard. Temperature indication is provided to alert personnel of excessive charcoal bed temperature.

l ese. w 2 9.9-29 June 1998 /

1,

'~

\\

MNPS 2 FSAR For the hot shutdown case, the event is initiated by a rapid opening of the atmospher.,

dump valves or the turbine bypass valves resulting in a steam flow increase of 41% of the nominal full power steam flow. A bounding value for the negative MTC was assumed as was the technical specification value of the shutdown margin. The results of this event for both the one pump and four pump case were found to be bounded by the full power, full flow event.

I i

l The responses of key system variables are given in Figures 14.1.31 to 14.1.3-7 for the rated power case. The sequence of events is given in Table 14.1.3 3.

14.1.3.7 Conclusion The results of the analysis demonstrate that the event acceptance criteria are met since the minimum DNBR predicted for the full power case is greater than the safety limit. The correlation limit assures that with 95% probability and 95% confidence, DNB is not l

expected to occur therefore, no fuelis expected to fail. The fuel centerline melt threshold of 21 kW/ft is not violated during this event.

14.1.4 Inadvertent Opening of a Steam Generator Relief or Safety Valve 14.1.4.1 Event initiator This event is initiated by an increase in steam flow caused by the inadvertent opening of a 5/10 l'

l secondary side safety or relief valve.

N'3 14.1.4.2 Event Description

}

The resulting mismatch in energy generation and removal rates results in an overcopling of the primary system. If the MTC is negative, the reactor power willincrease.

14.1.4.3 Reactor Protection 6

Reactor protection is provided by the variable overpower trip, LPD trip, TM/LP trip, low secondary pressure trip, and low steam generator water level trip. Reactor protection for the inadvertent Opening of a Steam Generator Relief or Safety Valve event is summarized in Table 14.1.4-1, 14.1.4.4 Disposition and Justification The inadvertent opening of a steam generator safety valve would result in an increased steam flow of approximately 6.75% of full rated steam flow. Each dump (relief) valve is sized for approximately 7.50% steam flow with the reactor at full rated power. As such, the consequences of any of these occurrences will be bounded by the events in Sec-l tion 14.1.3. The disposition of events for the Inadvertent Opening of a Steam Generator L

Relief or Safety Valve event is summarized in Table 14.1.4-2.

+i.1.5 Otcom4yotem Piping FodwceAnoido-end Outsid cf Centcim.cnt nsxc c

(

--44.1.5.1 E w n:l..::;ter 14st.un 14.1-5 April 1993 l 1

=

F l

MNPS-2 FSAR his event is initiated by a rupture in the main steam piping upstream of the MSIVs which 3

l re Its in an uncontrolled steam release from the secondary system.

14.1. 2 Event Description s

The increa in energy removal through the secondary system results in a severe overcool-ing of the pn ary system. In the presence of a negative MTC, this cooldown causes a decrease in the hutdown margin (following reactor scram) such that a return to power 1

might be possibi ollowing a steam line rupture. This is a potential problem because of the high power pea 'ng factors which exist, assuming the most reactive control rod to be stuck in its fully with awn position.

14.1.5.3 Reactor Protec 'on Reactor protection is provide y the low steam generator prc,ssure and water level trips, variable overpower trip, LPD trip TM/LP trip, high containment pressure trip, and SlAS.

IjU 4

Reactor protection for the Steam stem Piping Failures inside and Outside of Contain-ment event is summarized in Table

.1.5 1.

14.1.5.4 Disposition and Justification a

I At rated power conditions, the stored energ in the primary coolant is maximized, the 1

available thermal margin is minimized, and the re trip power level is maximized. These i

conditions result in the greatest potential for co own and provide the greatest challenge i

to the SAFDLs. Initiating this event from rated po er also results in the highest post-trip power since it maximizes the concentration of delay neutrons providing for the greatest power rise for a given positive reactivity insertion. A itional thermal margin is also provided at lower power levels by the automatically dec asing setpoint of the variable j

overpower trip. Thus, this event initiated from rated pow r conditions will bound all other cases initiated from at power operation modes.

For the zero power and suberitical plant states (Modes 2-6), th re is a potential for a return to-power at reduced pressure conditions. The most limiti steam line break (SLB) event at zero power is one which is initiated at the highest tempe ture, thereby providing the greatest capacity for cooldown. This occurs in Modes 2 and 3. Thus, the event initiated from Modes 2 and 3 will bound those initiated from Modes 4. Further, the limiting initial conditions will occur when the core is just critical. These onditions will maximize the available positive reactivity and produce the quickest and I est return to power. Thus, the SLB initiated from critical conditions in Mode 2 will boun the results of the event initiated from suberitical Mode 3 conditions.

)

The technical specifications (Reference 14.1-1) only require a minimum of one P to be lWI operating in Mode 3. One pump operation provides the limiting minimum initial co flow case. Minimizing core flow minimizes the clad to coolant heat transfer coefficient a d degrades the ability to remove heat generated within the fuel pins. Conversely, howeser, i

a maximum loop flow will' maximize the primary to secondary heat transfer coefficient, j

thus providing for the greatest cooldown. Higher loop flow will sweep the cooler fluid in the core faster, maximizing the rate of positive reactivity addition and the peak power

(

level.

14st m 2 14.1 6 April 1993 4

s MNPS 2 FSAR

~

The orst combination of conditions is achieved for the four pump loss of offsite power

(..

case, n this situation, the initial loop flow is maximized resulting in the greatest initial cooldowg, while the final loop flow is minimized providing the greatest challenge to the DNB SAFDL. Since the natural circulation flow which is established at the end of the transient wi be the same regardless of whether one or four pumps were initially operating, the results o he four pump loss of offsite power case will bound those of the one pump.

l case. Thus, o four pump operation need be analyzed for the Mode 2 case.

The event is analy ed to support a more negative MTC. This event must be analyzed both with and without a incident loss-of offsite power. Typically, thero are two single failures which are con idered for the offsite power available case. The first is failure of a High Pressure Safety in etion (HPSI) pump to start. The second is failure of an MSIV to close, resulting in a conti ed uncontrolled cooldown. However, Millstone 2 has combina-tion MSIV/ swing disc chec valves. A double valve failure would thus be required for i

steam from the intact steam enerator to reach the break. This is not deemed credible.

Thus, the single failure to be c nsidered with offsite power available is failure of a HPSI pump to start. For the loss of-o fsite power case, the limiting single failure is the failure of a diesel generator to start. This i assumed to result in the loss of one HPSI pump and one charging pump. The disposition of vents for the Steam System Piping Failures inside and Outside of Containment event is su arized in Table 14.1.5-2.

14.1.5.5 Definition of Events Analyze The SLB event is initiated by a double ende guillotine break of the main steam line at its largest point between the steam generators d the flow restrictors. This break location leads to an uncontrolled steam release from th secondary. The event occurs concurrent with the most reactive control rod stuck out of a core.

The increase in energy removal through the secon ry system results in a severe overcool-ing of the primary system. In the presence of a nog ive MTC, this cooldown results in a large decrease in the shutdown margin and a return to ower. This retum to power is exacerbated because of the high power peaking factors vhich exist, with the most reactive control rod stuck in its full withdrawn position.

The consequences for the event are bounded by analyzing at oth HZP and Hot Full Power (HFP) conditions. At HFP conditions the stored energy.in the p ' mary coolant is maxi-mized, the available thermal margin is minimized and the pre trip ower level is maximized.

These conditions result in the greatest potential for cooldown. Init ting this event from i

rated power also has the potential for the highest post-trip power sin e it maximizes the concentrations of delayed neutrons thus providina for the greatest po r rise for a given positive reactivity insertion. If the event occurred at lower power, addit nal thermal margin is provided by the automatically decreasing setpoint of the variabl verpower trip.

Thus this event, initiated from full rated power conditions, will bound all oth cases initiated from at power operation modes or power levels.

For the zero power and subcritical plant states there is also a potential for a return-to-power. The most limiting SLB event at zero power is one which is initiate at the highest temperature and pressure, thereby providing the greatest capacity for cooldo n.

(

The most limiting conditions will occur when the core is critical. This condition will maximize the available positive reactivity and therefore produce the quickest and largest 14si.w2 14.1-7 April 1993 l </

MNPS-2 FSAR p

r um to power. Thus, the SLB occurring from critical conditions will bound the results of

(,

the vent initiated from subcritical conditions.

As out ed in References 14.1-2 and 14.1-3, three computerized calculations are required prior to t final calculation of the Minimum Departure From Nucleate Boiling Ratio (MDNBR) v ues and the maximum Linear Heat Generation Rate (LHGR) values utilized in the determin ion of fuel failure. The NSSS response is computed with Siemens Power q1-is Corporation (S

-RELAP) (developed from RELAP5/ MOD 2 (Reference 14.1-4), with SPC qfg3 modifications (R erence 14.1-2), the detailed core and hot assembly power distributions and reactivity spot hecks are computed with the SPC three-dimensional core simulator model, XTG (Refere e 14.1-5), and the detailed core and hot assembly flow and enthalpy distributions are comp ted with XCOBRA-IllC (Reference 14.1-6). The modified Barnett correlation was utilized calculate MDNBR due to the reduced pressures occurring during the SLB event.

14.1.5.5.1 Analysis of Res is The SPC-RELAP analysis provides he NSSS boundary conditions for the XTG and the q 3-16 XCOBRA-IllC (Reference 14.1-6) c ulations. This section presents a description of the treatment of factors which can have significant impact on NSSS response and resultant MDNBR and LHGR values. The plant s ecific parameters used in th.is analysis are listed in Tables 14.1.5-3 to 14.1.5-5. Conservat ms are included in parameters or factors known to have significant effects on the NSSS pe formance and resulting MDNBR and LHGR values.

14.1.5.5.1.1 Break Location, Size, and Flow M del 5/qe The limiting break, a double ended gu?!!otine break,

• located inside containment between the steam generator outlet and the flow restrictors, his break location results in the 1

largest cross sectional flow area and will therefore pro uce the most rapid cooldown and

)

the highest return to power. The break flow areas for t affected and intact steam generators are listed in Table 14.1.5 3. These areas corre pond to the locations in the flow path where choked flow will occur.

The SPC-RELAP break mass flow rate is computed using the ody critical flow model lq3.is modified such that only steam flows out the break.

14.1.5.5.1.2 Boron injection Boron injection into the primary system acts to mitigate the return to p war. Injection of boron is modeled from two sources, the HPSI and the charging system.

he characteris-tics of the HPSI and charging systems are listed in Table 14.1.5-3. Both s tems are conservatively assumed in this analysis to take suction from the Refueling ter Storage Tank (RWST). The line volume between the check valves isolating these syst s pumps and the cold leg injection location is assumed to initially contain no boron. The ye required to sweep this unborated water from these lines with borat'ed water is inc ded as an integral part of the SPC-RELAP NSSS calculation. The delivery curve for the H I

(.

system used in this analysis is given in Figure 14.1.5-1.

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14st m 2 14.1 8 April 1993 s

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MNPS-2 FSAR l

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14.1.5.5.1.3 Single Failure Assumption The single failure assumed in the engineered safeguards system was the failure of one of t e two HPSI pumps required to be in service during normal operation. Also only one 1

i ch ging pump is assumed to be available. This assumption results in an additional delay in th time required for the boron to reach the reactor core. This delay is further amplified when mbined with the assumption of a stagnant upper head which serves to malatain l.

primary stem pressure due to flashing of the hot fluid in the upper head. Although one l

charging p was assumed available, the impact of crediting charging has been evaluated f[

and determi d not to invalidate the conclusions of this analysis, i

14.1.5.5.1.4 F dwater For the HZP scena 'os the AF flow was initialized such that steam flow equaled the heat generated by the R

s. No decay heat is assumed to maximize the cooldown. After the i

initiation of the transi t, the AF flow is all6wed to increase with decreasing steam generator pressure ass ing a fixed control valve setting. At 180 seconds (conservatively based on technical speci ations (Reference 14.1-1),the AF is increased to pump runout flow. All flow is directed i o the affected steam generator to me::imize the cooldown rate.

In the HFP cases the main feed ter flow will be terminated 30 seconds after the reactor trip occurs due to closure of the edwater regulator vs.lves. After reactor scram, the feedwater flow increases as the se ndary pressure decreases at the lowest possible fluid temperature until the regulators are c sed. Fluid temperature is determined by assuming all heating of feedwater ceases after t time of the break. The AF is modeled as in the HZP cases after 180 seconds.

14.1.5.5.1.5 Trips and Delays Trips for the HPSI, charging system, main feed ater valves, and MSIVs are given in Table 1.4.1.5-4. Biases to account for uncertaint s are included in the trip setpoints as shown. For the steam and feedwater valves, the lay times given are between the time the trip setpoint is reached and the time full valve el ure is reached. For the HPSI and charging pumps, the delay time given is from the time he setpoint is reached until the pumps have accelerated to rated speed. Additional del time required to sweep the lines of unborated water is accounted for by setting the boron oncentration of the injected flow to zero until the volume of the injection lines has been inje ed.

14.1.5.5.1.6 Neutronics The core kinetics input for this calculation consisted of the minim m required c'ontrol rod shutdown worth at the EOC, and EOC values associated with the r etivity feedback curves, delayed neutron fraction, delayed neutron fraction distributio and related time.

constants, and prom' t neutron generation time. The SPC-RELAP defa It fission product p

and actinide decay constants were utilized for this calculation.

The core reactivity is derived from input of several functions. These inclu e effects from control rod worth, moderator density changes, boron concentration, and Do pler effects.

The reactivity is weighted between the core sectors. Different reactivity fu er

.(

utilized where necessary for the HZP and the HFP cases. The SPC-RELAP anal:tions rses were I

usi.w 14.1-9 Jun 1994 l

q MNPS 2 FSAR 4

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(

pe formed with an MTC of -28 pcm/*F. A summary of the nuclear input and assumptions

(

is g en in Table 14.1.5 5.

14.1.5.

1.7 Decay Heat The presen of radioisotope decay heat at the initiation of the SLB event will reduce the rate and the e ent of cooldown of the primary system. For the HFP case, decay heat was calculated on th basis of infinite irradiation time prior to transient initiation. For the HZP case, there was n decay heat at the transient initiation, but decay heat was calculated based on the core p er and the time at power calculated during the transient. This treatment of decay he serves to maximize the stored energy in the HFP cases and to minimize it in the HZP ca

s. This treatment provides limiting stored energy conditions for the SLB cases.

14.1.5.5.1.8 Nodalization The NSSS transient calculations pr\\egnted in this report utilized the nodalization model described in Reference 2. The nodalizhtion treats all major NSSS components and subcomponents as discrete elements, w the exception of the secondary side of the b0 steam generators. In addition, all compon ts with long axial dimensions are divided into subcells adequate to minimize numerical dif ion and smearing of gradients, in order to simulate the asymmetric thermal hyd ulic and reactivity feedback effects that occur during an SLB transient, the core is nodalize into three radial sectors. One sector f

corresponds to the region immediately surrounding t assembly where the most reactive control rod is assumed stuck out of the core. This set or is termed the " stuck rod" sector.

The remainder of the region of the core which is directl ffected by the loop containing the break is the second sector and is termed the "affecte " sector. The remainder of the core and the other loop is termed either " unaffected" or "in et" sector or loop.

14.1.5.5.1.9 Interloop Mixing During an actual SLB transient, some mixing betwsen the parallel annels within the reactor pressure vessel will occur in the downcomer, the lower plen

, the core, and the upper plenum due to lateral momentum imbalances, and turbulence or ddy mixing. The mixing will act to reduce the positive reactivity feedback effects due to reduced rate and magnitude of cooldown of the affected loop and associated core sector.

In this analysis, no credit is taken for turbulent or eddy mixing of coolant be een loops or the parallel flow channels within the reactor pressure vessel (RPV). However nterloop mixing is calculated to occur due to flow in interloop junctions in the upper an ower plenums. Mixing in the lower plenum was reduced to a minimum by using an e emely high loss coefficient between the affected and intact sectors.

14.1.5.5.2 Minimum Departure From Nucleate Boiling and Linear Heat Generation ate MDNBR calculations require determination of the power, enthalpy, and flow distribution within the highest power assembly of the stuck rod core sector. Similarly, determination i

(,

of the maximum LHGR also requires characterization of the power distribution. The powe I

distribution within the core, including the highest powered assembly within the stuck rod

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14 w 2 14.1 10 April 1993 l'

+ *

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MNPS 2 FSAR r'

cor,e sector, is calculated with XTG (Reference 14.1-5). Flow and enthalpy distributions ll

(

wit the core, including the highest powered assembly within the stuck rod core sector,

=

are ca ulated with XCOBRA IllC (Reference 14.1-6). In order to obtain compatible flows, l Adeu%-Ig moderat densities, and powers within the high power assemblies, iteration between XTG and XCO

-IllC is conducted.

For this calcula 'on, the modified Barnett correlation was found to be suitable for the MDNBR calculati. The modified Barnett correlation is based upon closed channels and primarily uniform p er distribution data. The correlation is based on assembly inlet (or upstream) fluid condi ons rather than on local fluid conditions as is the case with subchanne. based corr tions. Use of the correlation is limited to the range of the data base unless conservative xtrapolations can be made.

14.1.5.6 Analysis Results A summary of calculated results portant to this analysis is presented in Table 14.1.5 6 for the four scenarios analyzed. Th MDNBR values are listed together with the corre-sponding core power values at the ti of MDNBR which corresponds to the maximum post scram power level. For cases wh e offsite power is available for operation of the primary coolant system pumps, MDNBR d maximum LHGR occurs at the m?@: um 5[10 power condition. For cases where offsite wer is lost and the primary 9ste ornps coast down, the maximum LHGR and MDNB however, occur when the. worst cambina-tion of core power, flow, inlet temperature, an ressure is present. Thes. wnditions occurred at the time of peak power in this analys.

The scenario which results in the highest post scram ower level and largest LHGR is that initiated from HZP with offsite power available for oper ion of the primary coolant pumps.

The general post trip response of the NSSS for the HFP s nario with offsite power.

available is comparable to that for the HZP scenario. One caption is the post scram seberitical core power response during the initial portion of t transient. Tne post scram suberitical power response is different for the HFP case due to elayed neutron and stored energy effects. In the HFP case the scram shutdown margin is rge enough that by the time the reactivity reaches zero most of the delayed neutrons are o longer in the system inhibiting a return to power. Because the HZP case results in a high r power level and higher LHGR, it is presented in detail.

The NSSS responses for the scenarios with loss of offsite power for op sation of the primary system coolant pumps are different from those scenarios where difsite power is available throughout the transient due to the pump coastdown and subseg nt natural circulation of the primary coolant. Post scram maximum power levels attain during the transient are significantly lower. Lower power levels result from lower positiv moderator feedback. The positive moderator feedback is reduced due to the coolant densi reduc-tions that occur axially upwards in the core at low core flow rates, even for low c re power levels. Lower power levels cause MDNBR values to increase, but lowering w

rates cause MDNBR values to decrease. Overall, the combination of factors results a lower MDNBR values for the reduced flow condition than for the full flow condition.

Of the two loss of offsite power scenarios analyzed, the HZP case results in lower MDN

(

values. The general response of the HFP and HZP cases with loss of offsite power is comparable. Again, the exception is the post scram subcritical core power response

/

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us t.w2 14.1-11 April 1993 /

MNPS 2 FSAR ring the initial portion of the transient. The post scram subcritical power response is dif rent for the HFP case due to delayed neutron and stored energy effects. Because the two enarios are quite similar in terms of their general response, only the limiting MDNBR case (i.., HZP without offsite power) is presented in detail.

14.1.5.6.1 Hot Zero Power with Offsite Power Available 93-11r The SPC-RELA simulation of the NSSS during the HZP transient with offsite power l 3!?

available is illustr ed in Figures 14.1.5-2 through 14.1.5 7. A tabulation of the sequence of events is presen d in Table 14.1.5-7. The SPC-RELAP computation was terminated l 9)*9 600 seconds after br k initiation. This is well beyond the time of MDNBR or peak LHGR.

Beyond 600 seconds, c re reactivity would become more subcritical as dryout of the steam generator occurs f owing AF termination. Termination of the AF by manual operator action was assum to occur 600 seconds after initiation of the break.

14.1.5.6.1.1 Secondary Syste Thermal Hydrau'lic Parameters Steam flow out the break is the so e of the NSSS cooldown. Steam flow for the affected generator is plotted in Figure 14.1.5-2. The affected steam generator continues to blow down through the break throug out the transient. The pressure and mass flow rate drop rapidly at first and then proces downward at a slower decay rate until 238 seconds. At that time, the cool AF conde es a significant quantity of steam and the g30 break flow essentially goes to zero. The co own of the secondary side produces a change in heat transfer regimes between the

' mary and secondary which results in a heatup of the primary coolant. The higher temp rature reduces the reactivity present and power drops rapidly.

The intact steam generator blows down for a short p iod until the MSIVs completely close approximately 10.5 seconds after the break is ini ted. The pressure recovers as the intact steam generator equilibrates with the primary stem and then slowly decays as the intact steam generator begins to act as a heat source t the primary system.

14.1.5.6.1.2 Primary System Thermal Hydraulic Parameters The primary system coolant temperature and pressure responses esulting from the break flow are illustrated in Figures 14.1,5 3 through 14.1.5-5. The pri ry system pressure decays rapidly as the coolant contracts due to cooldown and the pr surizer liquid empties. The MSIVs close at 10.5 seconds, ending the blowdown o he intact steam generators and reducing the rate of energy removal from the primary fl 'd. Primary system pressure recovers somewhat at that point, and then increases slowly for he duration of the transient.

14.1.5.6.1.3 Reactivity and Core Power The reactivity transient calculated by SPC-RELAP is illustrated in Figure 14.1.5-Initially,

{

the core is assumed to be. critical at HZP. All control rods, except the most reacti e one, are assumed to be inserted into the core following the first reactor trip signal. The reactivity transient then proceeds. Cooldown of both the coolant and fuel brings the core

[

critical due to moderator and Doppler reactivity feedback. Shortly thereafter, powergins

\\

to rise steadily due to the dominating positive reactivity feedback from the moderator, usw 14,g.12 April 199

/

MNPS-2 FSAR g

e HPSI and charging flow is actuated 45.2 seconds after the break. Borated water

(

pa es through the core 153 seconds after the break initiation following a 30 second HPSI dela a 40 second charging pump gartup delcy, a Ine flushing delay, and a transport delay tween the cold leg injection point and the core. Entry of borated water into the core hel override the positive moderator feedback ed helps to terminate the increase in core powe Core power then begins a decline as the concentration of boric acid increases with time. A 238 seconds, the power drops rapidly due to increasing primary coolant temperature.

llowing the rapid power drop, the power declines much slower as the boron concentrat n increases. Terminating AF 600 seconds after the break will subse-quently cause the imary coolant to heat up. This, combined with the ever increasing bc on concentration, ill terminate the SLB event.

Figure 14.1.5-7 shows t e transient reactor power. The maximum power level is 686 5/To MWt or 25% of rated po r at 153 seconds after the break initiation.

14.1.5.6.1.4 XT3 and XCOB

-IllC Results 43-i f The XTG calculation is made initia on the basis of SPC-RELAP input. Each assembly

!Wi3 within the three channels is assume to have a uniform flow corresponding to the sector flows calculated with SPC-RELAP. D to high power peaking in the region of the stuck q3 4 control rod, large moderator density re etions are calculated to occur in the top portions of several assemblies in this region of th core in the XTG calculation. This moderator density decrease is a major factor in the fl tening of the axial and radial profiles, and the significant reduction in reactivity observed en XTG is compared to SPC-RELAP.

l U~II The SPC-RELAP reactivity and power calculatio has considerable inherent conservatism.

l *8 To demonstrate this, a comparison of the chang 'n reactivity at the maximum LHGR time is made. A comparison of the overall change in re etivity between SPC-RELAP and XTG g.g shows that SPC-RELAP conservatively underestimat s the negative reactivity by 2.314 at j

the start of the transient and overestimates the reacts ity at maximum LHGR time by 8.74 6, thus indicating that the SPC-RELAP power calculatio is conservative. It should be l 91-1g noted that the XTG calculated reactivities are best estim te at both the initial and maxi-mum LHGR conditions.

An XCOBRA-IllC core analysis was conducted to define the f w and enthalpy distribution within the high power assembly. The XCOBRA-IllC core flow 'stribution analysis indicates that the flow, and therefore moderator density, in the pper elevations of the high power assembly is greater in the closed channel XTG calcula 'on than the open channel XCOBRA IllC calculation. The power distribution and react ity calculated by XTG are therefore conservative.

14.1.5.6.1.5 Departure From Nude'te Boiling Ratio and Linear Heat Ge ration Rate For the MDNBR portion of the calculation, the radial power distribution'was odified to conservatively account for local rod power distribution affects within the hot ssembly.

This was done by raising the power of the hot assembly by an additional 15% o bound the peak rod power.

{

On the bases of these conservative assumptions, the MDNBR value was calculate to be 2.40. This compares to a 95/95 DNBR limit of 1.135 for the modified Barnett corre tion.

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1est.Mn 14.1 13 April 1. 3 !

t

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MNPS 2 FSAR

.T refore, no fuel rods would be expected to fait during this transient scenario from an i

MD R standpoint.

L The ana is of the peak LHGR also comes from the XTG and XCOBRA-lllC analysis. The peak LHG

's calculated from the SPC RELAP total core power and the XTG radial and l qptg HM3 axial peakin The peak LHGR,20.994 kW/ft, was calculated for the case of HZP with offsite power allable. Comparing this LHGR with a centerline melt criteria of 21 kW/ft, it is apparent that nterline melt is not predicted to oc::ur. Thus, no fuel failures are predicted to occur ue to violation of the centerline melt criteria.

i 14.1.5.6.2 Hot Zero ower with Loss of Offsite Power l

The SPC-RELAP NSSS sim ation of the most limiting SLB scenario from an MDNBR lM standpoint (i.e., HZP with lo of offsite power) is illustrated in Figures 14.1.5-8 through 14.1.5 13. A tabulation of th equence of events is presented in Table 14.1.5 8.

Termination of the AF by manua perator action was assumed to occur 600 seconds after initiation of the break. This is wel eyond the tir-of MDNBR and maximum LHGR.

Following termination of AF, core re tivity woutc come more subcritical due to continued addition of boron and event I dryout of the affected steam generator.

j 14.1.5.6.2.1 Secondary System Thermal ydraulic Parameters Steam flow out the break is the source of the. SSS cooldown. Steam flow for the affected steam generator is plotted in Figure 14.

5-8. The affected steam generator continues to blow down through the break throug ut the transient. The pressure and mass flow rate drop rapidly at first and then proces downward at a slower decay rate.

The intact steam generators blow down for a short per d until the MSIVs completely close approximately 10.5 seconds after the break is initi ed. The' pressure recovers as 1

the intact steam generator equilibrates with the primary s tem. Subsequently, the intact steam generator pressure remains essentially constant as th primary intact coolant loop approaches natural circulation conditions.

14.1.5.6.2.2 Primary System Thermal Hydraulic Parameters The primary system core coolant temperature and pressure respons resulting from the break flow are illustrated in Figures 14.1.5-9 through 14.1.5-11. The rimary system pressure decays rapidly as the coolant contracts due to the cooldown a d the pressurizer empties. Continued pressure reduction in the primary system causes the elatively hot stagnant liquid in the head of the RPV vessel to flash. The flashing in the per head, coupled with near equilibration of other NSSS parameters, retards the press e decay from that point forward. The elevated pressere acts to limit the delivery of boron i to the core due to the pressure versus flow characteristics of the HPSI system.

A comparison of intact and affected core sector inlet temperatures throughout th transient indicates significant differences due to the limited cross flow allowed bet een i

loops. The core sector flows all show the same trend due to the coastdown of the primary coolant pumps. That is, all flows decrease rapidly until natural circulation conditions are achieved in the two flow loops.

t j

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usun 14,114 April 1993

MNPS 2 FSAR N

r 1.1.5.6.2.3 Reactivity and Core Power l

(

The r ctivity transient calculated by SPC-RELAP is illustrated in Figure 14.1.5 12.

93'%

' Initially, he core is assumed to be critical at HZP. All control rods, except the most W

reactive o, are assumed to be inserted into the core at the start of the transient.

Cooldown o oth the coolant and fuel brings the core critical due to moderator and Doppler reacti 'ty feedback. After reaching criticality, the power spikes momentarily, but fuel temperature ises rapidly and Doppler feedback effects rapidly reduce core reactivity and power. Short thereafter, power rises again, then stabilizes as the affected core sector average mode tor temperature stabilizes. Although the affected core sector inlet temperature continues o decrease during this period, the flow rate is also decreasing, thus stabilizing the affected c e sector average moderator temperature.

The HPSI and charging flow actuated 48.7 seconds after the break and the shutdown effect of boron is superimpose upon the other reactivity feedback effects. Borated water

_ passes through the core 152 se nds after the break initiation, following a line flushing Sl'Io delay and a transport delay betwe the cold leg injection point and the core. Entry of borated water into the core initiates general power descent which would ultimately bring the reactor to a shutdown condition a the concentration of boron increases with time.

Terminating AF 600 seconds after the b ak will subsequently cause the primary to heat up. This, combined with the ever increasi boron concentration, will finally terminate the SLB event.

The transient experienced by the core power is 'lustrated in Figure 14.1.5-13. A small power spike is calculated to occur at 63 seconds fter the break is initiated. However, it is of such short duration that fuel temperatures an core heat flux do not increase sufficiently to cause any DNS concern at that partic ar point in time. The next maximum power levelis 293 MWt or 11% of rated at 169 seco ds after the break initiation. -

14.1.5.6.2.4 XTG and XCOBRA lllC Results The XTG calculation is initially made on the basis of SPC-R P predicted core power, f39 flow, pressure, and inlet temperatures. The XTG calculations rovide the radial and axial power distributions for use in the XCOBRA IllC code. Due to t high power peaking in the region of the stuck control rod, and the low core average nat ral circulation flow rates, large moderator density decreases are calculated in several assem ies in this region in the XTG calculation. This is a major factor in the flattening of the axial nd radial profiles, and the significant reduction in reactivity observed when XTG is compare to SPC-RELAP.

l93-tg XCOBRA-lllC analysis is also conducted to define the flow and enthalp distribution within i

the high power assembly.

The absolute difference in reactivity between SPC-RELAP and XTG indicates hat the SPC-q 3-ty RELAP power calculation is conservative. The SPC-RELAP reactivities at HZP nd the MDNBR point are calculated to be -8.00 $ and 0.046 0, respectively and the X values are calculated to be -10.31 $ and -9,91 $.

14.1.5.6.2.5 Departure From Nucleate Boiling Ratio and Linear Heat Generation Rate Results s

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14si m er 14.1-15 April 1993 /

i MNPS-2 FSAR L-A imilar approach to that taken for the HZP power available case was utilized for the limiting MDNBR case, which is the HZP scenario with loss of offsite power. The mo MDN R of the hot fuel assembly is calculated to be 1.18. This compares to a 95/95 DNBR it of 1.13S for the modified Barnett correlation. Therefore, no fuel rods would be expecte o fail during this transient scenario from an MDNBR standpoint.

As before th analysis of the peak LHGR comes from the XTG and XCOBRA-IllC analysis.

The peak LHG was 16.5 kW/ft. Comparing this LHGR with a centerline melt criteria of 21 kW/ft, it is a arent that centerlirw melt is not predicted to occur. Thus, no fuel l

failures are predic d to occur due to violation of the centerline melt criteria.

l 14.1.5.6.3 Hot Fui ower Without Offsite Power Available The sequence of events r the case is presented in Table 14.1.5-9. For reasons present-ed in Section 14.1.5.6 this ase was not discussed in detail.

14.1.5.6.4 Hot Full Power th Offsite Power Available The sequence of events for the ca e is presented in Table 14.1.5-10. For reasons presented in Section 14.1.5.6 this se was not discussed in detail.

5[90 14.1.5.7 Conclusions The HZP scenario with loss of offsite powe was determined to be the most limiting in this

'i analysis from an MDNBR standpoint. The H and HZP scenarios, with offsite power maintained for operation of the primary coolan pumps resulted in a return to higher power levels than the scenarios where offsite power is st. However, these scenarios provide substantially greater margin to the MDNBR limit ause of the higher coolant flow Tate.

In no scenario evaluated, however, was fuel failure alculated to occur as a result o.'

penetration of the MDNBR safety limit.

The HZP scenario with offsite power available was deter 'ned to be the most limiting in this analysis from the standpoint of centerline melt. This s nario results in the highest return to power and highest calculated LHGR of 20.9 kW/ft.

e HFP and HZP scenarios with offsite power maintained for operation of the primary coo nt pumps returned to higher power levels than the scenarios where offsite power is lo Even though these scenarios have substantially greater margin to the MDNBR limit be use of a higher i

coolant flow rate, the higher power levels in combination with the hi ly skewed power distribution due to the assumed stuck rod cluster resulted in them havi the least margin to the fuel centerline melt limit.

I(

l 5*sim 14.1 16 April 1993 l

1 MNPS-2 FSAR REFERENCES 14.1-1 gehnical Specifications for Millstone Unit 2, Docket No. 50 336, Updated through i

Amendment No.116.

14.12 SPC-84 3(P), Supp.1, " Steam Line Break Methodology for PWRs," Advanced Fuels Co any Richland, WA, June 1988.

14.1-3 XN-NF-84-93(

"Steamline Break Methodology for PWRs," Exxon Nuclear 91-ir Company, Richla, WA, November 1984.

y/93 AJa...

14.1-4 "RELAP5/ MOD 2 Code anual, Volume 1: System Models and Numerical Methods:

Volume 2, Users Guide' d input Requirements," NUREG/CR-4312, EGG-2396, Revision 1, EG&G Idaho, in., Idaho Falls, ID 83415, March 1987.

14.1-5 "XTG-A Two-Group Three Di nsional Reactor Simulator Utilizing Coarse Mesh Spacing," XN-CC-28(A), Revision Exxon Nuclear Company, Richland, WA 99352, October 1978.

14.1-6 "XCOBRA-IllC: A Computer Code to De rmine the Distribution of Coolant During Steady-State and Transient Core Operation," XN-NF-75-21(A), Revision 2, Exxon Nuclear Company.

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14s t.w2 14.1-17 April 1993 /

The following pages are " Insert C," Section 14.1.5 and Section 14

References:

e e

f 1

l l

MNPS-2 FSAR 14.1.5 Steam System Piping Failures inside and Outside of Containment l

Two separate analyses have been performed for the steam line break event. Section

)

14.1.5.1 describes the pre-scram analysis performed to determine Departure from Nucleate Boiling Ratio (DNBR) and Linear Heat Generation Rate (LHGR) up to and l

including reactor trip. This time period represents the highest reactor power condition and the assumptions have been selected to minimize DNBR and maximize LHGR during this time frame. Section 14.1.5.2 describes the post-scram analyses performed to determine MDNBR and LHGR during the return to power caused by the overcooling. A different set of assumptions and single failure were determined to minimize MDNBR and maximize LHGR for the return to power time frame.

14.1.5.1 Pre-Scram Analysis 14.1.5.1.1 Event Initiator The pre-scram SLB analysis is initiated by a rupture in the main steam piping whic'.i results in an uncontrolled steam release from the secondary system.

l i

14.1.5.1.2 Event Description The increase in energy removal through the secondary system results in a severe overcooling of the primary system. With a negative MTC, the primary system cooldown causes the reactor power level to increase. If the break is not large enough to trip the reactor on a Low Steam Generator Pressure signal, the cooldown will continue until the reactor is tripped on a Variable Overpower or TM/LP signal (for breaks outside containment) or a High Containment Pressure signal (for breaks inside containment) or until the reactor reaches a new steady-state condition at an elevated power level.

Although the SLB calculation is typically a cooldown event, for the pre-scram analysis the cooldown event is not significant for the limiting pre-scram case. The case with a loss of offsite power, also known as a " pumps off" case, credits the low reactor coolant flow trip for harsh conditions. In this case, the Reactor Coolant Pumps (RCPs) are tripped shortly after the initiatio'n of the transient. The sharp reduction in reactor coolant flow causes the pre-scram pumps off calculation to become a heat up transient very similar to a Loss of Coolant Flow (LOCF).

Therefore, the conditions for this case are biased as if it were a LOCF (i.e. BOC neutronics). This case becomes a combination of an MSLB and an LOCF event.

14.1.5.1.3 Reactor Protection l

Reactor protection is provided by the low steam generator pressure and war levr.i l

trips, variable overpower trip, LPD trip, TM/LP trip, high containment pressi.m is;r,

e l

low reactor coolant flow, and SlAS. Reactor protection for the Steam Syst x 9ing Failures inside and Outside of Containment event is summarized in Tab!a 14.1.5.1 1.

. -.. - -..~ - - - -

f MNPS-2 FSAR 14.1.5.1.4 Disposition and Justification t

HFP initial conditions are limiting for the pre-scram SLB cases since this is the highest power condition.

' The outside containment breaks do not cause harsh conditions inside containment, and therefore, do not cause the Low Reactor Coolant Flow trip to be degraded. If a loss of offsite power were concurrent with an outside containment break, the primary coolant flow rate would coastdown similar to an LOCF event, without the Low Reactor Coolant Flow trip being degraded. The outside containment break case with loss of offsite power is therefore bounded by the LOCF event.

The inside containment breaks do cause harsh conditions inside containment, and therefore, an increased allowance for instrument uncertainty was applied for the Low Reactor Coolant Flow trip. Therefore, only'the inside containment breaks will be analyzed with a loss of offsite power.

The following pre-scram HFP Steam Line Break cases for break sizes ranging up to a double-ended guillotine break in a main steam line were analyzed, with the effects of power decalibration and harsh containment conditions (where applicable) included in the analysis:

1.

Breaks outside containment and downstream of the check valves (symmetric cases) 2.

Breaks outside containment and upstream of a check valve'(asymmetric cases) 3.

Breaks inside containment with RCPs on (asymmetric cases) 4.

- Breaks inside containment with RCPs off (asymmetric cases)

The event is analyzed to support the technical specification EOC MTC limit. This event must be analyzed both with and without a coincident loss-of-offsite power.

j The single failure assumed in this analysis is the loss of one channel of Nuclear Instrumentation (NI) which provides power indication to the RPS. If one channelis i

out of service, the three remaining Ni safety channels will be in a 2-out-of-3 coincidence mode. With the assumption of a failure in one of these channels, both of the remaining channels are required for a trip, relying on the lowest power indication for the safety function.

The disposition of events for the Steam System Piping Failures inside and Outside of Containment event is summarized in Table 14.1.5.1-2.

14.1.5.1.5 Definition of Events Analyzed The pre-scram SLB event is initiated by a rupture in the main steam piping. The break location is downstream of the steam generator integral flow restrictor and either 1.

outside containment and upstream of the main steam line check valves (asymmetric break), or

i MNPS-2 FSAR 2.

outside containment and downstream of the main steam line check valves (symmetric break), or 3.

inside containment and upstream of the main steam check valves (asymmetric break).

1 Steam released through a break located downstream of the main steam line check valves flows to the break from both steam generators and, therefore, results in a l

symmetric transient. However, steam released through a break located upstream of one of the check valves flows to the break from the upstream steam generator only (because the check valve precludes backflow to the break from the other steam generator) and, therefore, results in an asymmetric transient.

Power decalibration is caused by density-induced changes in the reactor vessel downcomer shadowing of the power-range.ex-core detectors during heatup or cooldown transients. The nuclear power levels indicated by those instruments are lower than the actual reactor power levels when the coolant entering the reactor vessel is cooler than the normal temperature for full-power operation (and higher when the vessel inlet coolant is warmer than the normal full-power temperature).

This effect is included in the modeling of any power-dependent reactor trips credited in the analysis of full-power cooldown events and low-power events. The Variable Overpower trip, the Thermal Margin / Low Pressure (TM/LP) trip function, and the Local Power Density (LPD) trip all depend on the indicated nuclear power level.

Harsh containment conditions can be caused by the release of steam within the reactor containment. Under such conditions, only those trips which have been qualified for harsh environments are credited, and increased uncertainties are 1

included in the setpoints of all environmentally qualified trips which are credited.

As outlined in Reference 14.1-1, three computerized calculations are required prior to the final calculation of the Minimum Departure From Nucleate Boiling Ratio (MDNBR) values and the maximum Linear Heat Generation Rate (LHGR) values utilized in the determination of fuel failure. The NSSS response is computed using the Siemens Power Corporation (SPC) ANF-RELAP code (Reference 14.1-2), the detailed core and hot assembly power distributions and the reactivity at the time of peak post-scram power are calculated using the SPC XTGPWR code (Reference 14.1-3), and the detailed core and hot assembly flow and enthalpy distributions are calculated using the SPC XCOBRA-IllC code (Reference 14.1-4). The SPC XNB correlation was utilized to calculate MDNBR.

14.1.5.1.5.1 Analysis of Results The ANF-RELAP analysis provides the NSSS boundary conditions for the XTGPWR and the XCOBRA-IllC calculations. This section presents a description of the treatment of factors which can have a significant im~ pact on NSSS response and i

resultant MDNBR an'd LHGR values. The plant specific parameters used in this analysis are listed in Tables 14.1.5.1 -3 to 14.1.5.1 -5.

Conservatisms are included in parameters or factors known to have significant effects on the NSSS performance and resulting MDNBR and LHGR values.

MNPS-2 FSAR 14.1.5.1.5.1.1 Break Location, Size, and Flow Model l

The pre-scram SLB event analyzes breaks outside containment both downstream j

(symmetric cases) and upstream (asymmetric cases) of the main steam line check valves and breaks inside containment (asymmetric cases). A full range of break sizes, up to the double-ended guillotine break of a main steam line, were considered.

The ANF-RELAP break mass flow rate is computed using the Moody critical flow l

model modified such that only steam flows out the break.

14.1.5.1.5.1.2 Power Decalibration Power decalibration is caused by density-induced changes in the reactor vessel downcomer shadowing of the power-range ex-core detectors during heatup or cooldown transients. The nuclear power levels indicated by those instruments are lower than the actual reactor power levels when the coolant entering the reactor vessel is cooler than the normal temperature for full-power operation (and higher when the vessel inlet coolant is warmer than the normal full-power temperature).

This effect is included in the modeling of any power-dependent reactor trips credited in the analysis of full-power cooldown events and low-power events. The Variable i

Overpower trip, the Thermal Margin / Low Pressure (TM/LP) trip function, and the Local Power Density (LPD) trip all depend on the indicated nuclear power level.

l 14.1.5.1.5.1.3 Harsh Containment Conditions Harsh containment conditions can be caused by the release of steam within the reactor containment. Under such conditions, only those trips which have been qualified for harsh environments are credited, and increased uncertainties are included in the setpoints of all environmentally qualified trips which are credited, i

14.1.5.1.5.1.4 Boron Injection Boron injection into the primary system acts to mitigate the return to power.

Injection of boron is modeled from the HPSI system. The HPSI system is conservatively modeled to take suction from the Rcfueling Water Storage Tank (RWST) at 35'F with a boron concentration of 1720 ppm. Initially, the line volume between the check valves isolating the system pumps and the cold leg injection location is assumed to be filled with unborated water. The time required to flush this l

unborated water from the safety injection lines is included as an integral part of the l

ANF-RELAP NSSS calculation. In the pre-scram SLB event, the analysis is terminated shortly after reactor trip, therefore injection of borated water is not a factor in the enalysis.

14.1.5.1.5.1.5 Single Failure Assumption

I l

MNPS-2 FSAR l

In order to simulate the asymmetric thermal hydraulic and reactivity feedback effects l

that occur during the pre-scram SLB event, the core is divided into an affected sector (1/2 of the core) and an unaffected sector (1/2 of the core). The single failure assumed in this analysis is the loss of one channel of Nuclear Instrumentation (NI) which provides power indication to the Reactor Protection System (RPS). If one l

channel is out of service, the three remaining Ni safety channels will be in a 2-out-of-3 coincidence mode to cause a reactor trip. The excore detectors are placed around the reactor vessel is positions that result in one detector seeing the flux only from the l

affected region, one seeing the flux only from the unaffected region, and two detectors seeing nearly equal flux from both regions. If one of these latter two is out of service, and the other is assumed to be a single failure, the remaining two channels will be required to cause an RPS trip (high power or TM/LP). Since the power in the affected region will always be higher than in the unaffected region, it is sufficient to model the NI channel reading the unaffected region only.

14.1.5.1.5.1.6 Feedwater Normal MFW flow is assumed to be delivered to both SGs. The MFW flow increases as the secondary pressure decreases at the lowest possible fluid temperature until the feedwater regulator valve closes. Fluid temperature is determined by assuming heating of the feedwater ceases at the same time the break is initiated. The MFW flow is terminated 14 seconds after receiving the isolation signal.

14.1.5.1.5.1.7 Trips and Delays Actuation signals and delays are given in Table 14.1.5.1-4. Biases to account for uncertainties are included in the trip setpoints as shown. In the pre-scram SLB event, the analysis is terminated shortly after reactor trip, therefore injection of

)

borated water is not a factor in the analysis.

j 14.1.5.1.5.1.8 Neutronics The core kinetics input for this calculation consisted of the minimum required control rod shutdown worth at EOC, and EOC values associated with the reactivity feedback curves, delayed neutron fraction, delayed neutron fraction distribution and related time constants, and prompt neutron generation time. The ANF-RELAP default fission product and actinide decay constants were utilized for this calculation.

1 1

l The core reactivity is derived from input of several functions. These include effects l

from control rod worth, moderator density changes, boron concentration, and Doppler effects. The reactivity is weighted between the core sectors. The ANF-RELAP analyses for cases with offsite power available were performed with an MTC of -28 pcm/ F. The ANF-RELAP analyses for cases with a loss of offsite power were performed with an MTC of +4.0 pcm/ F. A summary of the nuclear input and assumptions is given in Table 14.1.5.1-5.

1 l

MNPS 2 FSAR 14-1.5.5.1.9 Decay Heat The presence of radioisotope decay heat at the initiation of the SLB event will reduce the rate and the extent of cooldown of the primary system. The initial decay heat is calculated on the basis of infinite irradiation time at a power of 2754 MW prior to transient initiation. This treatment of decay heat serves to maximize the stored

- energy and provide limiting stored energy conditions for the SLB cases.

14.1.5.1.5.1.10 Nodalization The NSSS transient calculations utilized the nodalization model described in

' Reference 14.1-1. The nodalization treats all major NSSS components and subcomponents as discrete elements, with the exception of the secondary side of the steam generators, in addition, all components with long axial dimensions are divlded into subcells adequate to minimize numerical diffusion and smearing of gradients.

in order to simulate the asymmetric thermal-hydraulic and reactivity feedback effects i

that occur during the pre-scram SLB event, the core is divided into an affected sector (1/2 of the core) and an unaffected sector (1/2 of the core).

14.1.5.1.5.1.11 Interloop Mixing During an actual SLB transient, some mixing between the parallel channels within the reactor pressure vessel will occur in the downcomer, the lower plenum, the core, and the upper plenum due to lateral momentum imbalances, and turbulence or eddy mixing. The mixing will act to reduce the positive reactivity feedback effects due to a reduced rate and magnitude of cooldown of the affected loop and associated core sector.

In this analysis, no credit is taken for turbulent or eddy mixing of coolant between loops or the parallel flow channels within the reactor pressure vessel. However, interloop mixing is calculated to occur due to flow in interloop junctions in the upper and lower plenums. Mixing in the lower plenum was effectively reduced to zero by using an extremely high loss coefficient between the affected and intact sectors.

14.1.5.1.5.2 Minimum Departure From Nucleate Boiling Ratio and Linear Heat Generation Rate Analysis The XTGPWR (Reference 14.1-3) cora neutronics code is used to calculate the core radial power distributions for XCOBRA-IllC (Reference 14.1-4) during the asymmetric transients with offsite power available only. The XTGPWR model is a three-dimensional representation of the entire core, with four radial nodes and 24 axial nodes for each fuel assembly.

i

. Based on the overall core conditions calculated by ANF-RELAP for the symmetric cases (or ANF-RELAP and XTGPWR for the asymmetric cases with offsite power available) at the peak heat flux time-point, the XCOBRA-lllC fuel assembly thermal-

. ~..

MNPS-2 FSAR hydraulic code is used to calculate the flow and enthalpy distributions for the entire core and the DNB performance for the DNB-limiting assembly. The XCOBRA-IllC model consists of a thermal-hydraulic model of the core (representing each assembly by a single " channel") linked to a detailed thermal-hydraulic model of the limiting assembly (representing each subchannel by a single " channel"). The limiting assembly DNBR calculations are performed using the XNB DNB correlation (Reference 14.1 -4).

For the asymmetric transients, the radial power peaking is augmented above the Technical Specification limit to account for the increase in radial power peaking which occurs during the transient. ' The increase in peaking is determined by XTGPWR.

14.1.5.1.6 Analysis Results A summary of calculated results important to this analysis is presented in Table 14.1.5.1-6 for the limiting MDNBR and LHGR cases. The MDNBR values are listed together with the corresponding core power values at the time of MDNBR which corresponds to the maximum power level. For cases where offsite power was available for operation of the primary coolant system pumps, the MDNBR and the maximum LHGR occurred at the time of the maximum power condition. For cases -

where offsite power is lost and the primary system pumps coast down, the maximum LHGR and the MDNBR occur when the worst combination of core power, flow, inlet temperature, and pressure are present. These conditions occurred at the time of peak power in this analysis.

The scenario which resulted in the highest power level and the largest LHGR is the 2

HFP 3.50 ft symmetric break outside containment with offsite power available for-operation of the primary coolant pumps. This case is presented in detail.

j The scenario which resulted in the limiting MDNBR is the HFP case with a loss of offsite power and is also presented in detail.

2 14.1.5.1.6.1 - Hot Full Power 3.50 ft Break Outside Containment and Downstream of a Check Valve with Offsite Power Available The ANF-RELAP simulation of the NSSS during the HFP symmetric break transient with offsite power available is illustrated in Figures 14.1.5.1 through 14.1.5.1-6.

A tabulation of the sequence of events is presented in Table 14.1.5.1-7. The ANF-RELAP computation was terminated 60 seconds after break initiation. This is well beyond the time of MDNBR or peak LHGR. The general response of the reactor was the same for all the symmetric break sizes but the occurrence of events was delayed as the break size decreased.

14.1.5.1.6.1.1 Secondary System Parameters

MNPS-2 FSAR l

Upon break initiation the break flow increased sharply and then began to decline in l

response to falling secondary side pressure. When the turbine trip occurred, the break flow increased due to a local pressure increase. The main steam line flow rate from each generator initially increased (see Figure 14.1.5.1-6) in response to the break and the assumed instantaneous full opening of the turbine control valves. The increased steam flow creates a mismatch between the core heat generation rate and the steam generator heat removal rate. This power mismatch causes the primary-to-secondary heat transfer rate to increase, which in turn causes the primary system to cool down (see Figure 14.1.5.1 -2). When the reactor scram occurred, the turbine valves closed and steam flow declined sharply. At this point, the MFW flow may exceed the steam flow as the control system ettempts to restore steam generator mass. Both steam flow and MFW flow were terminated when the main steam 1

isolation valves closed.

14.1.5.1.6.1.2 Primary System Parameters 1

Approximately five seconds after the break occurred, the core inlet temperature began to decline. With a negative MTC (see Figure 14.1. 5.1 -3), the primary system cooldown caused the reactor power level to increase. The core power continues to increase until reactor scram on low steam generator pressure occurs. This terminated the power excursion. The pressurizer pressure and level began to decline as the volume of water in the primary system shrank. The core inlet mass flow rate increased due to the increasing density of the primary system fluid while the reactor coolant pumps' speed remained constant.

14.1.5.1.6.1.3 Departure From Nucleate Boiling Ratio and Linear Heat Generdtion Rate Results The MDNBR value for this scenario was calculated to be 1.298 which is above the 95/95 XNB correlation limit. Therefore, no fuel rods would be expected to fait during this transient scenario from an MDNBR stand point.

2 The peak LHR for the LHR-limiting case (3.50 ft break outside containment and downstream of a check valve) is calculated to be 19.7 kW/ft. Comparing this LHGR value with a centerline melt criteria of 21 kW/ft, it is apparent that centerline melt is not predicted to occur. Thus, no fuel failures are predicted to occur due to violation of the centerline melt criteria.

2 14.1.5.1.6.0.

Hot Full Power 3.51 ft Inside Containment Asymmetric Break Concurrent with a Loss of Offsite Power l

l The ANF-RELAP NSSS simulation of the most limiting pre-scram SLB scenario from i

an MDNBR standpoint (i.e., HFP 3.51 ft* inside containment asymmetric break concurrent with a loss of offsite power) is illustrated in Figures 14.1.5.1-7 through 14.1.5.1-11. A tabulation of the sequence of events is presented in Table 14.1.5.1-

8. The ANF-RELAP computation was terminated 60 seconds af ter break initiaticn.

This is well beyond the time of MDNBR or peak LHGR.

MNPS-2 FSAR The transient is initiated by the opening of the break. The RCPs tripped shortly after transient initiation. The sharp reduction in the reactor coolant flow causes this pre-trip pumps off calculation to become a heat up transient very similar to a Loss of Coolant Flow event. Typically, the Steam Line Break calculation is a cooldown event. Because this case is a heat up event the most positive BOC neutronics conditions are used, and the maximum inside containment asymmetric break size is used. The maximum break size causes the biggest decrease in primary pressure.

Maximizing the primary system pressure decrease causes the maximum decrease in moderator density and the maximum positive moderator feedback. The RCP trip causes the RCS flow to decrease rapidly throughout this transient. The decreasing RCS flow causes the transient time of the fluid in the core to increase and the fluid temperature begins to rise. The increasing fluid temperature causes positive moderator feedback, which in turn causes an increase in core power. However, the decreasing RCS flow causes the heat transfer to the fluid to decrease. The increase in core power is offset by the decrease in h' eat transfer from the fuel rods, such that, the fuel rod heat flux decreases slightly until reactor scram. The reactor scrams on the low reactor coolant flow trip signal.

14.1.5.1.6.2.3 Departure From Nucleate Boiling Ratio and Linear Heat Generation Rate Results 2

The MDNBR value for the pre-scram 3.51 ft asymmetric break inside containment with a loss of offsite power was calculated to be 0.88 which is below the 95/95 XNB correlation limit. The number of failed assemblies is determined by comparing the core power distribution to the assembly power where DNB occurs. This results in a predicted failure of 3.7% of the fuel rods in the core.

2 The peak LHR for this case is bounded by the 3.50 ft outside containment symmetric break. Therefore, the LHGR for this case is below the criteria of 21.0 kW/ft and no fuel failures are predicted to occur due to violation of the centerline melt criteria.

14.1.5.1.7 Conclusions 2

The HFP 3.50 ft break outside containment and downstream of a check valve (symmetric break) with offsite power available was determined to be the most limiting in this analysis from an LHGR standpoint (19.7 kW/ft). In no scenario evaluated, however, was fuel failure calculated to occur as a result of violating the 21 kW/ft fuel centerline melt criteria.

2 The HFP 3.51 ft asymmetric break inside containment coincident with a loss of offsite power was determined to be the most limiting in this analysis from the standpoint of MDNBR. The MDNBR was calculated to be 0.88 which is below the 95/95 XNB correlation limit. This results in a predicted failure of 3.7% of the fuel rods in the core.

m. _ _

g MNPS-2 FSAR 14.1.5.2 Post-Scram Analysis' 14.1.5.2.1 Event Initiator l

This event is initiated by a rupture in the main steam piping downstream of ths L

integral steam generator flow restrictors and upstream of the MSIVs which recula in an uncontrolled steam release from the secondary system.

l l

14.1.5.2.2 Event Description The increase in energy removal through the secondary system results in a severe overcooling of the primary system. In the presence of a negative MTC, this j

cooldown causes a decrease in the shutdown margin (following reactor scram) such that a return to power might be possible following a steam line rupture. This is a potential problem because of the high power peaking factors which exist, assuming the most reactive control rod to be stuck in its fully withdrawn position.

14.1.5.2.3 Reactor Protection Reactor protection is provided by the low steam generator pressure and water level trips, variable overpower trip, LPD trip, TM/LP trip, high containment pressure trip, l

and SIAS. Reactor protection for the Steam System Piping Failures inside and i

Outside of Containment event is summarized in Table 14.1.5.2-1.

l 14.1.5.2.4 Disposition and Justification At rated power conditions, the stored energy in the primary coolant is maximized, the available thermal margin is minimized, and the pre-trip power level is maximized.

These conditions result in the greatest potential for cooldown and provide the greatest challenge to the SAFDLS. Initiating this event from rated power also results in the highest post-trip power since it maximizes the concentration of delayed neutrons providing for the greatest power rise for a given positive reactivity insertion.

Additional thermal margin is also provided at lower power levels by the automatically j

decreasing setpoint of the variable overpower trip. Thus, this event initiated from rated power conditions will bound all other cases initiated from at power operation i

modes.

For the zero power and subcritical plant states (Modes 2-6), there is a potential for a l

return-to-power at reduced pressure conditions. The most limiting steam line break (SLB) event at zero power is one which is initiated at the highest temperature, thereby providing th'e greatest capacity for cooldown. This occurs in Modes 2 and 3.

l Thus, the event initiated from Modes 2 and 3 will bound those initiated from Modes 4-6. Further, the limiting initial conditions will occur when the core is just critical.

These conditions will maximize the available positive reactivity and produce the quickest and largest return to power. Thus, the SLB initiated from critical conditions l-.

g m.w

--r--

--e--

r--m-y vrwr sus

MNPS-2 FSAR in Mode 2 will bound the results of the event initiated form suberitical Mode 3 conditions.

The technical specifications only require a minimum of one RCP to be operating in Mode 3. One pump operation provides the limiting minimum initial core flow case.

Minimizing core flow minimizes the clad to coolant heat transfer coefficient and degrades the ability to remove heat generated within the fuel pins. Conversely, however, a maximum loop flow will maximize the primary to secondary heat transfer coefficient, thus providing for the greatest cooldown. Higher loop flow will sweep the cooler fluid into the core faster, maximizing the rate of positive reactivity addition and the peak power level.

l The worst combination of conditions is achieved for the four pump loss of offsite power case. In this situation, the initial loop flow is maximized resulting in the greatest initial cooldown, while the final loop flow is minimized providing the greatest challenge to the DNB SAFDL. Since the natural circulation flow which is established at the end of the transient will be the same regardless of whether one or four pumps were initially operating the results of the four pump loss of offsite power case will bcund those of the one pump case. Thus, only four pump operation need be -

analyzed for the Mode 2 case.

The event is analyzed to support the technical specification EOC MTC limit..This event must be analyzed both with and without a coincident loss-of-offsite power.

Typically there are two single failures which are considered for the offsite power available case. The first is failure of a High Pressure Safety injection (HPSI) pump to start. The second is failure of an MSIV to close, resulting in a continued uncontrolled cooldown. However, Millstone 2 has combination MSIV/ swing disc check valves. A double valve failure would thus be required for steam from the intact steam generator to reach the break. This is not deemed credible. Thus, the single failure to be.

considered with offsite power avai!able is failure of a HPSI pump to start. For the loss-of-offsite power case, the limiting single failure is the failure of a diesel generator to start. This is assumed to result in the loss of one HPSI pump. The disposition of events for the Steam System Piping Failures inside and Outside of Containment event is summarized in Table 14.1.5.2-2.

14.1.5.2.5 Definition of Events Analyzed

. The post-scram SLB is initiated by a rupture in the main steam piping downstream of the integral steam generator flow restrictors and upstream of the MSIVs which results in an uncontrolled steam release from the secondary system. The effects of harsh containment conditions (where applicable) are included in the following analyses:

1.

HFP and HZP breaks outside containment with offsite power available 2.

HFP and HZP breaks outside containment with a loss of offsite power 3.

HFP and HZP breaks inside containment with offsite power available 4.

HFP and HZP breaks inside containment with a loss of offsite power The event is analyzed to support the technical specification EOC MTC limit. This event must be analyzed both with and without a coincident loss-of-offsite power.

MNPS-2 FSAR The single failure assumed in this analysis results in the disabling of one of the two HPSI pumps required to be in service during normal operation, in addition to the single failure, there is no credit taken for the charging pump system. This assumption results in an additional delay in the time required for boron to reach the core. The delay is amplified when combined with the assumption of a stagnant upper head which serves to maintain the primary system pressure due to flashing of the hot fluid in the upper head.

The increase in energy removal through the secondary system results in a severe overcooling of the primary system, in the presence of a negative MTC, this cooldown results in a large decrease in the shutdown margin and a return to power.

i This return to power is exacerbated because of the high power peaking factors which exist, with the most reactive control rod stuck in its full withdrawn position.

As outlined in Reference 14.1-1, three computerized calculations are required prior to the final calculation of the Minimum Departure From Nucleate Boiling Ratio (MDNBR) values and the maximum Linear Heat Generation Rate (LHGR) values utilized in the determination of fuel failure. The NSSS response is computed using the Siemens Power Corporation (SPC) ANF-RELAP code (Reference 14.1-2), the detailed core and hot assembly power distributions and the reactivity at the time of peak post-scram power are calculated using the SPC XTGPWR code (Reference 14.1-3), and the detailed core and hot assembly flow and enthalpy distributions are calculated using the SPC XCOBRA-lllC code (Reference 14.1-4). The modified Barnett correlation was utilized to calculate MDNBR due to the reduced pressures occurring during the SLB event.

14.1.5.2.5.1 Analysis of Results The ANF-RELAP analysis provides the NSSS boundary conditions for the XTGPWR and the XCOBRA-IllC calculations. This section presents a description of the treatment of factors which can have a significant impact on NSSS response and resultant MDNBR and LHGR values. The plant specific parameters used in this analysis are listed in Tables 14.1.5.2-3 to 14.1.5.2-5. Conservatisms are included in parameters or factors known to have significant effects on the NSSS performance and resulting MDNBR and LHGR values.

14.1.5.2.5.1.1 Break Location, Size, and Flow Model The post-scram SLB eveM is initiated by a double ended guillotine break of a main steam line downstream m :he integral steam generator flow restrictors and upstream of the MSIVs. The flow is choked at the integral steam generator flow restrictor, 2

which has an area of 3.51 ft. On the steam generator side of the break, steam flows out of the break throughout the entire transient. On the MSIV side of the break, break flow terminates after the MSIVs are fully closed. As an added conservatism, the main steam check valves are not credited in the analysis. The event occurs concurrent with the most reactive control rod stuck out of the core.

The break flow areas for the affected and intact steam generators are listed in Table 14.1.5.2-3. These areas correspond to the locations in the flow path where choked flow will occur.

MNPS-2 FSAR The ANF-RELAP break mass flow rate is computed using the Moody critical flow model modified such that only steam flows out the break.

14.1.5.2.5.1.2 Boron Injection Boron injection into the primary system acts to mitigate the return to power.

injection of boron is modeled from the HPSI system. The HPSI system is conservatively modeled to take suction from the Refueling Water Storage Tank (RWST) at 35 F with a boron concentration of 1720 ppm. Initially, the line volume between the check valves isolating the system pumps and the cold leg injection location is assumed to be filled with unborated water. The time required to flush this i

unborated water from the safety injection lines ic included as an integral part of the ANF-RELAP NSSS calculation. The characteristics of the HPSI system are listed in Table 14.1.5.2-3. The delivery curve for the HPSI system used in this analysis is given in Figure 14.1.5.2-1.

14.1.5.2.5.1.3 Single Failure Assumption The single failure assumed in the engineered safeguards systern results in the disabling of one of the two HPSI pumps required to be in service during normal operation in addition to the single failure, there is no credit taken for the charging pump system. This assumption results in an additional delay in the time required for boron to reach the reactor core. The delay is further amplified when combined with the assumption of a stagnant upper head which serves to maintain the primary system pressure due to flashing of the hot fluid in the upper head.

14.1.5.2.5.1.4 Feedwater For the HFP scenarios, normal MFW flow is assumed to be delivered to both SGs.

The MFW flow increases as the secondary pressure decreases at the lowest possible fluid temperature until the feedwater regulating valve closes. Fluid temperature is determined by assuming heating of the feedwater ' ceases at the same time the break is initiated. The MFW flow is terminated 14 seconds after receiving the isolation signal.

For the HFP scenarios, the AFW flow is assumed to be zero at break initiation. After 180 seconds, AFW is delivered at the maximum capacity of the AFW system with flow restrictors installed on the AFW delivery lines. For the HZP scenarios, the AFW flow is increased to the maximum capacity immediately at break initiation. For all scenarios, all of the AFW flow is directed to the affected steam generator to maximize the cooldqwn rate. The operator is assumed to terminate the AFW flow to the affected steam generator at 600 seconds.

14.1.5.2.5.1.5 Trips and Delays p

p

,.m-.9 y

c_-

g

MNPS 2 FSAR Trips for the HPSI, main feedwater valves, and MSIVs are given in Table 14.1.5.2-4.

Biases to account for uncertainties are included in the trip setpoints as shown. For the steam and feedwater valves, the delay times given are between the time the trip setpoint is reached and the time full valve closure is reached. For the HPSI system,

the delay time given is from the time the setpoint is reached until the pumps have accelerated to rated speed. Additional delay time required to sweep the lines of unborated water is accounted for by setting the boron concentration of the injected 4

flow to zero until the volume of the injection lines has been cleared.

14.1.5.2.5.1.6 Neutronics The core kinetics input for this calculation consisted of the minimum required control rod shutdown worth at the EOC, and EOC values associated with the reactivity feedback curves, delayed neutron fraction, delayed neutron fraction distribution and related time constants, and prompt neutron generation time. The ANF-RELAP default fission product and actinide decay constants were utilized for this calculation.

The core reactivity is derived from input of several functions. These include effects from control rod worth, moderator density changes, boron concentration, and Doppler effects. The reactivity is weighted between the core sectors. Different reactivity functions were utilized where necessary for the HZP and the HFP cases.

The ANF-RELAP analyses were performed with an MTC of -28 pcm/ F. A summary of the nuclear input and assumptions is given in Table 14.1.5.2-5.

14.1.5.2.5.1.7 Decay Heat

.The presence of radioisotope decay heat at the initiation of the SLB event will reduce the rate and the extent of cooldown of the primary system. For the HFP scenarios, the initial decay heat is calculated on the basis of infinite irradiation time at a power of 2700 MW prior to transient initiation. For the HZP scenarios, the initial decay heat is calculated on the basis of infinite irradiation time at a power of 1 W prior to transient initiation. For both scenarios, decay heat generated from return to power is calculated..This treatment of decay heat serves to maximize the stored energy in the HFP cases and to minimize it in the HZP cases. This treatment provides limiting stored energy conditions for the SLB cases.

14.1.5.2.5.1.8 Nodalization The NSSS transient calculations utilized the nodalization model described in Reference 14.1-1. The nodalization treats all major NSSS components and subcomponents as discrete etements, with the exception of the secondary side of the steam generators. In addition, all components with long axial dimensions are divided into subcells adequate to minimize numerical diffusion and smearing of gradients.

In order to simulate the asymmetric thermal hydraulic and reactivity feedback effects that occur during an SLB transient, the core is nodalized into three radial sectors.

One sector corresponds to the region immediately surrounding the assembly where

~..

i MNPS-2 FSAR the most reactive control rod is assumed stuck out of the core. This sector is termed the ' stuck rod' sector. The remainder of the region of the core which is directly affected by the loop containing the break is the second sector and is termed the

'affected' sector. The remainder of the core and the other loop is termed either the

' unaffected' or the ' intact' sector or loop.

l t

l' I

14.1.5.2.5.1.9 Interloop Mixing l

During an actual SLB transient, some mixing between the parallel channels within the l

reactor pressure vessel will occur in the downcomer, the lower plenum, the core, and I

the upper plenum due to lateral momentum imbalances, and turbulence or eddy mixing. The mixing will act to reduce the positive reactivity feedback effects due to a reduced rate and magnitude of cooldown of the affected loop and associated core sector.

[

In this analysis, no credit is taken for turbulent or eddy mixing of coolant between loops or the parallel flow channels within the reactor pressure vessel (RPV).

However, interloop mixing is calculated to occur due to flow in interloop junctions in the upper and lower plenums. Mixing in the lower plenum was reduced to a l

minimum by using an extremely high loss coefficient between the affected and intact

)

sectors.

l 14.1.5.2.5.1.10 Harsh Containment Conditions Harsh containment conditions can be caused by the release of steam within the reactor containment. Under such conditions, only those trips which have been' qualified for harsh environments are credited, and increased uncertainties are included in the setpoints of all environmentally qualified trips which are credited.

l 14.1.5.2.5.2 Minimum Departure From Nucleate Boiling Ratio and Linear Heat Generation Rate Analysis MDNBR calculations require determination of the power, enthalpy, and flow distributions within the highest power assembly of the stuck rod core sector.

Similarly, determination of the maximum LHGR also requires characterization of the power distribution. The power distribution within the core, including the highest powered assembly within the stuck rod core sector, is calculated with XTGPWR (Reference 14.1-3). Flow and enthalpy distributions within the core, including the l

highest powered assembly within the stuck rod core sector, are calculated with

- XCOBRA-IllC (Reference 14.1-4). In order to obtain compatible flows, moderator densities, and powers within the high power assemblies, iteration between XTGPWR

- and XCOBRA-lllC is conducted.

For this calculation, the modified Barnett correlation was found to be suitable for the l

MDNBR calculation. The modified Barnett correlation is based upon closed channels and primarily uniform power distribution data. The correlation is based on assembly inlet (or upstream) fluid conditions rather than on local fluid conditions as is the case l

l

. ~. _-.. - - -

1 I

MNPS-2 FSAR with subchannel based correlations. Use of the correlation is limited to the range of the data base unless conservative extrapolations can be made.

14.1.5.2.6 Analysis Results A summary of calculated results important to this analysis is presented in Table 14.1.5.2-6 for the limiting MDNBR and LHGR scenarios. The MDNBR values are i

listed together with the corresponding core power values at the time of MDNBR which corresponds to the maximum post-scram power level. The outside containment cases, regardless of whether or not offsite power was or was not available, were found to be the most limiting. For cases where offsite power was available for operation of the primary coolant system pumps, the MDNBR and the maximum LHGR occurred at the time of the maximum power condition. For cases 1

where offsite power is lost and the primary system pumps coast down, the maximum LHGR and the MDNBR occur when the worst combination of core power, flow, inlet temperature, and pressure are present. These conditions occurred at the time of peak power in this analysis.

l l

The scenario which resulted in the highest post-scram power level and the largest l

LHGR is that initiated from HFP with the break occurring outside containment and l

with offsite power available for operation of the primary coolan't pumps. This case is presented in detail.

The NSSS responses for the scenarios with loss of offsite power for operation of the primary system coolant pumps are different from those scenarios where offsite power is available throughout the transient due to the pump coastdown and-subsequent natural circulation of the primary coolant. Post-scram maximum 'p,ower levels attained during the transient are significantly lower. Lower power levels result from lower positive moderator feedback. The positive moderator feedback is reduced due to the coolant density reductions that occur axially upwards in the core at low l

core flow rates, even for low core power levels. Lower power levels cause MDNBR values to increase, but lowering flow rates cause MDNBR values to decrease.

Overall, the combination of factors results in lower MDNBR values for the reduced flow condition than for the full flow condition.

l Of the two loss of offsite power scenarios analyze'd, the HFP break occurring outside containment case resulted in lower MDNBR values. The general response of the HFP and HZP cases with loss of offsite power is comparable. Because the two scenarios are quite similar in terms of their general response, only the limiting MDNBR case (i.e., HFP break outside containment and without offsite power) is presented in detail, i

l 14.1.5.2.6.1 Hot Full Power Outside Containment with Offsite Power Available The ANF-RELAP simulation of the NSSS during the HFP transient with offsite power available is illustrated in Figures 14.1.5.2-2 through 14.1.5.2 9. A tabulation of the I

sequence of events is presented in Table 14.1.5.2-7. The ANF-RELAP computation l

was terminated 600 seconds after break initiation. This is well beyond the time of

._ ~

MNPS-2 FSAR MDNBR or peak LHGR AFW termination of the AFW by manual operator action was assumed to occur 600 seconds af ter initiation of the break.

14.1.5.2.6.1.1 Secondary System Thermal Hydraulic Parameters Steam flow out the break is the source of the NSSS cooldown. Break flow for the steam generators is plotted in Figure 14.1.5.2-2. Secondary pressure for the steam generators is plotted in Figure 14.1.5.2-3. After break initiation, the pressure in the affected steam generator decreased immediately and then stabilized around 180 seconds. The mass inventory in both steam generators decreased throughout the transient. The relatively high reactor power level caused the affected steam generator to dry out by 490 seconds. The affected steam generator drying out caused the primary-to-secondary heat transfer to deteriorate. As a result, the i

primary system temperature rose, the secondary side pressure decreased, and, since the break flow is determined by the secondary system pressure, the break flow also declined. The heatup of the primary coolant reduced the reactivity present and power dropped rapidly.

The intact steam generator blows down for a short period until the MSIVs completely close approximately 17 seconds after the break is initiated. The pressure recovers as the intact steam generator equilibrates with the primary system' and then slowly increases as the primary system begins to heat up.

14.1.5.2.6.1.2 Primary System Thermal Hydraulic Parameters The primary system coolant temperature and pressure responses resulting from'the break flow are illustrated in Figures 14.1.5.2-4 through 14.1.5.2-6. The primary system pressure decays rapidly as the coolant contract:: due to cooldown and the pressurizer empties. The MSIVs close at '17 seconds, ending the blowdown of the intact steam generators and reducing the rate of energy removal from the primary fluid. The pressurizer emptied at approximately 60 seconds and system pressure (which increased slowly for the duration of the transient) was thereafter established by the saturation temperature of the primary coolant in the upper head of the reactor vessel.

14.1.5.2.6.1.3 Reactivity and Core Power The reactivity transient calculated by ANF-RELAP is illustrated in Figure 14.1.5.2-8.

Initially, the core is assumed to be at full power. All control rods, except the most reactive one, are assumed to be inserted into the core following the reactor trip signal. The reactivity transient then proceeds. The total core reactivity, initially at 0.00$, decreased instantly due to the scram worth at reactor trip, but then steadily increased due to moderator and Doppler feedback associated with the primary system cooldown. Shortly thereaf ter, power begins to rise steadily due to the dominating positive reactivity feedback from the moderator. The reactor soon achieves a quasi-steady-state power level where the Doppler and the moderator reactivities balance the scram reactivity.

MNPS-2 FSAR Fifty-five seconds after break initiation, the RCS pressure dropped below the shutoff head of the HPSI system and HPSI flow to the RCS began. But, the elevated primary pressure limited the delivery of boron into the core due to the pressure versus flow characteristics of the HPSI system and unborated water never cleared the safety injection lines during the transient.

Figure 14.1.5.2-9 shows the transient reactor power. The reactor power initially declined due to insertion of the control rods. The severe cooldown resulted in power increasing after 52 seconds. A quasi steady-state reactor power 'evel was established by 260 seconds and a maximum power level of 378 MW or 14% of rated power occurred at 462 seconds.

14.1.5.2.6.1.4 XTGPWR and XCOBRA-IllC Results The XTGPWR calculation is made initially on the basis of ANF-RELAP input. Each assembly within the three channels is assumed to have a uniform flow corresponding to the sector flows calculated with ANF-RELAP. Due to high power peaking in the region of the stuck control rod, large moderator density reductions are calculated to occur in the top portions of several assemblies in this region of the core in the XTGPWR calculation. This moderator density decrease is a major factor in the flattening of the axial and radial profiles, and the significant reduction in reactivity observed when XTGPWR is compared to ANF-RELAP. An XCOBRA-IllC analysis is also conducted to define the flow and enthalpy distribution within the high power assembly.

The ANF-RELAP reactivity and power calculation has considerable inherent conservatism. To demonstrate this, a comparison of the change in reactivity at the maximum LHGR time is made. A comparison of the overall change in reactivity between ANF-RELAP and XTGPWR shows that ANF-RELAP conservatively underestimates the negative reactivity by 1.01 $ at the time of maximum LHGR thus indicating that the ANF-RELAP power calculation is conservative.

14.1.5.2.6.1.5 Departure From Nucleate Boiling Ratio and Linear Heat Generation Rate Results For the MDNBR portion of the calculation, the radial power distribution was modified to conservatively account for local rod power distribution affects within the hot assembly. This was done by raising the power of the hot assembly to bound the peak rod power.

On the bases of these conservative assumptions, the MDNBR value was calculated to be 2.28. This compares to a 05/95 DNBR limit of 1.135 for the modified Barnett correlation.

f Therefore, no fuel rods would be expected to fait during this transient scenario from an MDNBR stand point.

i The analysis of the peak LHGR also comes from the XTGPWR and XCOBRA-IllC analysis. The peak LHGR is calculated from the ANF-RELAP total core power and

l l

l MNPS-2 FSAR l

the XTGPWR radial and axial peaking. The peak LHGR, 24.27 kW/ft, was l

calculated for the HFP outside containment break with offsite power available event.

l When compared to a centerline melt criteria of 21.0 kW/ft, four assembly quadrants (one full assembly) or O.46% of the core, are predicted to fail due to violation of the centerline melt criteria.

14.1.5.2.6.2 Hot Full Power Outside Containment with Loss of Offsite Power The ANF-RELAP NSSS simulation of the most limiting SLB scenario from an MDNBR standpoint (i.e., HFP outside containment break with a loss of offsite power) is illustrated in Figures 14.1.5.2-10 through 14.1.5.2-16. A tabulation of the sequence of events is presented in Table 14.1.5.2-8. Termination of the AFW by manual operator action was assumed to occur 600 seconds after initiation of the break. This is well beyond the time of MDNBR and maximum LHGR. Termination of AFW would cause the affected SG to dry out and an increase in the primary system temperature. The increase in primary temperature, will drive the reactor subcritical and restore shutdown.

14.1.5.2.6.2.1 Secondary System Thermal Hydraulic Parame'ters Steam flow out the break is the source of the NSSS cooldown. Steam flow for the i

l affected steam generator is plotted in Figure 14.1.5.2-10. Secondary pressure for the steam generators is plotted in Figure 14.1.5.2-11. The affected steam generator blows down through the break throughout the transient. The pressure and mass flow rate dropped rapidly at first and then proceeded downward at a slower decay rate until natural circulation flow was established by approximately 250 seconds, i

The intact steam generators blow down for a short period until the MSIVs completely close approximately 16 seconds after the break is initiated. The pressure recovers as the intact steam generator equilibrates with the primary system. Subsequently, the intact steam generator pressure remains essentially constant as the primary intact coolant loop approaches natural circulation conditions.

14.1.5.2.6.2.2 Primary System Thermal Hydraulic Parameters The primary system core coolant temperatu e and pressure responses resulting from the break flow are illustrated in Figures 14.1.5.2-12 through 14.1.5.2-14. The primary system pressure decays rapidly as the coolant contracts due to the cooldown and the pressurizer empties. Continued pressure reduction in the primary system causes the relatively hot stagnant liquid in the head of the RPV vessel to flash. The flashing in the upper. head, coupled with near equilibration of other NSSS parameters, retards the pressure decay from that point forward.

A comparison of intact and affected core sector inlet temperatures throughout the l

transient indicates significant differences due to the limited cross flow allowed between loops. The core sector flows all show the same trend due to the 1

MNPS-2 FSAR coastdown of the primary coolant pumps. That is, all flows decrease rapidly until natural circulation conditions are achieved in the two flow loops.

14.1.5.2.6.2.3 Reactivity and Core Power The reactivity transient calculated by ANF-RELAP is illustrated in Figure 14.1.5.2 15.

Initially, the core is assumed to be at full power. All control rods, except the most reactive one, are assumed to be inserted into the core following the reactor trip signal. The reactivity transient then proceeds. The total core reactivity, initially at 0.00$, decreased instantly due to the scram worth at reactor trip, but then steadily increased due to moderator and Doppler feedback associated with the primary system cooldown. Shortly thereaf ter, power begins to rise steadily due to the dominating positive reactivity feedback frorn the moderator. The reactor soon achieves a quasi-steady-state power level where the Doppler and the moderator reactivities balance the scram reactivity.

Ninety seconds after break initiation, the RCS pressure dropped below the shutoff head of the HPSI system and HPSI flow to the RCS began. But, the elevated primary pressure limited the delivery of boron into the core due to the pressure versus. low characteristics of the HPSI system and unborated water never cleared the safety j

injection lines during the transient.

The transient experienced by the core power is illustrated in Figure 14.1.5.2-16. The reactor power declined to a decay heat level during the first 150 seconds of the transient. The maximum peak power level of 207 MW or 7.7% of rated power i

occurred at 488 seconds.

14.1.5.2.6.2.4 XTGPWR and XCOBRA-lllC Results The XTGPWR calculation is initially made on the basis of ANF-RELAP predicted core power, flow, pressure, and inlet temperatures. The XTGPWR calculations provide the radial and axial power distributions for use in the XCOBRA-IllC code. Due to the high power peaking in the region of the stuck control rod, and the low core average natural circulation flow rates, large moderator density decreases are calculated in several assemblies in this region in the XTGPWR calculation. This is a major f actor in the flattening of the axial and radial profiles, and the significant reduction in reactivity observed when XTGPWR is compared to ANF-RELAP. An XCOBRA-IllC analysis is also conducted to define the flow and enthalpy distribution within the high power assembly.

A comparison of the overall change in reactivity between ANF-RELAP and XTGPWR l

shows that ANF-RELAP conservatively underestimates the negative reactivity by 1.00$ at the time of MDNBR thus indicating that the ANF-RELAP power calculation i

is conservative.

14.1.5.2.6.2.5 Departure From Nucleate Boiling Ratio and Linear Heat Generation Rate Results l

l l

MNPS-2 FSAR The MDNBR of the hot fuel assembly is calculated to be 1.71 which is above the modified Barnett 95/95 DNBR correlation limit. Therefore, no fuel rods are expected to fail from an MDNBR standpoint.

As beforo, the analysis of the peak LHGR comes from the XTGPWR and the XCOBRA-IllC analysis. The peak LHGR was 17.96 kW/ft, Comparing this LHGR with a centerline melt criteria of 21 kW/ft, it is apparent that centerline melt is not predicted to occur. Thus, no fuel failures are predicted to occur due to violation of the centerline melt criteria.

14.1.5.2.7 Conclusions The HFP and HZP scenarios, with offsite power maintained for operation of the primary coolant pumps resulted in a return to higher power levels than the scenarios where offsite power is lost. However, these scenarios provide substantially greater margin to the MDNBR limit because of the higher coolant flow rate. In no scenario evaluated, however, was fuel failure calculated to occur as a result of penetration of the MDNBR safety limit. The HFP and HZP scenarios with offsite power maintained for operation of the primary coolant pumps returned to higher power levels than the scenarios where offsite power is lost. Even though these scenarios have substantially greater margin to the MDNBR limit because of a higher coolant flow rate, the higher power levels in combination with the highly skewed power distribution due to the assumed stuck rod cluster resulted in them having the least margin to the fuel centerline melt limit.

The HFP outside containment break scenario concurrent with a loss of offsite power was determined to be the most limiting in this analysis from an MDNBR standpoint.

The MDNBR of the hot fuel assembly is calculated to be 1.71 which is above the modified Barnett 95/95 DNBR correlation limit. Therefore, no fuel rods are expected to fail from an MDNBR standpoint.

The HFP outside containment break scenario with offsite power available was determined to be the most limiting in this analysis from the standpoint of centerline melt. This scenario results in the highest return to power and highest calculated LHGR of 24.27 kW/ft. When compared to a centerline melt criteria of 21.0 kW/ft, four assembly quadrants (one full assembly) or 0.46% of the core, are predicted to fail due to violation of the centerline melt criteria.

4 MNPS-2 FSAR 14.1.5.3 Radiological Consequences of a Main Steam Line Break The main steam line break is postulated to occur in a main steam line outside the containment. The radiological consequences of a main steam line break inside containment is bounded by the main steam line break outside containment. The plant is assumed to be operating with Technical Specification coolant concentrations and primary to secondary ler.kage. A 0.035 gpm primary to secondany leak is assumed to occur in both steam i

generators.

Two separate main steam line break cases are analyzed. In the first case, associated with this accident is that I fuel assembly experiences melting and releases the melted fuel into the RCS at the onset of the accident. One fuel assembly is equivalent to 0.46% melt. The l

activity associated with the melt condition is therefore available for release to the atmosphere via primary to secondary leakage. In the second case a pre-accident iodine spike is assumed to occur. In this case the primary coolant iodine concentrations are 60 times the plant technical specification activity level of I uCi/gm DE I-13 L In addition, the noble gas 7

activity in the primary coolant is assumed to be at technical specification levels.

The noble gases and iodines in the primary coolant that leak into the faulted steam generator during the transient are released directly to the environment without holdup or decontamination. An iodine partition factor of 0.01 is used for the releases from the unaffected steam generator. Off-site power is assumed to be lost, thus making the condenser unavailable. The-steam releases from the main steam line break are from the turbine building blowout panels as the atmospheric dispersion factor is greater for this release point than the enclosure building blowout panels. The steam releases from the intact steam generator are from the MSSVs/ADVs.

The radiological consequences of a main steam line break to the EAB, LPZ and Millstone 2 Control Room are reported in Tables 14.1.5.3-2 and 14.1.5.3-3. The assumptions used to perform this evaluation are summarized in Table 14.1.5.3-1.

The resulting dosn to the EAB and LPZ do not exceed the limits specified in 10CFR100.

The resulting dot.-s to the Control Room do not exceed the limits specified in GDC19.

t

MNPS-2 FSAR l

l REFERENCES 14.1-1 " Steam Line Break Methodology for PWRS," EMF-84-093(P), Revision 1, Siemens Power Corporation, Richland, WA, June 1998.

14.1-2 "ANF-RELAP Methodology for Pressurized Water Reactors: Analysis of Non-l LOCA Chapter 15 Events," ANF-89-151(P)(A), Advanced Nuclear Fuels Corporation, l

May 1992.

14.1-3 "XTG-A Two-Group Three Dimensional Reactor Simulator Utilizing Coarse l

Mesh Spacing (PWR Version)," XN-CC-28, Volume 1, Revision 5, Exxon Nuclear Company, Richland, WA 99352, July 1979.

14.1-4 "XCOBRA-lilC: A Computer Code to Determine the Distribution of Coolant During Steady-State and Transient Core Operation," XN-NF-75-21(A), Revision 2, Exxon Nuclear Company, January 1986.

\\

1 I

\\

l l

l

MNPS-2 FSAR

/Y.h S. /- /

g TABLE 14.*.5 1 i

AVAILABLE REACTOR PROTECTION FOR STEAM SYSTEM PIPING FAILURES INSIDE AND OUTSIDE OF CONTAINMENT-CVENT

\\

Gt2E-ScsMir, sostygzs Reactor Operatina Conditions Reactor Protection 1

Low Steam Generator Pressure Trip Low Steam Generator Water Level Trip low Reac/v Cos/a.,f flow Variable Overpower Trip Local Power Density Trip i

Thermal Margin / Low Pressure Trip High Containment Pressure Trip Safety injection Actuation Signal l

2

.)

Low Steam Generator Pressure Trip

+

Low Steam Generator Water Level Trip 1-ow 9mhs c%k./ Nm l

Variable Overpower Trip High Containment Pressure Trip Safety injection Actuation Signal l

l 3-6 Technical Specification Requirements on i

Shutdown Margin, inherent Negative Doppler Feedback l

I l

)

14S161.MP2 1 of 1 October 1994 {

l l

l l

MNPS-2 FSAR N./,d /- 2

)

TABLE 4 d '.5 2 DISPOSITION OF EVENTS FOR STEAM SYSTEM PIPING FAILURES INSIDE AND OUTSIDE OF CONTAINMENT EVENT l

ME -ScAdA1 AbdLYs%S Reactor Operatina Conditions Disposition 1

Analyze 2

Analyze

~

5lt0 l

3-6 Bounded by the above l

l r

I l

f i

(

)

1 i

14516 2.MP2 1 Of 1 October 1994)

MNPS-2 FSAR TABLE 14.1.5 3 1

ADVANCED NUCLEAR FUELS RELAP THERMAL-HYDRAULIC INPUT (STEAM LINE BREAK) linitial ondition Thes..i ! 'hd.sdie Inout HZE HEE Total Cor Power (watt) 1 2700'108 Primary Pres re (psia) 2250 2250 Core inlet' Tem ature (*F) 532 549 Primary Flow Rate m) 401,000 398,000 i

Pressurizer Level (% o pan) 40.0 -

65.0 i

Secondary Pressure (psia) 891 867 Secondary Temperature (*F) 531 527 Steam Flow Rate (Ib/s)- per Ste Generator 7

'1631 Feedwater Flow Rate (Ib/s)- per Steam Generator 7

1631 Feedwater Enthalpy (Bru/lb) 0.1 410.7 Secondary Fluid Mass (Ib) 223,000 143,000 Break Characteristics i

Minimum Flow Area Affected Steam Generator 6.31 ft2 Intact Steam Generator 2.35 ft' Location of Pine Break pstream of Affected eamline Flow Restrictor

'iniection Systems Total HPSI Pumps (2, normal,1 mounted spare) 3 3

Active HPSI Pumps 2

2 t

j Single Failure (No credit for mounted spare) 1 HPSI pump 1 HPSI pump Active Charging Pumps 1

1

/

1451&3.MP2 1 Of 2 October 1994

d MNPS-2 FSAR TABLE 14.1.5-3

' -)

3 ADVANCED NUCLEAR FUELS RELAP THERMAL-HYDRAULIC INPUT (STEAM LINE BREAK)

HZP life 4

Ref ling Water Storage. Tank Boron Concentration (ppm) 1720 1720 i

HPSI D ivery Curve Fig.14.1.5-1 Fig.14.1.5-1 1

Feedwater

-l Auxiliarv i

Flow, ma ' mum (Ibm /sec) 229.5 229.5 J

Temperatur (*F) 32.1 32.1 gg 5/to initial Flow / Steam enerator (Ibm /sec) 0.0 1631.1 Initial Temperature (*

N/A 432.1 u

i i

1 l

l J

14815 3.MP2 2 of 2 nnnhar 1004

MNPS-2 FSAR TABLE 14.1.5-4 ACTUATION SIGNALS AND DELAYS (STEAM LINE BREAK)

Analysis Paramet r Setnoints Setnoint Uncertainty Value i

1.

Low St m Line Pressure 500 psia

-22 psi 478 psia

\\

2.

Low Press izer Pressure 1600 psia

-22 psi 1578 psia MSIV Closure Reouired Actuatio sSional A. (1) Above Delav - 6.9 seconds HPSI Actuation

.Reouired Actuation Sional si[go A. (2) Above Delav - 30 seconds Main Feedwater Valve Closure Recuired Actuation Sional

d. (1) Above Delay - 30 seconds Reactor Scram Reovired Actuation Sional A. (1) or (2) Above Delay

.9 second instrument delay; 3.0 second insertion tirne.

't 9

Oi O OE

~ _. - _

MNPS-2 FSAR TABLE 14.1.5-4

?

ACTUATION SIGNALS AND DELAYS iSTEAM LINE BREAKI Charoin Pumo Actuation Reauirbd Actuation Slanal

gf9, A. (2 bove DS.htY - 40 se nds l

1

1. *,')

l

~'

-I l

J

\\

]

1481s-4.Me2 2 of 2 Octnhar 19A41 /

O t

MNPS-2 FSAR t

l TABLE 14.1.5-5 ADVANCED NUCLEAR FUELS-RELAP NUCLEAR INPUT AND ASSUMPTION (STEAM LINE BREAK) l Point Kinetic innut Value l

Effective Dela Neutron Fraction

.0049 Effective Neutron 'fetime (micro sec) 22.0 Minimum Shutdown activity Requirement 3.6% delta rho I

Stuck Rod Location S7jo Within half core section coole by the affected loop Fission Product and Actinide Deca Constants l

Default values in ANF-RELAP utilized i

)

I l

t I

l l

j f

2 14 sis.s.w2 1 of 1 n< enhar 1904 f

l

MNPS-2 FSAR

~

l TABLE 14.1.5-6 STEAM LINE BREAK ANALYSIS

SUMMARY

Maximum irntial Offsite Post Scram Maximum Power Power Retum to LHGR Level vailable Power (MWt)

MDNBR _

(kW/ft) h HZP Yes 686 2.40

< 21.0 HZP 294 1.18 16.5 HFP Yes 394 3.00 17.1 HFP No 147 4.60 5.7 4

usew2 1 Of 1 October 1994 /

MNPS 2 FSAR

)

TABLE 14.1.5-7 TEAM LINE BREAK SEQUENCE OF EVENTS-HOTZERO POWER-POWER T_imt*

Event O.

Reactor at hot zero power.

0. +

ouble-ended guillotine break located between affected steam generator and flow restrictors.

t 3.6 Main team isolation valve closure signal generated by low steam generator press e.

10.5 Main ste line isolation valves stop blowdown from intact steam generator 6.9 secon fter low steam generator pressure sign'al.

15.2 Safety injecti signal generated by low primary coolant pressure.

kg 32.

Reactor becomes ritical.

45.2 HPS! and charging p mps actuated.

153.

Thermal power reaches aximum level at 25% of rated power.

153.

First boron has passed thr ugh core.

180.

Auxiliary feedwater initiated affected steam generator.

600.

Auxiliary feedwater isolated ma ally.

600.+

Primary system temperature increa due to steam generator dryout and additional boron injection will termin e power excursion.

l i

'Eme after break, seconds 14515-7.MP2 1 O{ 1 Octnhor 1QQd \\

-. ~ -

MNPS 2 FSAR l

i 3

TABLE 14.1.5-8

.l

\\

STEAM LINE BREAK SEQUENCE OF EVENTS-HOT ZERO POWER-WITHOUT OFFSITE POWER Emg*

Event

{

0.

actor at hot zero power.

O. +

Dou le-ended guillotine break located between affected steam generator and the fl w restrictor.

3.6 Main ste m isolation signal generated by low steam generator pressure.

10.5 Main steam 'ne isolation valves stop blowdown from intact steam generator 6.9 seconds er low steam generator pressure signa' l.

18.7 Safety injection 'gnal generated by low primary coolant pressure.

48.7 HPSI and charging ps actuated.

50.

Reactor becomes criti 1.

5/to 152.

First boron has passed th ugh core.

169.

Thermal power reaches ma um level at 11 % of rated power. -

180.

Auxiliary feedwater initiated to ffected steam generator.

600.

Auxiliary feedwater isolated manu ly.

600.+

Primary system temperature increase ue to steam generator dryout and additional boron injection will terminat ower excursion.

i l

iI.

i i

• Time after break, seconds

1 MNPS-2 FSAR

-s J

TABLE 14.1.5-9 STEAM LINE BREAK SEQUENCE OF EVENTS-HOT FULL POWER-POWER AVAILABLF, TJ.ma Event O.

Reactor at hot full power.

0. +

ouble-ended guillotine break located between affected steam generator and t

flow restrictor.

3.5 Reac trip and main steam isolation valve closure signal generated by low

• steam enerator pressure.

10.4 Main stea line isolation valves stop blowdown from intact steam generator 6.9 seconds fter low steam generator pressure sigrial.

13.8 Safety injection ' nal generated by low primary coolant pressure.

43.8 HPSI and charging ps actuated.

174.

Reactor becomes critica 180.

Auxiliary feedwater initiate o affected steam generator.

204.

Thermal power reaches maximu level at 15% of rated power.,

E/90 )

204.

First boron has passed through core 600.

Auxiliary feedwater isolated manually.

600.+

Primary system temperature i;ncrease due t steam generator dryout and additional boron injection will terminate pow excursion.

k

)

' Time after break, seconds 34 sis-o.un 1 of 1 October 1994 1 /

MNPS-2 FSAR TABLE 14.1.510 STEAM LINE BREAK EVENT SEQUENCE - HOT FULL POWER -

WITHOUT OFFSITE POWER T1!ng*

Event i

O.

eactor at hot full power.

O. +

Do'uble-ended guillotine break located between affected steam generator and the how restrictor.

3.6 Reactor rip and main steam isolation signal generated by low steam generator ressure.

10.5 Main steam lerte isolation valves stop blowdown from' intact steam generator g/9o 6.9 seconds after low steam generator pressure signal.

16.0 Safety injection si nal generated by low primary coolant pressure.

HPSI and charging p\\

46.0 u ps actuated.

180.

Auxiliary feedwater initia d to affected steam generator.

l 224.

First boron has passed throu core.

235.

Reactor becomes critical.

250.

Thermal power reaches maximum i el at 5.4% of rated power.

600.

Auxiliary feedwater isolated manually.

600.+

Primary system temperature increase due steam generator dryout and j

additional boron injection will terminate po r excursion.

' Time after break, seconds 54 sis.io. w 2 1 of 1 nr.enhor 1 o04 1.

I MNPS-2 FSAR i

Table 14.1.5.1-3 ANF-RELAP Thermal-Hydraulic input (Pre-Scram Steam Line Break) i Initial Condition Thermal-Hydraulic Inout HE Reactor Power (MW) 2754 l

Pressurizer Pressure (psia) 2250 Pressurizer Level (%)

65 l

Cold Leg Coolant Temperature ( F) 549 Total Primary Flow Rate (Ibm /sec) 37,640 Secondary Pressure (psia) 881 Core Bypass Flow Rate (Ibm /sec) per Loop 753 Main Feedwater Temperature ('F) 432 Steam Generator Mass inventory (Ibm) 167,237 i

l l

l l

l l

[

MNPS-2 FSAR l

Table 14.1.5.1-4 l

Actuation Signals and Delays (Pre-Scram Steam Line Break)

Non-Harsh Harsh Containment Containment Condition Condition Reactor Trio SetDoint Setooint Delav Variable Overpower (ceiling) 111.6% of rated Not credited 0.9s Low Reactor Coolant Flow Credited 85% flow 0.65 s High Containment Pressure Not appficable 5.83 psig 0.9 s Low Steam Generator Pressure 658 550 0.9 s 1728 psia 1700 psia 0.9 s 7)

TM/LP(function)

Evaluated from function Not credited 0.9 s given in Technical Specification

l MNPS-2 FSAR l

l-Table 14.1.5.1-5 ANF-RELAP Neutronics input and Assumptions (Pre-Scram Steam Line Break) l l

l' Point Kinetics Inout Value l

l Effective Delayed Neutron Fraction 0.0054 i.

l Moderator Temperature Coefficient (pcm/'F)

Offsite Power Available (Technical Specification most negative 28 l

limit)

Loss of Offsite Power (Technical Specification most positive limit

+4 above 70 % RTP)

HFP Scram Worth (pcm) 6628 Shutdown Margin Requirement (pcm) 3600

. Doppler Coeflicient Offsite Power Available 1.20 x most-negative value at EOC Loss of O!Tsite Power 0.80 x least-negative value at BOC Fission Product and Actinide Decay Constants Default values in ANF-RELAP utilized

MNPS-2 FSAR Table 14.1.5.1-6 MDNBR and Peak Reactor Power Level Summary (Pre-Scram Steam Line Break)

Location of Type of Size of Peak Reactor Power MDNBR Break Cooldown Break

(% of rated)

Outside containment, downstream of check 2

Symmetric 3.00 ft 1.310 130.01 %

valves 3.' 0 ft 2

5 1.298 130.91 % '

2 l.20 ft 1.254 124.42 %

Outside containment, Asymmetric i

upstream of check 2

1.6011 1.302 124.87 %

valve 2

1.80 fi 1.334 124.92 %

2 0.40 ft 1.299 117.85 %

Inside containment.

Asymmetric upstream of check 2

0.80 ft 1.262 122.26 %

valve 2

1.80 ft 1.318 125.51 %

Inside containment, 2

2 Asymmetric 3.51 ft 0;o8 106.86 %

upstream of check valve with loss of offsite power The peak LHRs for all pre-scram breaks are bounded by the peak LHR for the 3.50 ft break outside containment and downstream of a check valve.

2 The MDNBRs for all pre-scram breaks are bounded by the MDNBR for the 3.51 ft break inside containment and upstream of a check valve with the loss of offsite power.

I MNPS-2 FSAR Table 14.1.5.1-7 LHGR-Limiting Pre-Scram Steam Line Break Sequence if Events:

HFP 3.50 ft* Symmetric Break Outside Containment with Offsite Power Available Time (se_c)

Event 0

Break downstream of main steam line check valves opens 0

Turbine controlvalves open fully 7

Low steam generator pressure trip setpoint reached 8

Tmbine trips on reactor scram signal 9

Scram CEA insertion begins 9

Reactor power reaches maximum value 10 MDNBR occurs l

4 1

l l

1

MNPS-2 FSAR Table 14.1.5.1-8 l

l MDNBR-Limiting Pre-Scram Steam Line Break Sequence of Events: HFP 3.51 ft* Asymmetric Break inside Containment with Loss of Offsite Power Time (sec)

,Eleni 0

Break occurs O

RCPs trip 0.

Peak LHGR(kW/ft) 2 Scram signal on low flow trip 3

Scram CEA Insertion begins 3

Max Power (Fraction of RTP) 4 MDNBR l

1 j

l l

l~

l i

MNPS-2 FSAR l

tv. i. s. a - i

)

TABLE 14.1.5-;

l AVAILABLE REACTOR PROTECTION FOR STEAM SYSTEM PIPING FAILURES INSIDE AND OUTSIDE OF CONTAINMENT EVENT -

0057-Sc eam 446 L V.szs Reactor Operatina Conditions Reactor Protection 1

Low Steam Generator Pressure Trip Low Steam Generator Water Level Trip Variable Overpower Trip Local Power Density Trip Therma 1 Margin / Low Pressure Trip High Containment Pressure Trip 5l90 Safety injection Actuation Signal 2

Low Steam Generator Pressure Trip Low Steam Generator Water Level Trip Variable Overpower Trip High Containment Pressure Trip Safety injection Actuation Signal 3-6 Technical Specification Requirements on Shutdown Margin, inherent Negative Doppler Feedback 14S161.MP2 1 of 1 October 1994 l

MNPS-2 FSAR N /.6,2-2 i

TABLE 14.1.5-2 DISPOSITION OF EVENTS FOR STEAM SYSTEM PIPING FAILURES INSIDE AND OUTSIDE OF CONTAINMENT EVENT

[Q 5 7 - scpi,tm Mc V5xS Reactor Oneratino Conditions Disposition 1

Analyze 2

Analyze 5l10 3-6 Bounded by the above i

)

14S16 2.MP2' 1 Of 1 October 1994]

MNPS-2 FSAR Table 14.1.5.2-3 ANF-RELAP Thermal-Hydraulic Input (Post-Scram Steam Line Break)

Initial Condition Thermal-Hydraulic Inguj HFP HZP Core Power (MW) 2700 lE-6 Primary Pressure (psia) 2250-2250 Pressurizer Level (%)

65 40 Cold Leg Temperature ( F) 549 532 Primary Flow Rate per Loop (Ibm /sec) 18,820 19,241 Secondary Pressure (psia) 880 892 Steam Generator Mass Irwentory (Ibm) 167,237 253,989 Total Steam Flow (Ibm /sec) per Steam Generator 1634 4

Total MFW Flow (Ibm /sec) per Steam Generator 1634 4

MFW Temperature ( F) 432' 432 Total AFFlow(Ibm /sec) 184 184 RWST Boron Concentration (ppm) 1720 1720 AF Temperature (*F) 32' 32 Break Characteristics Minimum Flow Area 2

Affected Steam Generator (ft )

3.51 2

Unaffected Steam Generator (f1 )

3.51 Location of Pine Break Downstream of steam generator integral flow restrictor and upstream of MSIV 1 of 2 l

f f

[-

MNPS-2 FSAR r

Table 14.1.5.2-3 l

t j

ANF-RELAP Thermal-Hydraulic input (Post-Scram Steam Line l

Break) l l

l i

Iniection Systems HFP HZP Total HPSI Pumps -

3 3

Active HPSI Pumps 2

2 Single Failure (No credit for mounted spare) 1HPSIpump 1 HPSI pump Active Charging Pumps 0

0 Refueling Water Storage Tank Boron Concentration (ppm) 1720 1720 HPSI Delivery Curve Fig.14.1.5.2-1 Fig.14.1.5.2-1 Feedwater Auxiliary Flow, maximum (Ibm /sec) 183.6 IF7.6 Temperature (*F) 32.1 32.1 Mai.11 Initial Flow per Steam Generator (Ibm /sec) 1634.1 0.0 Initial Temperature (*F) 432.4 N/A 2 0f 2 i

I I

r l

l i

l l

l1 MNPS-2 FSAR I

Table 14.1.5.2-4

- Actuation Signals and Delays (Post-Scram Steam Line Break)

Inside Outside Parameter Setooints Containment Containment

1. Low Steam Generator Pressure Trip 550 psia 658 psia

2. Low Pressurizer Pressure SIAS 1500 psia 1578 psia i -

3. Low Steam Generator Pressure MSI 370 psia 478 psia l

MSIV Closure Reauired Actuation Sianal (3) Above t

Delay - 6.9 seconds HPS' Actuation 1

l Reauired Actuation Signal l

l (2) Above~

Delay - 25.0 seconds Main Feedwater Valve Closure B_eguired Actuation Sinnat (3) Aixwe Delay - 14.0 seconds i

Reactor Scgm Reauired Actuation Sinnal (1) Above Delay - 0.9 second instrument delay l

3.0 second insertion time l

i i

l

.~_.

MNPS-2 FSAR Table 14.1.5.2-5 ANF-RELAP Neutronics input and Assumptions (Post-Scram Steam Line Break) l P_qint Kinetics Input Value i

Effective Delayed Neutron Fraction 0.0054 1

Moderator Temperature Coefficient (pcm/ F)

-28.0 i

HFP Scram Worth (pcm) 6438.0 Shutdown Margin Requirement (pcm) 3600.0 l

Stuck Rod Location l

Within half-core section cooled by affected loop Fission Product and Actinide Decay Constants Default values in ANF-RELAP utilized i

l l

~ _. _ _. _

.. -.. _ _ _.. _ _ -. _... _ _ _ _.. ~ ~. _ _ _.

l^

MNPS-2 FSAR Table 14.1.5.2-6 Post-Scram Steam Line Break Analysis Summary Initial Power Offsite Power Break Maximum MDNBR Maximum Fuel Failure Level Available Location Post Scram LHGR

(% of Core) l Return to Power (kW/ft)

(MW)

HFP No outside 207.5 1.71 17.96 0.0 containment HFP Yes outside 378.0 2.28 24.27 0.5 containment l

HZP No outside 182.9 1.89 15.76 0.0 containment HZP Yes outside 343.5 2.37 23.47 0.3 l

containment I

Initial Offsite Break ANF-RELAP XTGPWR Conservatisms in Net i

Power Power Location Reactivity Reactivity input Parameters Conservatism in l

Level Availability Change Change (MTC, Doppler, and ANF-RELAP j

[

($)

($)

Scram Worth Bias)

. model l

($)

($)

l HFP No outside

+0.00

-6.30

+5.30

+1.00 containment HFP Yes outside

+0.00

-5.87

+4.86

+1.01 containment HZP No outside

+6.69

+3.00

+2.72

+0.97 cont,ainment HZP Yes outside

+6.68

+3.43

+2.34

+0.91 containment l

l l

l I

l l

l i

I

MNPS 2 FSAR Table 14.1.5.2-7 LHGR-Limiting Post-Scram Steam Line Break Sequence of Events: HFP Outside Containment Break with Offsite Power Available Time (sec)

Event O.

Reactor at HFP 0.+

Double ended guillotine break.

4 Low steam generator pressure trip, Reactor trip 11-MSIV and MFW valves closure trip signal 16 SI signal

~17

- MSIVs closed 25 MFW valves closed 41 Si pumps at reted speed (25 s delay) 180 AFW starts 462 Peak post-scram power reached (378.03 MW)

N/A Si lines cleared. Boron begins to enter primary system 490 Steam generator dry out 600 Calculation te.Tninated. Power decreasing.

r i

i t

l l

l l

l

)

i MNPS-2 FSAR i

Table 14.1.5.2-8 MDNBR-Limiting Post-Scram Steam Line Break Sequence of Events: HFP Outside Containment Break with Loss of Offsite Power Time (sec)

Event O.

Reador at HFP 0.+

Double ended guillotine break. Loss of offsite power.

4 Low steam generator pressure trip, Reactor trip 9

MSIV and MFW valves closure trip signal 16 MSIVs closed '

i 18 Si signal 23 MFW valves closed 43 Si pumps at rated speed (25 s delay) 180 AF# starts 488 Peak post-scram power reached (207.47 MW)

N/A Si lines cleared. Boron begins to enter primary system 600 Calculation terminated. Powerdecreasing.

I I

l i

MNPS-2 FSAR TABLE 14.1.5.3-1 l

l ASSUMPTIONS USED IN MAIN STEAM LINE BREAK ANALYSIS Core Power Level (MWt) 2754 Primary to Secondary Leak Rate per Steam Generator 0.035 gpm Primary Coolant Iodine Concentration 1 uCi/gm DE I-131 Secondary Coolant Iodine Concentration 0.1 uCi/gm DE I-131 Primary Coolant Noble Gas Concentration 100/Ew i

Pre-accident Spike lodine Concentration 60 uCi/gm DE I-131 Melted Fuel Percentage 0.46%

Peaking Factor 1.45 Reactor Coolant Mass 430,000 lbs Intact Steam Generator Minimum Mass 100,000 lbs Safety Injection Signal Response 85 seconds 3

Site Boundary Breathing Rate (m /sec) 0 - 8 hr 3.47E-04 8 - 24 hr 1.75E-04 24 - 720 hr 2.32E-04 Site Boundary Dispersion Factors (sec/m3)

EAB: 0 - 2 hr 3.66E-04 LPZ: 0 - 4 hr 4.80E-05 4 - 8 hr 2.31E-05 8 - 24 hr 1.60E-05 24 - 96 hr 7.25E-06 96 - 720 hr 2.32E-06 Control Room Breathing Rate 3.47E-04 m /sec 3

Control Room Damper Closure Time 5 seconds l

Control Room Intake Prior To Isolation 800 cfm Control Room Inleakage During Isolation 130 cfm Control Room Emergency Filtered Recirculation Rate (t=10 mia) 2,250 cfm Control Room Intake Dispersion Factors (sec/m3)

PORVs/ADVs: 0 - 8 hr 3.19E-03 8 - 24 hr 2.05E-03 24 - 96 hr 7.61 E-04 96 - 720 hr 2.13E-04 Turbine Building Blowout Panels: 0 - 8 hr 4.23E-03 8 - 24 hr 2.85E-03 24 - 96 hr 1.12E-03 96 - 720 hr 3.63E-04 3

Control Room Free Volume 35,65011 Control Room Filter Efficiency (all iodines) 90 %

Thyroid Dose Conversion Factors ICRP 30 I

I i

_ ~. _ _..

i I

i MNPS-2 FSAR TABLE 14.1.5.3-2

SUMMARY

OF MILLSTONE 2 MSLB ACCIDENT DOSES (0.46% Melted Fuel) i Location Thyroid (rem)

Whole Body (rem)

Beta (rem)

EAB 4.8 0.06 N/A LPZ 2.3 0.02 N/A Control Room 29 0.03 0.5 l

l l

I I

r i

l l

o

(

i i

i i

MNPS-2 FSAR TABLE 14.1.5.3-3

SUMMARY

OF MILLSTONE 2 MSLB ACCIDENT DOSES (Pre-accident Iodine Spike)

Location Thyroid (rem)

Whole Body (rem)

Beta (rem)

EAB 0.935 0.010 N/A LPZ 0.176 0.002 N/A Control Room 5.314 0.003 0.039 i

I I

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d

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V MNPS-2 FSAR-MAY,1990 d

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0.

20.

40.

60.

80.

300.

FLOW TO PRIMARY SYSTEM (LBM/SEC)

\\s L.

FIGURE 14.1.5-1 ONE PUMP HIGH PRESSURE SAFETY INJECTION SYSTEM DELNERY VS. PRIMARY PRESSURE l

L_ :

0

_)

i MAY, 1990 lePS-2 FSAR

/

12000 i

10000 t

8000 T

t ao 6000 o

e f

n0 2

4000 -

t 20 l

I t

0 400 600

~

0 200 Time (s) l FIGURE 14.1.5-2 AFFECTED STEAM GENERATOR BREAK FLOW VS.

[

TIME - HOT ZERO POWER WITH OFFSITE POWER

~

i

L

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V MAY, 1990 MNPS-2 FSAR' MO i

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FIGURE 14.1.5-3 AFFECTED CORE SE., OR N.ET TEMPERATURE VS.

f TIME - HOT ZERO POWER WITH OFFSITE POWER i

I'

,.;1 t

t;!t i!

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MAY,,1990 l

/

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0190090000P l

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r 2500 l

+

l q

s 2000 6

t 1500 2

e a

9*

E j

1000 e

r

/

I O

O 200 400 600 Time (s)

(

FIGURE 14.t5-5 PRESSURIZER PRESSURE VS. TIME - HOT ZERO POWER WITH OFFSITE POWER i

\\.~~

V

_j MAY, 1990 m _.2 FMR 2

D O

m r

-2 nS

-4 a

e sr c

-6 I

t'

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200 400 600

~

Time (s)

FIGURE 14.1.5-6 ADVANCED NUCLEAR FUELS - RELAP CALCULATED

}

CORE REACTUITY VS. TIME - HOT ZERO POWER WITH OFFSITE POWER i

l

i

'w

\\d

..,/

MNPS-2 FSAR MAY, 1990 l

5 i

i T

H I

l iL.

i 4-ac b

h a.

e&~

t

/

i M.

400.0 000.9 0.0 200.0 MW t

FIGURE 14.1.5-7 CORE POWER VS. TIME - HOT ZERO POWER WITH OFFSITE POWER f

.w.

MPS-2 FSAR

. r M

i i

10000 8000 ao MM os:

a m

l O

3 4000

[

4 i

O I

0 200 400 600 Time (s)

[

FIGURE 14.1.5-8 AFFECTED STEAM GENERATOR BREAK FLOW VS.

TIME - HOT ZERO POWER WITHOUT OFFSITE POWER

[

i r._.

'w l#PS-2 FSAR MAY, 1990 700 f

/

ostoomocawVI 800 e

i G

i 500

~

b i

.T g@o l

E i

h 30 i

i t

yl l

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200 400 600 Time (s) i i

FIGURE 14.1.5-9 AFFECTED CORE SECTOR IftET TEMPERATlRE VS.

TIME - HOT ZERO POWER WITHOUT OFFSITE PnWER i

i i

G' 1

MNPS-2 FSAR gy, 39g M

i i

l 082001000,0HIMPF r

600 i

E L

300

'1 E

/

[

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l

=S -@0 o

30 [

i l

l 200 "

0 M0

@0 MO Time (s) t i

L FIGURE 14.1.5 - 10 INTACT CORE SECTOR litET TEMPERATURE VS.

TIME - HOT ZERO POWER WITHOUT OFFSITE POWER e

-.e i)

T3

)

p i

i 8e 8

N i

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d

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3 MNPS-2 FSAR MAY, 1990 0

2 G

V

~

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Time (s)

FIGURE 14.1.5 - 12 ADVANCED NUCLEAR FUELS - RELAP CALCULATED CORE REACTNITY VS. TIME - HOT ZERO POWER WITHOUT OFFSITE POWER

-rrem+.m

[$>(( f). 't

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i a

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t FIGURE 14.1.5 - 13 CORE POWER VS. TIME - HOT ZERO POWER -

WITHOUT OFFSITE POWER l

[

s r

MNPS-2 FSAR 1

f insert 13 crossed out figures: 14151.. - -- 13 1

i l

l -

l I

... -. ~...

--.. ~ ~ _ ~

i' MNPS-2 FSAR l

l 1.4 i

• 1.3 2

1.2 1.1

,,,,,.g.M

(

5 5 1.0

~~ ~~E ~l~.3 #

\\

2 O

.9 k.

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j

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l g

,7 Indicated thermal power Indicated nuclear power k

g

.6 Core-overage heat flux

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6 7

8 9

10 11 12 Time (s) 2 Figure 14.1.5.1-1 Normalized Core Power (Symmetric 3.50 ft Break Outside Containment with Offsite Power Available) l t

I i

l i

l r

MNPS-2 FSAR 610 i

i 600 7""-~~-

~

2.'.7C,y,- -..,

~

590 N,' N

's g

7580 5

5 570

- - - - Loop 1 hot leg E

Loop 2 hot leg 5s 560 Loop 1 cold legs

- -- Loop 2 cold legs 8

550 o

8

'm 540

'N 530

% ~~' % :.: m-520 t

O 1

2 3

4 5

6 7

8 9

10 11 12 Time (s)

Figure 14.1.5.1-2 Core inlet Temperatures (Symmetric 3.50 ft*

Break Outside Containment with Offsite Power Available)

MNPS-2 FSAR I

1.0

.0

._.. _.,_... / *f,

)

5

=~=:::. : ::--~...., r.,,

b

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v v

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-4.0

- 5.0

-6.0 I

I I

I I

I i

i i

i 1

0 1

2 3

4 5

6 7

8 9

10 11 12 Time (s)

Figure 14.1.5.1-3 Reactivity Feedback (Symmetric 3.50 ft' Break Outside Containment with Offsite Power Available) 4 T

.... _ ~ -

L i

i l

MNPS-2 FSAR f

r i

t l

l t-I 2260 2240 l

l 2220 7

g 2200 v

_l2180

[ 2160 Eq 2140 ll E.2120 n.

2100 2080 2060 I

I i

e i

i i

i e

i i

0 1

2 3

4 5

6 7

8 9

10 11 12 Time (s) 1

)

Figure 14.1.5.1-4 Pressurizer Pressure (Symmetric 3,50 ft" Break Outside Containment with Offsite Power Available) i i

.m.

MNPS-2 FSAR i

880 i

i i

i i

i i

i i

i g

--- SG 1

-\\'\\

~

860 SG 2 840 7

l 820

'i',s g

v

$g\\.

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2 800 5j780

\\'\\

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0 720

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I

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n n

a e

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1 2

3 4

5 6

7 8

9 10 11 12 Time (s)

Figure.14.1.5.1-5 Steam Generator Pressures (Symmetric 3.50 2

ft Break Outside Containment with Offsite Power Available) l i.

I i

m.. - -

MNPS-2 FSAR I

l i

l l

i 6000

--- Turbine l

Break i

5000 -

?

l N

En c, 4000 -

Oa

$ 3000

\\..

-.s,'

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i:

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2000 E

2

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\\.

\\

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i.

i O

1 2

3 4

5 6

7 8

9 to 11 12 Time (s) 2 Figure 14.1.5.1-6 Main Steam Line Flow (Symmetric 3,50 ft Break Outside Containment with Offsite Power Available) l l

l l

MNPS-2 FSAR 1.3 i_ i i

i i

i i

i i

i i

2*1.2 Reactor Power E

Core-overage heat flux 1.1 o

l-c 1,0 o

_'s s

y o.

.9 s

t V

s

'.8 s

x l

2

\\

.7 s

"Ei

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i 0 1 2

3 4

5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 Time (s)

Figure 14.1.5.1-7 Normalized Power and Heat Flux (Asymmetric 3.51 ft* Break inside Containment with Loss of Offsite Power) l.

l-l l

l 3.

l

. - ~.,

MNPS-2 FSAR l~

650 i

i i

i i

i i

i i

i i

i i

i i

i i

i i

i 640 630 620 610

,.s w

600 s

2

's 2 590

's s

2

's

• 580

's c.

s b570

' 's

~

s 7 560 s

O

's j 550-

%,_._._,_,~ -.~._.._._....

,]

u 540

' N,,, N 530 Loop 1 hot leg

..N Loop 2 hot leg s.N 520 Loop 1 cold legs 510.

-- Loop 2 cold legs s,,' N. s D

500 i

i i

i i

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i i

i i

i i

i i

i

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2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 l

Time (s) l l-Figure 14.1.5.1-8 Reactor Coolant Temperatures (Asymmetric 3.51 ft Break inside Containment with Loss of Offsite Power) 2 l

l l

L t

i l.

i

=.

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MNPS 2 FSAR 100 i

i i

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20 to 0

i i

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i 0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 Time (s)

Figure 14.1.5.1-9 Normalized RCS Flow Rate (Asymmetric 2

3.51 ft Break inside Containment with Loss of Offsite Power) e 4

1

-.w r 1---

w

' ' ' ' ' -* - =

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i i

4 MNPS-2 FSAR l

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2250 2200 o

l

'E 2150 3

$2100

. [ 2050 5

.t! 2000 5

m E 1950 o.

1900 1850 1800 t

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i i

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0 1 2

3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 lime (s)

Figure 14.1.5.1-10 Pressurizer Pressure (Asymmetric 3.51 ft*

Break inside Containment with Loss of Offsite Power) 1 I

t-I 1

i

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. -. -. ~

l l

MNPS-2 FSAR f:

12m 1100 1000-900 a

W E 800

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i 0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 Time (s)

Figure 14.1.5.1-11 Steam Generator Pressures (Asymmetric 3.51 ft* Break ins,ide Containment with Loss of Offsite Power)

.... -. ~.

MNPS-2 FSAR l

1200 A

\\

s

. 3,.1000 -

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ta> 400 -

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0 l

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10 20 30 40 50 60 70 80 Flow to Primary System (Ibm /sec)

Figure 14.1.5.2-1 One Pump High Pressure Safety injection System Delivery vs. Primary Pressure (Post-Scram Steam Line Break) 4 4

)

i l

lc I

l' t

MNPS-2 FSAR Millstone Cycle ~ 13 ' MSLB HFP Pumps On 3000.0

,i Affected SG

---

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~

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E

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4

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.0 100.0 200.0 300.0 400.0 500.0 600.0 Time (sec)

Figure 14.1.5.2-2 Steam Generator Break Flow (HFP Post-Scram Steam Line Outside Containment Break with Offsite Power Available)

MNPS-2 FSAR t

Millstone Cycle 13 MSLB HFP Pumps On l

1000.0 l

Unaffected SG Affected SG l

800.0 l

a l 600.0 v

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8 n

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4 200.0

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.0 100.0 200.0 300.0 400.0 500.0 600.0 Time (sec)

Figure 14.1.5.2-3 Steam Generators' Secondary Pressures (HFP Post-Scram Steam Line Outside Containment Break with Offsite Power Available) l l

l

MNPS-2 FSAR i

Millstone Cycle 13 MSLB HFP Pumps On 600.0 i ' ' ' = i - -

i.

Affected Region

~

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---

Unaffected Region r

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8 3

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ba

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E 350.0 v

H 300.0

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~

250.0 200.0

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.0 100.0.

200.0 300.0 400.0 500.0 600.0 Time (sec)

Figure 14.1.5.2-4 Core inlet Temperatures (HFP Post-Scram Steam 1.ine Outside Containment Break with Offsite Power Available) l i

i l

MNPS-2 FSAR Millstone Cycle 13 MSLB HFP Pumps On 2500.0 2000.0 o

m O

8 1500.0 a

g 3-f 1000.0 500.0

..i*

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.... i

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.0 100.0 200.0 300.0 400.0 500.0 600.0 Time (sec) i Figure 14.1.5.2-5 Pressurizer Pressure (HFP Post-Scram Steam Line Outside Containment Break with Offsite Power Available) 1 i

(

l

MNPS-2 FSAR Millstone Cycle 13 MSLB HFP Pumps On 70.0 60.0 2

1 q 50.0 o

O.

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,o tt v

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> 20.0

=-

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i L

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e

.0 100.0 200.0 300.0 400.0 500.0 600.0 Time (sec) 1 Figure 14.1.5.2-6 Pressurizer Level (HFP Post-Scram Steam Line Outside Containment Break with Offsite Power Available) w

__..-..__....-=__

_m MNPS-2 FSAR I

1 g

Millstone Cycle 13 MSLB HFP Pumps On 25.0 7

22.5

~

Unoffected SG i

Affected SG l

20.0 7

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~

2 l

T 15.0 f., ^g 2

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1

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.0 100.0 200.0 300.0 400.0 500.0 600.0 Time (sec)

Figure 14.1.5.2-7 Steam Generators' Secondary Mass (HFP Post-Scram Steam Line Outside Containment Break with Offsite Power Available) l i

c l-4

.I 9'

i MNPS-2 FSAR'-

t 1

Millstone Cycle 13 MSLB HFP Pumps On 14.0 12.0

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~ 10.0 Moderator j

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.t

... i

...i....i.

.. i....-

.0 100.0 200.0 300.0 400.0 500.0 600.0 Time (sec) i Figure 14.1.5.2-8 Reactivity Components (HFP Post Scram Steam Line Outside Containment Break with Offsite Power Available) l i

1

]

i e

I

l MNPS-2 FSAR l

Millstone ' Cycle 13 MSLB HFP Pumps On 500.0 450.0 400.0

{

^

3 350.0 2

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100.0 200.0 300.0 400.0 500.0 -

600.0 Time (sec)

Figure 14.1.5.2-9 Reactor Power (HFP Post-Scram Steam Line Outside Containment Break with Offsite Power Available) l-l l

t

MNPS-2 FSAR

-Millstone Cycle 13 MSLB HFP Pumps Off 3000.0

[1 Affected SG Unoffected SG 2500.0 ti

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. s 0 2000.0

's E

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a

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i.

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.0 100.0 200.0 300.0 400.0 500.0 600.0 Time (sec)

Figure 14.1.5.2-10 Steam Generator Break Flow (HFP Post-Scram Steam Line Outside Containment Break with Loss of Offsite Power) l

\\

i

a a

4 4-

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a a.sa a

.s.a s

=a

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Unaffected SG

---

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Figure 14.1.5.2-11 Steam Generators' Secondary Pressures (HFP Post-Scram Steam Line Outside Containment Break with Loss of Offsite Power) l e

l l

l l-

{

l

MNPS-2 FSAR

. Millstone Cycle 13 MSLB HFP Pumps Off 600.0 Affected Region 550.0

---

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{

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.0 100.0 200.0 300.0 400.0 500.0 600.0 Time (sec)

Figure 14.1.5.2-12 Core inlet Temperatures (HFP Post-Scram Steam Line Outside Containment Break with Loss of Offsite Power) i

(

l l

1 i

l I

MNPS-2 FSAR Millstone Cycle 13 MSLB HFP Pumps Of f 2500.0 2000.0 o

eg O

o 1500.0

~

CL 1000.0 500.0

.n

.0 100.0 200.0 300.0 400.0 500.0 600.0 Time (sec)

. Figure 14.1.5.2 13 Pressurizer Pressure (HFP Post-Scram Steam Line Outside Containment Break with Loss of Offsite Power)'

l 1

i

MNPS-2 FSAR t

Millstone Cycle 13 MSLB HFP Pumps Off 70.0 I

60.0

.p 50.0 Do.

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v 30.0 m

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> 20.0 10.0

.k

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.600.0 Time (sec)

Figure 14.1.5.2-14 Pressurizer Level (HFP Post-Scram Steam Line Outside Containment Break with Loss of Offsite Power)

l i

MNPS-2 FSAR Millstone Cycle 13 MSLB HFP Pumps Off 14.0 1

[

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i Figure 14.1.5.2-15 Reactivity Components (HFP Post Scram Steam Line Outside Containment Break with Loss of Offsite Power)

_~... - -. -.....- -

4 MNPS 2 FSAR Millstone Cycle 13 MSLB HFP Pumps Off

~ 500.0 450.0 2

400.0 2

m 3 350.0 2

2 v

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200.0 3

o e 150.0 fr 100.0 50.0 i

.0

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Figure 14.1.5.2-16 Reactor Power (HFP Post-Scram Steam Line Outside Containment Break with Loss of Offsite Power) l i

(

MNPS 2 FSAR Assumotion (1)

Activity in Containment available for release 100% Noble Gases 25% lodines (k-t }

(2)

Initial lodine Chemical Form:

91 % elemental 4% organic 5% particulate (3)

Purging Occurs 5 days after initiation of LOCA (4)

Breathing Rate = 2.32 x 10d m*/sec (5)

Power Leve! = 2700 MWt (6)

Dose Conversion Factors Reg. Guide 1.109 (7)

Purge Rate = 50 ft*/ min (8)

Containment Building Volume = 1.899 x 10' ft*

(9)

Release Point Unit 1 Stock (10) Filter Efficiencies:

90% elemental iodine 70% organic iodine 90% particulate iodine (11). Duration of Purge = 30 days (12) X/Q (sec/m*)

LPZ (0 30) days = 6.97 x 10

14.8.4 Radiott,,

Consequences of the Design Basis Acciden 1gggRT D

+

N 14.8.4.1 General The DBA involves a gross release of activity from the fuel to the containment building.

This section discusses the consequences of such a release.

14.8.4.2 Method of Analyses The radiological consequences of a Design Basis LOCA at Millstone 2 were analyzed for a low wind speed condition and a high wind speed condition. These are represented by cases A and B, respectively, unru.a.up:

14.8-11 June 1996l

l MNPS-2 FSAR se A - Low Wind Speed Condition This se assumes meteorological conditions exist which will give 95 percent highest X/O values (e.g., low wind speeds). For this scenario the activity which leaks from the containm) building enters the enclosure building where it is treated by the enclosure building filt

' n system (EBFS) before being released through the Unit 1 stack. A small percenta e (1.69 percent) of the containment leakage bypasses the EBFS and is released at groun\\ level for the entire accident (30 days). All containment leakage for the first 110 seconb(is assumed to be a ground level release. This is due to the fact that it takes 110 seconds for the enclosure building to achieve negative pressure and thus assure that leakaghwill be into the enclosure building rather than out. All assumptions used in this h alysis are given in Table 14.8.4-1.

The radiological taaluation us d thyroid dose conversion factors consistent with those stated in Regulatory Guide 1.1b8 Rev.1. In addition to the Staff's acceptance of these dose conversion factors in the puMished Safety Evaluation Report for Erie Nuclear Power Plant (NUREG-0423), NNEC offers the following justification (1)

Dose Conversion Factors The NRC has published a revised version of Regulatory Guide 1.109 (Octo-ber,1977) for use in Appendix I keulations. The inhalation dose conver-sion factors (DrFs) contained in this guide are lower than those previously used in radiological evaluations to me reactor siting criteria of 10CFR100.

The source of the iodine DCF previously\\ sed in radiological off-site dose calculations is TID-14844(March,1962).}f mfinity and input paramete hey were derived using the acute intake model, a dose commitment o from ICRP-ll. Parameters from ICRP-il such as\\ effective half life, fraction of the isotope reaching the organ, and the effective energy released per disin-tegration were based on the best available dataht that time (1959). Due to a lack of information, various conservatisms were\\ employed in determining these parameters.

The DCF's in Regulatory Guide 1.109 were based on i QP VI and X. They use a chronic intake model with a dose' commitment extbnding for 50 years after intake. Credit is given for hold up of the nuclide in the lung before it reaches the thyroid. This has the effect of reducing the frakion of the nuclide reaching the organ of interest. The fraction was calco(ated based on information in ICRP X. The source of biological half lives wa's also based on ICRP X.

There is an appreciable difference between several factors upon whi DCF's from TID-14844 and Regulatory Guide 1.109 were based. The 'rst l

point of difference is the fraction (fa) of the nuclide which is deposited i the organ. As stated above, Reg. Guide 1.109 (Rev.1) takes credit for retention of the iodine isotopes in the lung. This is a more realistic ap-proach since it is expected that fewer of the short lived isotopes would be able to reach the thyroid than those that are long lived. Another factor where there are differences is the effective energy deposited in the organ.

MP214-8.MP2 14.8-12 June 1996 l

MNPS-2 FSAR j

l 1

The differences are only slight and are the result of different decay schemes used to compute this parameter. The Reg. Guide values were based on Nmore recent data and hence are expected to be more accurate. It should

,lso be pointed out that the Reg. Guide values are slightly higher than lC P 11 values which tend to increase the DCF's. The third factor where dif(fetences occur is in the half lives. ICRP 11 reported a biological half g

138 day while Reg. Guide 1.109 used a value of 100 days. Once again the Reg.

ide value was based on ICRP X. It is not clear what the reason is for the di rence in this factor. The differences in biological half lives is not critical sin e the effective half life is directly proportional to the DCF's.

The effective ha (lives do not significantly change. The change in biologi-

]

cal half lives, there ore, has little effect on DCF's.

In sumruary, the majo ifference between the DCF's presented in TID-14844 and Reg. Guide 109 is the credit taken by the Reg. Guide for hold up of iodine in the lung.

-14844 based its DCF's on the best informa-tion that was obtainable in 1 59 whereas those presented in the Reg.

Guide reilect the best informat n obtainable toclay. It is the conclusion of NNECO that the dose conversion actors in Reg. Guide 1.109 (Revision 1) are more applicable to offsite dose alculations and are therefore assumed in this dose analysis for Millstone Un No.2.

(2)

Calculational Methods in order to calculate offsite doses, the com ter code TACT 111 was em-ployed. This code evaluates the activities, an integrated doses at a site following the instantaneous or continuous relea e of halogens and noble gases from a control volume.

The input to'the program consists of the tirne-depe ent variables described below, the volume of the primary system, filter effici cies, etc.

Eighteen isotopes are included in the model, including K ptons, Xenons, and lodines. The isotope inventory may be input, or the ogram will calculate it based on TID source terms: decay is evaluated, s well as i

filtration. The primary containment leak rate, atmospheric d persion factors, and breathing rates may vary with time, at the optio of the user.

Site dose calculations use the semi-infinite cloud dose models uggested by Regulatory Guide 1.4.

Case B - High Wind Speed Conditions An analysis was performed to determine the effect on the enclosure building of 'gh wind speeds. The wind speed at which the enclosure building will begin to exfiltr te is one in which the corresponding wind velocity pressure is greater than the enclosur building negative pressure. The effect of wind on the enclosure building is discusse in Section 6.7.1.2.

The enclosure building filtration region (EBFR) design negative pressure is 0.25 in w.g.

The wind velocity corresponding to a velocity pressure equivalent to this EBFR design ura ws.ur2 14.8-13 June 1996 l

i MNPS 2 FSAR I

I i

essure is 25 mph. Above this velocity displacement of the enclosure building at. osphere with outside air would begin. However, for conservatism it is assumed that is displacement would begin with a 23 mph wind.

l t

The mo rmulated for site boundary and control room doses is as follows:

l (1) wind is from the (plant) North direction.

I (2)

The h h wind condition exists for the first 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> following the incident (5 perc t of thirty days).

(3)

Only those reas of the EBFR above grade are exposed to the wind effects.

Therefore, onigthe enclosure building structure is subject to air displace-ment due to wi effects.

(4)

The amount of pos ccident containment leakage is assumed as O.5 volume percent per day per chnical specification 3.6.1.

(5)

The amount of exfiltratiop based conservatively on a 30 mph wind is less than 10 percent of the ELKS exhaust capabil ties assuming only one fan operating. Therefore,10 pe cent of the EBFR atmosphere is conservatively assumed to be an unfiltered ound release.

All other assumptions and methodologies ar the same for Case A.

14.8.4.3 Dose Calculations 14.8.4.3.1 Thyroid Doses and Whole Body Exposure

~

'The results of the calculated doses for Cases A and B e shown in Table 14.8.4-2 and are within the limits of 10CFR100.

14.8.4.3.2 Control Room Habitability As a result of Three Mile Island (TMI) Action Plan item til.D.3.4 the potential radiologi-cal doses to the MP-2 control room operators have been reevalu ed. The analysis is based on the control room assumptions and meteorological param ers given in Tables 14.8.4-3 and 14.8.4-4.

The control room is designed to be occupied for the duration of the ac ent (30 days).

Two (2) basic sources of radiation have been evaluated. They were: (1 direct dose from sources outside the control room, and (2) the dose received from air rne activity which enters the control room. The analyses ensure that the operators wil e ade-quately protected from all sources of radiation.

The radiation design objective for the control room walls is to limit the whole b y dose to personnelinside the control room to less than 5 rem during any DBA. The ext rnal sources considered in the shielding evaluation are: (1) containment, (2) enclosure reactor building, (3) filtration systems, and (4) piping sources. The affect of extern sources from DBA's at Millstone Units 1 and 2 were evaluated in the shielding analy 's.

MP2M 8.MP2 14.8-14 June 1996 \\

MNPS-2 FSAR cause the containment as well as other sources at Millstone Unit No. 3 are separated fro the Unit 2 control room by a relatively large distance as well as other structures, a shiel g analysis from a Unit 3 accident was determined to be unnecessary.

The assu tions used in the shielding evaluation are listed in Table 14.8.4 5. The resulting do s from the shielding analysis are given in Table 14.8.4-6.

An EBFS signal om Millstone 2 initiates control room isolation and after a 42-second delay the control om emergency ventilation system will be operating at 2,500 cfm.

Isolation of the con I room will be complete within 5 seconds after reception of an isolation signal. Durin this 5 second interval the normal outside air flowrate through the damper was assum to vary linearly from 2,000 cfm to O cfm. The MP2 control 1

room emergency ventilati system recirculates air from inside the control room and filters the air through high e 'ciency particulate air (HEPA) and charcoal filters before returning it to the control roo Two separate cases were analyze for the design basis LOCAs at MP2. These cases are representative of high wind spee and low wind speed conditions. As described in Section 14.8.4.2 (Case B)it has been ssumed that the high wind speed condition exists for 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> after the LOCA and 0 percent of the activity in the enclosure building bypasses the EBFS resulting in a ound level release to 'the environment.

i Displacement of the enclosure building atm phere would begin at wind speeds above 25 mph. However, for conservatism,it is ass med that this displacement would begin with a wind speed of 23 mph. The low wind s ed conditions used assumptions consistent with those for Case 1 and given in Ta e 14.8.4-1, except for the 1.69 percent bypass leakage. For the control room nalysis, the bypass leakage was-reevaluated without assuming a seismic event and w s determined to be negligible.

The calculated whole body, beta and thyroid doses are p sented in Table 14.8.4-8 and are below the General Design Criterion 19 limits. For acci gnts at either Unit 1 or Unit-3 (and for several Unit 2 accidents not involving a signalto automatically activate the EBFS) the Unit 2 control room will not automatically isolat nd must rely on a high i

radiation signal to perform the isolation.

Under normal conditions, air is provided to the control room operato by the air intake duct. The duct is equipped with redundant radiation monitors which 'll automatically isolate the control room upon a high radiation signal. Approximately 23.1 seconds of continuous unfiltered air intake is assumed to enter the control room subsgquent to j

isolation by a signal from either radiation monitor. After 42 seconds the co trol room emergency ventilation system will be operating. Control room air will be reci ulated through HEPA and charcoal filters.

l Since other operating reactors are located on the site, an assessment was made the habitability of the Millstone 2 control room subsequent to an assumed design basis LOCA at either Millstone 1 or 3. Assumptions used for each of these plants are give l

in their respective FSARs. Because of the close proximity of the Millstone 1 turbine l

building with respect to the Unit 2 control room intake duct, an assessment was also i

made of a steam line break (SLB) accident at Unit 1 on the Unit 2 operators. The l

assumptions used in this accident are given in Table 14.8.4-7.

l MP214 8.MP2 14.8-15 June 1996 l

MNPS 2 FSAR L

The Iculated Millstone 2 control room dose from Millstone 1 and Millstone 3 released

'are pre nted in Table 14.8.4 8.

1 14.8.4.4 onclusions it is concluded t t the exclusion boundary and low population zone (LPZ) guideline dose values of 10 R 100 would not be exceeded even for the DBA.

l t

1 l

l l.

i t-l l-MP214-8.MP2 14,8-16 June 1996 l

MNPS-2 FSAR The following pages are " Insert D," Sections 14.8.4.1 through 14.8.4.5:

l l

i

14.8.4.1 General l

A LOCA would increase the pressure in the containment resulting in a containment isolation and initiation of the ECCS and containment spray systems. A SIAS signal l

automatically starts the Enclosure Building Filtration System (EBFS) which maintains a negative pressure within the enclosure building during accident conditions. The nuclide inventory assumed to be initially available for release from within containment consists of l

100 percent of the core noble gasses and 25% of the iodines, as described in Regulatory l

Guide 1.4. A SIAS also isolates the control room by closing the fresh air dampers within 5 l

seconds. Within 10 minutes after control room isolation, the control room emergency ventilation (CREV) starts. CREV recirculates air within the control room through a 90 percent charcoal filter at 2,500 cfm (i 10%) to remove iodines from the control room l

envelope.

The radiological consequences of a Design Basis LOCA at Millstone 2 were analyzed for a low and high wind speed condition. The low wind speed case was found to bound the high wind speed case. Therefore only the low wind speed case will be presented here.

14.8.4.2 Release Pathways The release pathways to the environment subsequent to a LOCA are leakages from containment and the enclosure building, which are collected and processed by EBFS and leakages from containment and the RWST which bypass EBFS.

Containment Leakage The containment is assumed to leak at the design leak rate for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the accident.

After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, since the pressure has been decreased significantly, Regulatory Guide 1.4 allows for the leak rate to be reduced to one-half the design leakage rate.

l All containment leakage for the first 110 seconds is assumed to bypass EBFS and is released through the MP-2 vent. This is due to the fact that it takes 110 seconds for EBFS to achieve the required negative pressure in the enclosure building, thereby ensuring that leakage will be into the enclosure building rather than out.

EBFS collects most of the containment leakage and processes it through HEPA and charcoal filters and releases it up the Unit I stack. All containment leakage is collected and filtered by EBFS except for the small amount that is assumed to bypass EBFS and is released out the MP-2 vent.

Credit is taken for iodine removal due to containment sprays. The sprays are effective within 3 minutes post-LOCA and are assumed to shutoff 30 minutes later.

ESF System Leakane Pathway l

Post-accident radioactive releases from the ESF system are derived from fluid leakages l

assumed during recirculation of the containment sump water through systems located i

e outside containment. The nuclide inventory assumed to be available for release from this pathway consists of 50%'of the core iodines. The quantity ofleakage _is based on the assumption that the ESF equipment leaks at twice the maximum expected operational leak rate and that 10 percent of the iodine nuclides contained in the leakage fluid become airborne in the enclosure building. The nuclides which become airborne are collected and released to the environment through EBFS to the Unit I stack.

RWST Backleakage Pathway Post-accident radioactive releases from the ECCS system are a result of ECCS subsystems containing recirculated sump fluid backleaking to the RWST. The backflow rate to the RWST, as a result ofisolation valve leakage, is pre-defined and time dependent. Due to this time dependency, the contaminated sump fluid from backleakage does not enter into

. the RWST until 25.45 hours5.208333e-4 days <br />0.0125 hours <br />7.440476e-5 weeks <br />1.71225e-5 months <br /> post-LOCA. Since the RWST is vented to atmosphere, the release is a result of the breathing rate of the RWST due to solar heating. The EAB dose is a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> dose therefore it is not affected by backleakage.

14.8.4.3 Control Room Habitability The radiation design objective of the control room is to limit the dose to personnel inside the control room to 5 rem whole body, or its equivalent, during a DBA. The potential radiation dose to a control room operator is evaluated for the LOCA. The analysis is J

based on the assumptions and meteorological parameters (X/Q values) given in Tables 14.8.4-3 and 14.8.4-4.

1 l

The control room is designed to be continuously occupied for the duration of the accident, 30 days. Two basic sources of radiation have been evaluated: leakage of airborne activity into the control room from sources described in 14.8.4.2 and direct dose from sources outside the control room The control room shielding serves to protect the operators from direct radiation due to the passing cloud of radioactive effluent assumed to have leaked from containment, enclosure building and the RWST. The control room walls also provide shielding protection for radiation emanating from the CREV filters and containment shine.

A SIAS from Millstone 2 initiates control room isolation within 5 seconds by securing the fresh air intake dampers. Within 10 minutes CREV is operation recirculating air in the control room envelope through 90% efficient charcoal filters to remove radioactive iodines from the atmosphere. The calculated thyroid, whole body and skin doses from a Millstone 2 LOCA are presented in Table 14.8.4-8 and are below the General Design Criteria 19 limits.

Normally outside air is provided to the control room via an air intake duct, which is equipped with redundant radiation monitors. These radiation monitors isolate the control room within 10 seco,nds after a high radiation signal. This method ofisolation will occur after a Millstone 3 LOCA. The calculated thyroid, whole body and skin doses from a Millstone 3 LOCA are presented in Table 14.8.4-8 and are below the General Design Criteria 19 limits.

~

14.8.4.4 Dose Computation The radiological off'-site dose consequences resulting from a postulated Millstone 2 LOCA are reported in Table 14.8.4-2. The off-site dose analysis show that the consequences to the EAB (0 - 2 hr) and LPZ (0 - 30 day) are less than the limits of 300 rem thyroid and 25 rem whole body as specified in 10CFR100. The assumptions used to perform the radiological analysis are summarized in Table 14.8.4-1.

14.8.4.5 Conclusion Analysis shows that the off-site radiological consequences are within 10CFR100 guidelines and the control room radiological consequences are within GDC19 criteria.

2 i

MNPS-2 FSAR TABLE 14.8.4.-1 g/q, LOSS OF COOLANT ACCIDENT [Off fyy g ASSuntr.)~Zows)

Assumption

@SV (1)

Core power level = WOO MWt 12)

Opc ct%; t".c - ? ;:.

2 Gir)

Core released fractions: Noble gases = 100%,lodines = 25%

3(f) llc;ege.u composition:

91% elemental Jodme 4% organic 5% particulate V

(l!i)

Reactor building leak rate:

.5%/ day.g_24 hrs.

.25%/ day > 24 hrs.

6 (g)

Enclosure Building Eiltration System charcoal filter efficiencies:

p 90% for elemental

- (EBfS) 70% for organic

)

90% for particulate

/, Od E-oy 3, (,c, f oy C

j,9 2,@ E-oS

y. go g.og

(/)

Bypass leakage fraction = +-69%

3,0Y/-oc d,3/ g.og 9

Dra wolown ~n'm k A* I? E ~o6 MOE-of (y)

EBFS negctive p.,c:;;u:c :nlt ctica = 110 seconds

/. O y' E-ot 7.15 8-d6 8

A43 6-01 ~

3.32 f.og (y). X/Os:

Location Time Period Elevated Ground Release j

-6B-EMS (0-2) hrs.

03 x 10-39 x 1 LPZ (0-4) hrs.

3.1x1

~5 2.2x (4-8) hrs.

1.7 x 0-*

2.1 10

(8 24) hrs.

2.62 104 4.7 108 (24-96) hrs.

1., x 04

3. 4 10-*

(96-720) brs.

7x1

.3x1

• 9 (JC) Thyroi? Inhalation DCFs from Rcg. 08; 1 100 _IC#FSo i

/0 3

(M) Containment Free Air Volume = * """

" ^^ '~ ' /,900 E AoG f1 11 i

()2) Breothing Rates -(0-8) hr. = 3.47 x 104 m*/sec (8-24) br. = 1.75 x 104 m*/sec

)

(24-720) hr. = 2.32 x 104 m*/sec A

IN SE2 7" A 5 4ss4.i.ue:

1 of 1 October 1994 l

=

1 INSERT A Assumption

12) Release Points:

Filtered - MPl Stack Bypass - MP-2 Vent

13) Containment Spray Removal Coefficients:

elemental = 3.827 per hour particulate = 1.707 per hour ~

14) Containment Spray Effectiveness Time: 3 - 33 minutes post LOCA 15)ESF Leakage: 24 gallons per hour 16)ESF leakage begins at 25 minutes post LOCA

17) Sump Volume: 2.86E+5 gallons 18)RWST Backleakage begins at 25.45 hours5.208333e-4 days <br />0.0125 hours <br />7.440476e-5 weeks <br />1.71225e-5 months <br />

19) RWST Backleakage amount: 0.01 - 019 gpm

20) Iodine DF: 100 i

i l

MNPS-2 FSAR TABLE 14.8.4-2 l

r/qc

SUMMARY

OF DOSES FOR LOSS OF COOLANT ACCIDENT DOGE (rems)

/

CASE A CASE B ORGAN N

BOU DARY PQ HYDROGEN PURGE BOU DARY l

y NOTINCLUDED NOTINCLUDED Thyroid 56.3 12 Whole Bodyp 3.8 1.4 1.4 x

{

/

N 1

f

[

l

'J-h i s;)

b)N:/c bl y

y

_l ggg y.STf40/

h 9.YG'E+00 L f ~f 9./9 E -f of 9Y/ E~o/

\\

\\

^

6 14tB4 2.MP2 1 Of 1 October 1994l

I MNPS-2 FSAR i

j TABLE 14.8.4 -3 i

ggg g og coaut7 (CONTROL ROOM ASSUMPTIONS)

Accr oExq 1.

Control Room Volume = 7.7 x i O' ii' 3.56 f g / cv //#

l 130 1

2.

Control Room Unfiltered inleakage in Recirculation Mode = 4GO cfm 100 3.

Control Room Normal Makeup Air Flowrate = EGOO cfm InP-3 loco 4.

Time from Start of.^::it:nt to Time when Dampers Close = 5 sec.

% e fn.n Co.,ti / Ram hoyA ra lia/uv A 7) e u,A r< ba pus closc : /o Sec.

j, 5.

Time when Control Room Emergency Ventilation, System Operating at Full Speed = 42 xc.

Control Room Emergency Ventilation (System Flor.* ate = +;500 cfm vour++p

.;t.aso

/0 inw.

6.

7.

Charcoallodine Filter Efficiency = 90 percent

• NOTES: For t analysis of assumed A at MP2 For accide s at either Unit 1 or 3 damper clos time is 23.1 sec. to ount for monitor respop and damper closure time. Ot r Unit 1 and Unit 3 ass ptions are given'in the f owing references, f

REF ENCES:

1. Unit 1 - W. G. Cou il,1981 (NUSCO) Letter to. M. Crutchfield (NRC), ansmitting Millstone Nuclea ower Station Unit 1, Syste atic Evaluation Progra.Section XV, Topics: Desig asis Events.

2. Unit 3 -

illstone Unit No. 3 FSAR. \\

7.

Oypess l ea h A m ou,,+

// cc/hr-

=

\\

$s 0 CAY'Jl $0cm $btcl$//)y l f4/cr ff (z)q //,* g'

(,, f(

W'5 6 **// !

/,:r ' c,ocre/c c e r p A

? 's ec//wr wh.e 4 tr 9 ' concsc fe c

S"Ih Wall M. S 'o / / 'co,ncich ncyl 9 l* 5's we // V(o ):('/o<ry Eats F Wa // !

.2 ' Con c/e f.

l 000f l d ' Con cre /t Flow.'

/ ' c m e<e h

)

l l

34sa4.s uP2 1 of 1 October 1994 I

l l

MNPS-2 FSAR l

TABLE 14.8.4 4 ATMOSPHERIC DISPERSION DATA FOR MILLSTONE UNIT 2 CONTROL ROOM N

i MP-2 CONTROL ROOM X/Qs (sec./m*)

Release Point

/

MP-1

/MP-2 MP-[

a.

Ground L el l

(0-8) hr.

4.43 x 10 2 2.69 x 10 3

.78 x 10 4 (8-24) r.

3.05 x 10~2 1.90 x 10~8 3.29 x 104 (1-day 1.20 x 0~2 7.56 x 10" 1.20 x 104 l

4 30) day 3.

x 10'2 2.3C x 10" 2.26 x 10-5 (0-24) Mr.

9.14 x 1 N/A (24 36)hr.

N/A 9.14 10~5 N/A

b. Elevated Release - nit 1 Stack (0-8) hr.

2.0 x 10~'

2.0 x '10-'

N/A (8-24) br.

1.0 x 10-8 1.0 x 10

N/A (1 4) 2.5 x 10

2.51 x 10-'

N/A (4-

) day 5.01 x 10

5.01 x 10'"

A y

0 I

insee 4 up:

1 of 1 October 1994l

1 i

L INSERT B Release Point MP-2 RWST:

0 - 8 HR:

1.87E-3 8 - 24 HR:

1.20E-3 1 - 4 DAYS: 3.83E-4 4 - 30 DAYS: 5.81E-5 MP-2 Enclosure Building:

' 0 - 8 HRf 5.46E-3 8 - 24 HR:

3.45E-3 1 - 4 DAYS: 1.27E-3 4 - 30 DAYS: 3.98E-4 MP-2 Vent 0 - 8 HR:

2.92E-3 8 - 24 HR:'

l.89E-3 1 - 4 DAYS: 6.18E,

4 - 30 DAYS: 1.05E-4 MP-1 Stack 0 - 4 hrs:

2.51 E-4 4 - 8 hrs:

1.96E-5 8 - 24 hrs:

5.46E-6 24 - 96 hrs:

2,06E-7 96 - 720 hrs: 2.58E-9 MP-3 Containment 0 - 8 hr:

9.19E-4 8 - 24 hr:

5.29E-4 24 - 96 hr:

1.65E-4 96 - 720 hr:

2.75E-5 MP-3 Ventilation Vent 0 - 8 hr:

1.25E-3 8 - 24 hr:

7.49E-4 24 - 96 hr:

2.46E-4 96 - 720 hr:

4.08E-5 MP-3 MSVB l

0 - 8 hr:

2.47E-3 8 - 24 hr:

1.48E-3 l

24 - 96 hr:

4.87E-4

k i

i i

~

96 - 720 hr:

8.18E-5 MP-3 ESF Bldg 0 - 8 hr:

2.08E-3 8 - 24 hr:

1.18E-3 24 - 96 hr:

3.88E-4 i

96 - 720 hr:

6.12E-5 9

8 2

1 a

Y 4

4 4

k

MNPS-2 ?SAR TABLE 14.8.4-5 hSSUMPTIONS USED TO CALCULATE DOSES FROM EXTERNAL SO 1.

As mptions Used to Calculate Dose from Containment Source:

i a.

ource Term:

100 percent core noble gas inventory rel sed to containment 1

50 percent core iodine inventory rei sed to containment.

l b.

Source ssumed to be uniformly distributed to co ainment free air volume.

c.

Containm t Volumes:

Unit 1 = 2.

8 x 10 ft 5

Unit 2 = 1.89 x 10' f t*

d.

Containment Con ete Wall Thicknes :

Unit 1 = 5' Unit 2 = 3.75'(walls) 3' (done e.

Control Room Concrete ll hickness:

For wall facing Unit 1 co ai ent = 3'-6" For wall facing Unit.2 e ntain nt =. 2'-0" 2.

Assumptions Used to Cal late Dose fr Enclosure / Reactor Building Source:

a.

Containment Lea Rate:

Unit 1 = 1.2 ercent/ day Unit 2 = 0.

percent / day

)

1 b.

Volume Enclosure / Reactor Building Unit 1 eactor Building = 1.728 x 10' ft Unit Enclosure Building = 1.44 *x 10' ft c.

V ntilation Rate Unit 1 Reactor Building = 100 percent / day Unit 2 Enclosure Building = 6.000 cfm Control Room Wall Thickness:

I Control Room Wall Facing Enclosure Building = 2'-0" i

Control Room Wall Facing Reactor Building = 3'-6" i

)

meaup:

1 of 2 October 1994 k

MNPS-2 FSAR TABLE 14.8.4-5

.SSUMPTIONS USED TO CALCULATE DOSES FROM EXTERNAL SOURCE 3.

Assu tions Used to Calculate Dose from Filtration Systems:

a.

Filtr ion Systems Considered:

Mi stone Unit 1 Standby Gas Treatment System GTS)

Mill one Unit 2 Enclosure Building Filtration Si tem (EBFS)

Millst e Unit 2 Control Room Filters b.

Thickness of C crete Between Control Room d:

Millstone Un 1 SGTS = '9'-0" Millstone Unit EBFS = 18'-0" Millstone Unit 2 ontrol Room Fil r = 2'-0" 4.

Assumptions Used to Calculat Dose fro Overhead Plume:

a.

Millstone Unit 2 Contro1 Roo iling Concrete Thickness = 2'-0" b.

Filtration Sysi em Filter Effici n s:

t SGTS = 90 perce (all form of iodine)

EBFS = 90 perc nt (element and particulate idtfine)

= 70 per ent (oiganic io 'ne) c.

Plume Centerline /Os:

(0-8) hr. = 4. 4 x 10'* sec/m (8 24) hr.

4.19 x 10-* sec/m8 (1-4) da = 1.65 x 104 sec/m 8 (4 3 day = 9.92 x 10' 5.

Assum ions Used to Calculate Dose from Piping Sources:

Sources in the vicinity of the Unit 2 Control Room = Unit 1 core spray line, a.

b Source Term:

50 percent core iodine inventory 1 percent solid fission products c.

Concrete thickness of Control Room wall = 3'-S" 14 S84-5.MP2 2 of 2 October 1994 l

l.

MNPS-2 FSAR TABLE 14.8.4-6 UMMARY OF DOSES FROM EXTERNAL SO ES (1) 5/40 30-D DOSE (MREMS) TO MI TONE UNIT 2 ONTROL ROOM ERATORS UNIT 1 ACCIDENT OVERHEAD DIRE FRO SECONDARY CORE SPRAY TYPE PLUME PRI RY CONT.

CONTAINMENT LINE TOTAL l

Millstone 7.44 10 3.288 x 10 '

3.493 x 10 4.846 x 10' 5.972 x 10' Unit 1 LOCA Millstone 3.424 x 10" 4.044 x 10' 5.214 x 102 9.292 x 102 l

Unit 2 LO l

)

1 l

1

)

1 14 S B4-6.MP2 l of l QCtOber 1994 l

1

- - -. - ~.---- -.

. _ _ _ =

MNPS-2 FSAR TABLE 14.8.4-7 A

UMPTIONS USED IN A MILLSTONE UNIT 1 MAIN STREAMLINE BAK 1.

Mass of C lant/ Steam Released = 1.753 x 10' gm 2.

' Coolant Conce ration:

DEQ (1 131).= 0.2 icro Ci/gm -

Noble Gas ='100/E 'cro Ci/gm 3.

Duration of Release = 5.

sec.

4.

MP-2 Control Room Normal V tilation stem Flowrate = 2,000 ft / min. (Note:

8

. This flowrats assumed for entire

.5

c. since monitor response and damper closure = 2L' 1 sec.)

5.

Time for MP-2 Control Room. ntilatio System to Operate at Full Speed = 42 sec.

6.

MP-2 Recirculation Sy m Flowrate = 2,50 fm 7.

Tirne When Opera rs are Assumed to Don Scott ir Paks = 20 minutes 8.

Effectiveness f Scott Air Paks = 10,000

.i 9.

Time Wh Operators are Assumed to Purge the Control R m = 30 min.

10.

Purg ime Span = 30 min. to 4 hrs.

11.

rge Flowrate = 16,500 1

Control Room X/O (0-8 hrs.) = 4.43 x 10'2 sec/m S

I l

14 S84-7. MP2 I II October 1994 l

MNPS-2 FSAR 5

(

TABLE 14.8.4)(

DOSE TO MILLSTONE UNIT 2 CONTROL ROOM OPERATORS Whole Body (l)

Beta Thyroid Dose Gamma Dose Skin Dose Release (Rems)

(Rems)

(Rems)

Mills e 1 (1 OCA)(2)

)<27 x 10' 7.

x 10 2

.45 x 10-'

j) illstone 1

'i 2.62 x 10' 4.23 x 10-'

5.87 1

Millstone 2 (LOCA) 0,25 x 10"

.e M 5,,10 '

- 2,00 x ; 0 '

Luvv V'..A Svuod Cundidunk

),Sg xio' 9, 7 g,,o /

3,3 y,,o

~

o Wind S ee ons)

/

Millstone 3 (LOCA)

-2.40 ^ i G'

-2.00 x 10'

-2.G7 x i G^

0 * / S X /O '

/ WY.t/o*

/.yo xjo' NOTES:

(1)

Dose through wall and ceiling from external sources included.

d for Tu[Buildin%st t me o s is a

]

t I

{

}

14884 8 MP2 I of 1 October 1994 l

_. _. _. ~.. _ _ _ _...... _ _ - _ _.. _ _...

L Docket No. 50-336 B17413 l

l Millstone Nuclear Power Station, Unit No. 2 4

l Proposed Revision to Technical Specifications Control Room Ventilation System Requests for Additional Information l

l l

September 1998 j;

l

l l

I U. S. Nuclarr Regulttory Commission B17413/Attrchmant 6/Page 1 Proposed Revision to Technica! Specifications Control Room Ventilation System Requests for Additional Information in a letter dated November 6,1997,W the NRC requested the following additional information.

l Question 1 l

Calculation M2CRM2-01156-R2 evaluates the control room doses from a Design Basis l

Accident Loss-of-Coolant Accident. A containment bypass fraction of 4.025E-7 is determined on page 9 of the calculation.

This value was subsequently used to 4

determine the containment bypass release rate of 8.388E-11 hr for the period T = 110 seconds to T = 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Millstone 2 Technical Specification 3.6.1.2 establishes a bypass limit of 0.017 x LA (LA = 0.5%/ day). The staff believes that the appropriate release rate to be:

4 0.5%/ day x 1/100 x 0.017 x day /24 hr = 3.54E-6 hr The methodology used in the Millstone 2 calculation differs from the corresponding evaluation in Millstone 3 calculation M2CRM3-01146-R2, which d

developed a bypass leak rate of 0.0278%/ day or 1.16E-5 hr.

A comparison of the staff exclusion boundary area and low-population zone analyses results with those documented in the Updated Final Safety Analysis Report suggests that the suspect release rate was not used in the evaluation of offsite doses.

Justify the use of the 4.025E-7 value when the containment could be considered operable with a bypass fraction as high as 0.017.

If the bypass fraction assumption cannot be supported, the affected calculation should be revised and provided to the staff.

Reply The current licensing basis for Millstone Unit No. 2 is presented below. (it is important to note that this information is based on historical documentation. It may not accurately reflect the current configuration of Millstone Unit No. 2.

However, the conclusion reached is correct for the current Millstone Unit No. 2 configuration.)

Based upon the Atomic Energy Commission (AEC) requirements set in the N

D. G. Mcdonald Jr. letter to NNECO, " Request for Additional Information Relating to the Control Room Ventilation System, Millstone Nuclear Power Station, Unit No. 2 (TAC NO.

M92879)," dated November 6,1997.

U. S. Nuclur Regulatory Commission B17413/ Attachment 6/Page 2 May 10,1974,* Safety Evaluation Report (SER) and the correspondence leading up to the SER, the Control Dose assumptions of negligible bypass leakage (6.44x10* scfm) were not questioned while a bypass leakage of 1.7% of the daily containment leakage rate was mandated for calculating the 10 CFR 100 offsite doses. The licensing basis for calculating the control room dose following an i

event is therefore 6.44x10* scfm.

i

Background

l In a letter dated December 29,1972,m the AEC sent the Millstone Point Company a Request for Additional Information, in the form of a series of questions, to support the AEC review of the Millstone Unit No. 2 Final Safety Analysis Report (FSAR). AEC Questions 5.39 and 14.1 requested the following information:

5.39 Provide a detailed evaluation of penetration through-line leakage that can bypass the enclosure building. Detail each leakage path and describe provisions made for initial and periodic testing of these penetrations to measure leakage. Provide proposed Technical Specification limits for allowable leakage through these penetrations.

l 14.1 Discuss the dose calculational model used in Section 14.18.3.3 of the FSAR to determine the gamma and beta doses from iodine and noble gases inside the control room and report these doses separately.

Differentiate the leakage and the doses associated with the leakage which bypasses the enclosure building from the leakage and the doses associated with the leakage which is filtered and passed out of the stack.

In a letter dated February 16, 1973,W the Millstone Point Company sent Amendment 15 to the MP2 License Application to the AEC.

A response to questions 5.39 and 14.1 were included in this amendment:

5.39 - Response To evaluate the through-line leakage that can bypass the enclosure building filtration region (EBFR), the fluid systems penetrating containment are categorized as follows:

i 1.

Piping System open to the containment post-incident atmosphere.

W O. D. Parr (AEC) letter to The Millstone Poirt Company, Safety Evaluation for Millstone I

Nuclear Power Station, Unit 2, dated May 10,1974.

W K. R. Goller (AEC) letter to The Millstone Point Company, Request for additional Information, dated December 29,1972.

W The Millstone Point Company letter to AEC, " Amendment No.15 To Licensee Application in Docket No. 50-336," dated February 16,1972.

l

i U. S. Nuclear Regulttory Commission B17413/ Attachment 6/Page 3 2.

Piping Systems which are closed and therefore not exposed to the containment post-incident atmosphere.

The following assumptions are made to postulate the maximum hypothetical i

conditions:

l l

1.

There is either a seismic occurrence and all Seismic Class 2 lines are broken or either there is no seismic occurrence and all Seismic Class 2 lines are intact.

2.

The single failure applies to Seismic Class 1 components only, f

The line from the normal sump to the aerated drain tank could provide a potential path for bypass leakage. Assuming the leakage through the valve is proportional to the square root of the pressure differential, the maximum leakage l

through the two (2) series containment is 3 cc/hr at containment post-incident l

design conditions. After approximately one (1) hour the containment pressure is less than 10 psig and the leakage rate is less than 1.0 cc/hr. This valve leakage is diluted in the aerated waste system.

The refueling water purification penetrations could provide a potential path of bypass leakage into the auxiliary building. The maximum leakage is approximately 4.0 cc/hr through each penetration during the first hour. After this the maximum leakage is approximately 1.0 cc/hr. Assuming normal system I

alignment (Figure 9.5-1), the leakage is diluted and contained within the closed l

process piping.

From the preceding analysis, only the leakage through the normal sump to aerated drain' tank and refueling water purification penetrations may be considered as through line leakage. The maximum potential leakage rate is 11.0 cc/hr during the first hour and less than 3.0 cc/hr thereafter. All other leakage is either contained within the process system or within the EBFR.

The provisions for initial and periodic leak testing of containment penetrations and maximum allowable leakage are specified in Subsection 15.4.5. and Table 5.2-11 of the FSAR.

I' l

l,-

U. S. Nucicer Regulatory Commission B17413/ Attachment 6/Page 4 14.1 - Response The iodine and noble gas concentrations in the control room are calculated as a function of time. The thyroid and whole body doses for exposure to a semi-infinite cloud at these concentrations are calculated using the inhalation model (dose conversion factors tables 14.1-1 and 14.1-2) of TlD14844, and the semi-l infinite cloud model given in safety guide 4, and in Meteorology and Atomic Energy-1968. Additional information is given in Subsection 14.18.3.3.

The thyroid and whole body doses inside the control room previously stated in Subsection 14.18.3.3 were based isolating the control room 20 minutes after incident. The reduced doses given above results from design change in which control room is isolated on a EBFS or AEAS signal.

All leakage from the containment will be into the enclosure building. Since the enclosure building is kept at a slightly negative prer;' a all leakage from the containment will be exhausted through EBFS filters to ti.e Unit 1 stack. Thus, the leakage and doses previously given are from the stack releases.

In a letter dated immy 21, 1973,* the AEC sent the Millstone Point Company another Request for Additional Information, also in the form of a series of questions, to support the AEC review of the Millstone Unit No. 2 FSAR.

j Questions 6.16.1, 6.16.2, 6.16.3, and 6.16.4 concerned the possibi

.eakage of j

radioactive materials from containment:

6.16 Provide the following information concerning the containment leakage.

6.16.1 List each potential leakage path which presents a potential pathway for 4

release of radioactivity from the containment atmosphere (1) directly to the external atmosphere, (2) to the auxiliary building, and (3) to the enclosure building filtration area.

6.16.2 Provide an estimate of'the fractions of the total amount of containment leakage which can bypass the enclosure building filtration area and be released (1) directly to the atmosphere and (2) to the auxiliary building.

I Describe the tests and test frequencies that will be used to detect and l

linit these leakage fractions and the total containment leakage.

6.16.3 Provide the Technical Specification limit that assures that the leakage through potential leak paths that bypass the enclosure building filtration 4

• K. R. Goller (AEC) letter to The Millstone Point Company, Request for additional Information, dated May 21,1973.

U. S. Nuclur Regulatory Commission B17413/ Attachment 6/Page 5 area will not exceed the fraction of total leakage assumed in the dose calculations.

l I

I 6.16.4In the response to item 5.39 of Amendment 15, a seismic occurrence is not considered in the evaluation of through-line leakage that can bypass the enclosure building. Assuming a seismic occurrence, (1) describe any non-Category I system which could become open to the containment atmosphere, and terminate outside the enclosure building; (2) provide the information requested in 6.16.1 through 6.16.3 above for these additional l

pathways.

In a letter dated June 27,1973,5 the Millstone Point Company sent Amendment 16 to the MP2 License Application to the AEC.

Responses to questions 6.16.1 through 6.16.4 were included in this amendment:

6.16.1 - Response The detailed evaluation of penetration potential through-line leakage that could bypass the enclosure building filtration region (EBFR) was provided in the response to AEC Question 5.39 in Amendment 15. However, to supplement that response, containment potential leakage paths are again evaluated, as equested, on an individual case basis.

The following is the basis formulated for the analysis:

1.

There is no seismic event, therefore all systems remain intact.

2.

The model for liquid valve seat leakage through a closed valve is 2.0 cc/hr. per inch nominal valve diameter and 1.0 cc/hr. per inch diameter for two closed valves in series. The basis for this model is described in response to AEC Question 5.39 in Amendment 15.

3.

The model for gaseous valve seat leakage through a closed valve is 0.10 SCFH per inch of nominal valve diameter and 0.05 SCFH per inch for two valves in series. The former is the acceptance criterion per Manufacturers Standardization Society SP-61, Hydrostatic Testing of Steel Valves,1961 Edition.

The Millstone Point Company letter to AEC, " Amendment No.16 To Licensee Application in Docket No. 50-336," dated June 27,1973.

U. S. Nuciar Reguirtory Commission B17413/ Attachment 6/Page 6 Attached to this response is Table 6.16-1 which states:

Potential Containment Leakage Paths Leakage Path to Auxiliary Bldg Penetration No.14, Normal [ Containment] Sump 4

Rate: 1.76x10 [SCFM]

Penetration No. 67, Refueling Water Purification Rate: 2.34x10* [SCFM]

Penetration No. 68, Refueling Water Purification 4

Rate: 2.34x10 [SCFM]

Total Leakage Rate to Auxiliary Bldg.: 6.44x10* [SCFM)

Notes:

1.

Potential containment leakage paths following a LOCl without assuming a seismic event.

2.

Rate is expressed in units of standard cubic feet per minute.

3.

The percentage of leakage through the given path compared to the assumed containment leakage rate during the first day following the incident.

6.16.2 - Response The analysis for the design basis incident (DBI) is described in Section 14.18 of the FSAR. A containment leak rate of 1.5 volume percent per day was assumed i

during the first day and 0.75 volume percent per day thereafter as discussed in FSAR Subsection 14.18.2. The containment leakage is calculated as 48 SCFM thereafter.

In order to tabulate liquid leakages as a percentage of the total containment leakage a model is formulated. Due to the minute liquid leakages it is conservatively assumed that all the liquid evaporates into a vapor of the same constituents as the containment atmosphere. Therefore, these liquid leakages can be considered as vapor leakages.

The percentages of containment leakages which can be released directly to the atmosphere, auxiliary building and enclosure building filtration region (EBFR) shown in the preceding Table 6.16-1.

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U. S. Nuclear Regulatory Commission B17413/ Attachment 6/Page 7 Attached to this response is Table 6.16-2 which states:

Potential Containment Leakage Paths Leakage Path to Auxiliary Bldg Total Leakage Rate: Negligible Notes:

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1.

Potential containment leakage paths following a loss-of-coolant incident assuming a seismic event.

2.

Rate is expressed in standard CFM.

6.16.3 - Response The amount of containment leakage bypassing the EBFR is negligible as shown in the preceding Table 6.16-1. The potential containment post-incident leakage rate into the EBFR is calculated as 0.8 SCFM or approximately 100 lbs. during l

the first day. The allowable post-incident containment leakage rate per Technical l

Specification 15.4.5.A.2.a of the FSAR is approximately 1750 lbs. during the first day (approximately 1050 lbs. during the first day is allowed for the analyzed containment penetrations). The post-incident containment leakage rate assumed for the dose calculations was approximately 5200 lbs. during the first day.

Therefore, the amount oi leakage assumed for the post-incident dose calculations is approximately five times the Technical Specification limit and more than fifty times that calculated. The dose calculation assumed all potential containment post-incident leakage to be into the EBFR.

The detailed analyses in response to Question 5.39 and 6.16.1 in this j

Amendment indicate that the only potential leakage that can bypass the EBFR is a minute quantity of liquids. T.hese leakages in actuality will be diluted into the process system fluids and therefore will be contained in that system. On this basis and on the results of the preceding analyses it is not deemed necessary to impose Technical Specification limitations on these three penetrations.

6.16.4 - Response The analysis in the response to AEC Question 5.39 was based on the assumption that either there is a seismic occurrence and all Seismic Category 2 lines are broken or there is not a seismic occurrence and all Seismic Category 2 L

lines remain intact. The former case was not considered in the above analysis since any line that is Seismic Category 2 within containment was also Seismic Category 2 outside containment within the EBFR. Should any Seismic I.

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U. S. Nuciser Regulatory Commission B17413/ Attachment 6/Page 8 Category 2 lines break within containment, these must also break outside containment. Since all cor:tainment penetrations pass through the EBFR prior to entering any other area, any containment leakage due to a broken line must then be vented to the EBFR. Therefore, for a valid analysis of throughline leakage the conservative approach should assume that Seismic Category 2 lines remain intact, as was done.

However, as requested, a re-evaluation of potential containment leakages on an individual penetration basis assuming that all Seismic Category 2 lines are broken following the postulated seismic event was done. The penetration leakage models as formulated in the response to AEC Question 6.16.1 in this Amendment are valid.

As discussed in the response to AEC Question 6.16.2, the percentages of containment leakages which can be released directly to the atmosphere, auxiliary building and EBFR are shown in the preceding Table 6.16-2.

The above information establishe.s the Millstone Point Company position that leakage bypassing the containment was considered to be negligible. Both offsite and control room doses were originally calculated using this assumption. On May 10,1974,m the AEC produced the original SER for Millstone Unit No. 2.

Within the SER, the following statements were made concoming Control Room Habitability and Offsite Dose Calculations:

6.5 Habitability Systems We have calculated the potential radiation doses to control room personnel following a LOCA. The resultant doses are within the guidelines of GDC 19. On this basis, we conclude that the design of the control room ventilation system is acceptable for the purpose of preventing significant radiological exposures to operating personnel.

15.2 Loss-of-Coolant Accident Dose Model Unit 2 is a pressurized water reactor with a low leakage concrete primary containment and a sheet metal secondary structure forming an enclosure building. The enclosure build lng is maintained at a negative differential pressure after the postulated design basis LOCA. This assures treatment of most released activity by filtration systems and release from an elevated stack.

M O. D. Parr (AEC) letter to The Millstone Point Company, Safety Evaluation for Millstone Nuclear Power Station, Unit 2, dated May 10,1974.

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U. S. Nucicer Regulatory Commission B17413/ Attachment 6/Page 9 Upon receipt of a high radiation signal, filtration units will be activated such that the enclosure building is drawn down to a negative differential pressure within one minute. In analyzing the capability of the proposed enclosure building to further minimize the direct outleakage of fission products, the staff considered three specific points: (1) the minimum negative differential pressure throughout the enclosure building, (2) the amount of time required to achieve a minimum negative differential pressure.of.25 inches water gauge within the enclosure l

building, and (3) the fraction of the primary containment leakage that could bypass the filtration system and be released directly to the atmosphere.

We have evaluated the accident analyses for Unit 2 and have determined that the applicants' dose calculations were not consistent with staff reviews of other l

dual containment systems. Accordingly, we have calculated the loss-of-coolant l

accident (LOCA) dose for Unit 2 including an incremental dose resulting from the first minute after the postulated design basis loss-of-coolant accident. During I

this first minute, the staff assumes a direct ground level release from the l

containment with no holdup or filtering of fission products. After the first minute, l

the enclosure building region has reached a minimum negative differential l

pressure of 0.25 inches water gauge, sufficient to assure the retention and i

filtering of fission products assumed to be leaking from the reinforced concrete containment.

l in order to limit the total dose to well within 10 CFR Part 100 guidelines we have advised the applicants that we will require an integrated containment leak rate of l

0.5 percent per day. At the containment leak rate proposed by the applicants (1.5 percent per day), the total. dose to the thyroid at the site boundary, including the incremental dose from the first minute and the dose resulting from bypass i

leakage, exceeds the guideline value specified by 10 CFR Part 100. It should be noted that based on a leak rate of 0.5 percent per day, the bypass leakage function was taken as 1.7 percent per day of the containment leak rate.

L in order to achieve doses well within 10 CFR Part 100 guidelines, the staff will l

require the applicant to lower the containment leak rate to 0.5% per day. We believe that this value can be easily met. As can be seen from Table 15-1, the LOCA doses, assuming a primary containment leak rate of 0.5%/ day, are well within the guidelines of 10 CFR Part 100.

Conclusion Based upon the AEC requirements set in the May 10, 1974* SER and the correspondence leading up to the SER, the Control Dose assumptions of negligible bypass leakage (6.44x10* scfm) were not questioned while a bypass O. D. Parr (AEC) letter to The Millstone Point Company, Safety Evaluation for Millstone Nuclear Power Station, Unit 2, dated May 10,1974.

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U. S. Nucle r R::gulttory Commission B17413/ Attachment 6/Page 10 4

leakage of 1.7% of the daily containment leakage rate was mandated for calculating the 10 CFR 100 offsite doses. The Licensing Basis for calculating the Control Room dose follow!ng an event is therefore 6.44x10* scfm.

Question 2 The release activity in the Millstone 1

main steamline break analysis appears to be based, in part, on Appendix i source terms. The staff does not believe that this is an appropriate source of isotopic fractions for the design basis calculation.

Appendix I source terms are typically based on projections of releases over a year and often discount nuclides with short half-life by including a short period of decay. The isotopic fractions are being used in an analysis involving the release of fresh reactor coolant system (RCS) activity. The Millstone 1 Updated Final Safety Analysis Report does not contain a listing of the design basis RCS activity.

Provide a listing of the Millstone 1 design basis RCS activity for the halogens, xenons, and kryptons; and a brief description of the basis (e.g., percent failed fuel, x.x Cilgm, etc.).

Repiv Millstone Unit No.1 design basis accidents, loss of coolant and main steam line break, will no longer be evaluated for impact on Millstone Unit No. 2 control room habitability.

This credits the decision to decommission Millstone Unit No.1.8) a U

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B. D. Kenyon letter to the NRC, " Millstone Nuclear P wer Station, Unit No.1 Certification of Permanent Cessation of Power Operations and tha'. Fuel Has Been Permanently Removed from the Reactor," dated July 21,1998.

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U. S. Nuclur Reguintory Commission B17413/ Attachment 6/Page 11 In a letter dated November 25,1997,"* the NRC requested the following additional information.

Question 1 The proposed Technical Specification Bases contains the position that the nominal recirculation system flow rate of 2500 cubic feet per minute (cfm) can be used in lieu of the minimum acceptable flow of 2250 cfm since the range of flow fluctuation is overwhelmed by other conservatisms found in the control room dose calculations. The staff finds this position to be unacceptable.

If the flow value of 2250 cfm is in some way limiting, the proper approach is to explicitly identify the site-specific analysis assumption value deemed to be conservative that can be modified to compensate for the reduced recirculation flow while still maintaining an adequate margin of safety.

The staff requests that the licensee delete this language from the proposed amendment.

Hep.!g The control room dose calculations have been revised and now use a value of 2250 cfm. The proposed change to the Bases of Technical Specification 3.7.6.1, " Plant Systems - Control Room Emergency Ventilation System," states that the minimum flow of 2250 cfm is used in the associcted calculations.

Question 2 In the licensee's analysis of the Unit 3 loss-of-coolant accident, it was assumed that the activity released to the containment and available for release to the environment was 25% rather than the 50% assumed in prior versions of the calculation. However, the licensee's calculation tabulated spray parameters that are incompatible with the assumption of 25% core inventory. The difference between the assumed activities available for release is the method for crediting plate out of the activity released from the Reactor Coolant System (RCS). The spray parameters tabulated in the licensee's calculation appear to be based on Revision 2 to the Standard Review Plan (SRP) 6.5.2.

However, it is the staff's position that Revision 2 to the SRP is appropriate only for use with an assumption of 50% core inventory available for release. In assuming 25% core inventory and the spray parameters used in the calculation, the licensee is crediting plate out twice - the first being the 50% deterministic credit and the second being the plate out lambda of 3.1 hr". Revision 0 of SRP 6.5.2 provided for an assumption of 25% core inventory available for release. However, this revision also limited the spray lambda to 10 hr" with an overall iodine decontamination factor limited to 100.

"* D. G. Mcdonald Jr. letter to NNECO, " Supplemental Request for Additional Information Relating to the Control Room Ventilation System, Millstone Nuclear Power Station, Unit No. 2 (TAC NO. M92879)," dated November 25,1997.

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U. S. Nuclear Regulatory Commission B17413/ Attachment 6/Page 12 The staff requests that the licensee correct this deficiency in the analysis.

Reply Calculation #2 UR(B)-453, MP-2 Control Room Operator Doses Following a MP-3 LOCA, assuming Duct Leakage and Damper Bypass, uses Standard Review Plan (SRP) 6.5.2 Rev 2 and Regulatory Guide (RG) 1.4 in calculating doses. In accordance with the SRP and RG, the calculation assumes instantaneous plateout of 50% of elemental iodine initially released from the core and takes no credit for elemental iodine removal after a DF of 200 is reached relative to the amount of elemental iodines initially released from the core (i.e. the DF of 200 is relative to the initial release of 50% or a i

DF of 100 relative to the 25% iodines remaining after instantaneous plateout).

i The Millstone Unit No. 2 LOCA calculations, which now take credit for sprays, starts with the instantaneous plateout of 50% and only assumes sprays operate for 30 minutes. The elemental iodine DF reached is 6.78 of the 25% iodines remaining after instantaneous plateout or 13.56 relative to the initial release of 50% of the elemental lodines.

l Question 3 In the analysis of the Unit 1 main steamline break (MSLB), the licensee did not assess the dose consequences of an MSLB with the RCS activity at the maximum technical specification value of 4.0 pCi/g. This is provided for by Safety Guide 5, and by SRP 15.6.4. The licensee's analysis took the position that assuming a preincident spike was unnecessary because of the low probability of an MSLB accident in the 48-hour period when coolant activities are at 4.0 Ci/g. This position is not acceptable, j

Since the preincident iodine spike value represents an increase by a factor of 20 in concentration level, it is unlikely that doses will be acceptable.

The staff requests that the licensee provide an analysis of the control room dose associated with an MSLB with RCS concentration at 4.0 pCi/g dose equivalent 1-131.

Reply Millstone Unit No.1 design basis accidents, loss of coolant and main steam line break, will no longer be evaluated for impact on Millstone Unit No. 2 control room habitability.

This credits the decision to decommission Millstone Unit No.1.00 i

"" 8. D. Kenyon letter to the NRC, " Millstone Nuclear Power Station, Unit No.1 Certification of j

Permanent Cessation of Power Operations and that Fuel Has Been Permanently Removed j

from the Reactor," dated July 21,1998.

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