ML20071P456

Proposed Tech Spec Allowing Startup & Power Operation W/One Inoperable Reed Switch Position Indicator Per Control Element Assembly Group & Expanding Allowable Vol Band for Safety Injection Tanks

ML20071P456
Person / Time
Site:	Millstone
Issue date:	10/31/1982
From:	NORTHEAST NUCLEAR ENERGY CO.
To:
Shared Package
ML20071P451	List: A02456, Application to Amend License DPR-65,revising Tech Specs to Allow Startup & Power Operation W/One Inoperable Reed Switch Position Indicator Per Control Element Assembly Group ML20071P456
References
TAC-49061, NUDOCS 8211020519
Download: ML20071P456 (39)

Text

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Docket No. 50-336 1

I Attachment 1 Millstone Nuclear Power Station, Unit No. 2 1

Proposed Revisions to

+

Technical Specifications b

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4 October,1982 8211020519 821022 PDR ADOCK 05000336 PDR p

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ri REACTIVITY CONTROL SYSTEMS POSITION INDICATOR CHANNELS (Continued)

LIMITING CONDITION FOR OPERATION b) The CEA group (s) with the inoperable position indi-cator is fully inserted and subsequently maintained fully inserted, while maintaining the withdrawl sequence and THERMAL POWER Level required by Specifi-cation 3.1.3.6 and when this CEA group reaches its fully inserted position, the " Full In" limit of the CEA with the inoperable position indicator is actua-ted and verifies this CEA to be fully inserted.

Subsequent operation shall be within the limits of Specification 3.1.3.6.

4. If the failure of the position indicator channel (s) is during STARTUP, the CEA group (s) with the inoperable position indicator channel must be moved to the " Full Out" position and verified to be fully withdrawn via a

" Full Out" indicator within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The Provisions of -

Specification 3.0.4 are not applicable.

c. With a maximum of one reed switch position indicator channel per group or one pulse counting position indicator channel per group inoperable and the CEA(s) with the inoperable position indicator channel at either its fully inserted position or fully withdrawn position, operation may continue provided:

1. The position of this CEA is verified immediately and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter by its " Full In" or

" Full Out" limit (as applicable),

2. The fully inserted CEA group (s) containing the inoperable position indicator channel is subsequently maintained fully inserted and 3 Subsequent operation is within the limits of Specification 3.1.3.6.

d. With more than one pulse counting position indicator channels inoperable, operation in MODES 1 and 2 may continue for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> provided all of the reed switch position indicator channels are OPERABLE.

SURVEILLANCE REQUIREMENTS i

4.1.3.3 Each position indicator channel shall be determined to be OPERABLE by verifying the pulse counting position indicator channels and the reed switch position indicator channels agree within 6 steps at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> except during time intervals when the Deviation circuit is inoperable, then compare the pulse counting position indicator and reed switch position indicator channels at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

3/4 1-25

3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) ...

SAFETY INJECTION TANKS LIMITING CONDITION FOR OPERATION 3.5.1 Each reactor coolant system safety injection tank shall be OPEPAELE with:

a. The isolation valve open and the power to the valve operator removed,

b. Between 1080 and 1190 cubic feet of boreted water,

c. A minimum boron concentration of 1720 PPM, and

d. A nitrogen cover-pressure of between 200 and 250 psig.

AFFLICABILITY: MODES 1, 2 and 3.*

ACTION:

a. With one safety injection tank inoperable, except as a result of a closed isolation valve, restore the inoperable tank to OPERABLE status within one hour or be in HOT SHUTDOWN within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

b. With one safety injection tank inoperable due to the isolation valve being closed, either immediately open the isolation valve or be in HOT STANDBY within one hour and be in HOT SHUTDOW'.

within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.5.1 Each safety injection tank shall be demonstrated OPERABLE:

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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:

1. Verifying the water level and nitrogen cover-pressu e in the tanks, and

2. Verifying that each safety injection tank isolation valve is open.

• With pressurizer pressure 1,1750 psia; MILLSTONE - UNIl 2 3/4 5-1

t 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.1 Primary CONTAINMENT INTEGRITY shall be maintained.

APPLICABILITY: MODES 1,2,3 and 4.

ACTION:

Without primary CONTAINMENT INTEGRITY

• restore CONTAINMENT INTEGRITY within one hour 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 />.

SURVEILLANCE REQUIREMENTS 4.6.1.1 Primary CONTAINMENT INTEGRITY shall be demonstrated:

a. At least once per 31 days by verifying that all penetrations not capable of being closed by OPERABLE containment automatic isolation valves and required to be closed during accident conditions are closed by valves, blind flanges or deactivated automatic valves secured in their positions, except as pro-vided in Table 3.6-2 of Specification 3.6.3.1.

b. At least once per 31 days, by verifying the equipment hatch is closed and sealed,

c. By verifying the containment air lock is OPERABLE per Specification 3.6.1.3

d. After each closing of a penetration subject to type B testing (except the containment air lock), if opened following a Type A or B test, by leak rate testing the seal with gas at P psig) and verifying that when the measured leakage rate for(54 these seals is added to the leakage rate determined pursuant to Specification 4.6.1.2.d for all other Type B and C pene-trations, the combined leakage rate is less than or equal to 0.60 La .

• Operation within the time allowances of the ACTION statements of Specification 3.6.1.3 does not constitute a loss of CONTAINMENT INTEGRITY.

3/4 6-1

CONTAINMENT SYSTEMS CONTAINMENT AIR LOCKS LIMITING CONDITION FOR OPERATION 3.6.1.3 Each containment air lock shall be OPERABLE with:

a. Both doors closed except when the air lock is being used for normal transit entry and exit through the containment, then at least one air lock door shall be closed and

b. An overall air lock leakage rate of less than or equal to 0.05 L, at P, (54 psig).

APPLICABILITY: MODES 1,?,3 and 4.

ACTION:

a. With one containment air lock door inoperable:

1. Maintain at least the OPERABLE air lock door closed and either restore the inoperable air lock door to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or lock the OPERABLE air lock door closed.

2. Operation may then continue until performance of the next required overall air lock leakage test provided that the OPERABLE air lock door is verified to be locked closed at least once per 31 days.

3 Otherwise, 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 />.

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4. The provisiorie of Specification 3.0.4 are not applicable if the outer air lock door is inoperable.

b. With the containment air lock inoperable, except as the result of an inoperable air lock door, maintain at least one air lock door closed; restore the inoperable air lock to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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 />.

3/4 6-6

CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS 4.6.1.3 Each contain:nent air lock shall be demonstrated OPERABLE:

a.* After each opening, except when the air lock is being used for multiple entries, then at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, by verifying no detectable seal leakage by pressure decay when the volume between the door seals is pressurized to greater than or equal to 25 psig for at least 15 minutes,

b. At least once per 6 months by conducting an overall air lock leakage test at Pa (54 psig) and by verifying that the overall air lock leakage rate is within its limit and

c. At least once per 6 months by verifying that only one door in each air lock can be opened at a time.

• Exemption to Appendix "J" of 10CFR50.

3/4 6-7

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Docket No. 50-336 Attachment 2 Millstone Nuclear Power Station, Unit No. 2 Large Break Loss-of-Coolant Accident Analysis Results October,1982

. The Loss of Coolant Accidsnt (LOCA) has been reanalyzed for Millstone Unit 2 with 15.3% (1300 tubes) steam generator tube plugging and 5.5% primary flow reduction. The following information. amends the Safety Analysis Report section on Major Reactor Coolant System Pipe Ruptures. The results.are consistent .

with. acceptance. criteria provided in Reference [1].

The description of the various aspects of the Westinghouse LOCA analysis methodology is given in Reference [2]. This document describes the major phenomena modeled, the interfaces among the computer cooes, ano the features of the codes which ensure compliance with the Acceptance Criteria. The SATAN-VI, WREFLOOD, C0CO, and LOCTA-IV codes wnich are used in the LOCA analysis are described in detail in References [3]

through [6]; code moditications' are specified in References [7] througn

[13]. These codes are used to . assess the core heat transfer geometry and to determine if the core remains amenable to cooling throughout ano subsequent to the blowdown, refill, and refloca phases of the LOCA. The SATAN-VI computer code analyzes the thermal-hydraulic" transient in the I RCS tring bIow'down, and thAWREFLOOD computer code is used to calculate this transient & ring the refill and reflood phases of the accident.

The COCO canputer code is used to calculate the Containment pressure transient throughout the LOCA analysis. Similarly, the LOCTA-IV compu-ter code is used to compute the thermal transient of the hottest fuel rod & ring the entire analysis.

I SATAN-VI is used to calculate the RCS pressure, enthalpy, density, and l

the mass and energy flow rates in the RCS, as well as steam generator energy transfer between the primary and secondary systems as a function l of time & ring the blowdown phase of the LOCA. SATAN-VI also calculates the accumulator water flow rates and internal pressure and the pipe l

break mass and energy flow rates that are assumed to be vented to the l Containment W ring blowdown. At the end of the blowoown phase, these l

1 1 .

The mass anc 'en rgy ' release data are transferred to the WREFLOOD coce.

rites curing blowoown are utiliz c in the COCO coce for use in the oetermination of the Containment pressure response curing enis first pnase of the LOCA. Accitional SATAN-VI oucout cata incluaing tne . core

'fow rates ano enthalpy, the core pressure, an,o the core power ' decay

transient, are transferreo to the LOCTA-IV coce.

With initial information from the SATAN-VI code, WREFLOOD uses a system thermal-hydraulic model to determine the core flooding rate (i.e., the rate at which coolant enters the bottom of the core), the coolant pres-sure and temperature, and the core water level ering the refill anc reflood phases of the LOCA. WREFLOO'O also calculates the mass ano energy flow addition to the Containment through the breax. Since the mass flow rate to the Containment depenas upon the core flooding rate and the local core pressure, which is a function of the Containment backpressure, the WREFLOOD and COCO codes are interactively linked.

WREFLOOD is also linked to the LOCTA-IV code in that thermal-nydraulic parameters from WREFLOOD are used by LOCTAIV in its calculation of the fuel temperature. LOCTA-IV is used throughout the analysis of the LOCA transien't to calculate the fuel clad temperature and metal-water reac-tion of the hattest rod in the core.

The analysis presented.here was performed with the 1981 version' of the evaluation model which includes the NUREG-0630.D3] Reactor Coolant pumps are assumed to i

l continue to run during blowdown unless othemise noted.

m.* .

O 2

Reaults The analysis of the loss of coolant ac:icent is performed at 102 percent of :ne ;icansac core power rating. The peak linear power and total c:re power useo in the analysis are given in Tahle 2. Since there is margin.

between the value of peak linear power density used in this analysis and the value of the peak linear power density expected during plant opera-tion, the peak clad tamperature calculated in this analysis is greater than the maximum clad temperature expected to exist.

Table 1 presents the occurrence time for various events throughout the accident transient.

Table 2 presents selected input values and results from the hot fuel rod thermal transient calculation. For these results, the hot spot is defined as the location of maximum peak clad tamperatures. That loca-tion is specified in Table 2 for the worst break case analyzed. The location is

(

indicated in feet which presents elevation above the bottom of the active fuel stack.

l Table 3 presents a sumscry of the various containment systems parameters and structural parameters which were used as input to the COCO ccmputer ,

f codeU3 used in this analysis.

Figures 1 torougn 14 present the parameters of principal interest from the larga break ECCS analysis. The following items are noted:

Figure 1: Hot spot clad temperature. .

Figure 2: Coolant pressure in the reactor core.

Figure 3: Water level in the core and downcomer during reflood.

Figure 4: Containment pressure transient .

Figure 5: Core flow during blowdown Figure 6: Fuel rod heat transfer coefficients.

Figure 7: Hot spot fluid temperature.

Mass released to Containment during blowdown.

l Figure 8:

l Figure 9: Energy released to containment during blowdown.

Figure 10: Fluid quality in the hot assembly Figure 11: Mass velocity Safety injection tank water flow rate into RCS during Figure 12:

blowdown (per tank).

Figure 13: Pumped safety injection water flow rate during reflood. .

Figure 14: Core reflooding rate.

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Conciusions " normal Anaivcas

.cr

:reaks up to and incluoing the doubie ended severanca of a reac cr c:ciant pice, the Eme' gency Core Ccoling Systen will meet :ne Ac:aotanca Critaria as presentad in 10C?R50.46.El3 That'is:

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1. The calculated peak clad tamperature does not exceed 2200 F based on a peak core linear power of 15.6 kw/ft.

2. The amount of fuel element cladding that reacts chemically with water or steam does not exceed 1 percant of the total amount of Zircalloy in the reactor.

3. The clad temperature transient is terminated at a time when the core geometry is still amenable to cooling. The cladding oxidation s

limits of 17*. are not exceeded during or after quenching.

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4. The core tamparature is reduced and decay heat is removed for an extended period of. time, as required by the long-lived radioactivity remaining in the core.

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?.EFERE? ICES

'Accectance :riteria for -Emergency C;ri ._ ~.ing Systems for Lignt aater lao;eo :iuclear ?Ower .ieactors,' . ;FR50.J6 ano Appencix K of l974

' :CF:.50. :eaeral Registar, loiume 19, :iumoer 3, January 4, Borcelon, :. M., Massie, H. W. ano Zoroan T. A., "Westingnouse dCCS 2.

Evaluation Model - Sumary," 'WCAP-8339, July 1974.

3. Bordelon, F. M., et al ., " SATAN-VI Program: Comprehensive Space-Time Dependent Analysis of Loss if Coolant," WCAP-8302 (Proprietary) ana WCAP-8306 (Non-Proprietary), June 1974.

4. Kelly, R. O., et al., " Calculational Model for Core Refloooing after a Loss of Coolant Accident (WREFL000 Code)," WCAPU170 (Proprietary) and WCAP-8171 (Non-Proprietary), June 1974.

Bordelon, F. M. and Murphy, E. T., " Containment Pressure Analysis 5.

Code (C0CO)," WCAP-8327 (Proprietary) and WCAP-8326 (Non- '

Proprietary)', June 1974.

6. .Bordelon, F. M. , et al . , "LOCTA-IV Program: Loss of Coolant Tran-sient Analysis," WCAP-8301 (Proprietary) and WCAP-8305 (Non-Proprietary), June 1974.

}

7. Ferguson, K. L.., and Kemper, R. M., ECCS Evaluation Mocel for Westinghouse Fuel Reloads of Combustion Engineering NSSS, WCAP-9528 (Proprietary) and WCAP-9529 (Non-Proprietary), June 1979.

8. Ferguson, K. L., and Kemper, R. M., Addendun to ECCS Evaluation Model for Westinghouse Fuel Reloads of Comoustion Engineering NSSS, October 1979.

9. Bordelon, F. M., et al., " Westinghouse ECCS Evaluation Mocel - Sup-plementary Information," '* CAP-8471 (Proprietary) anc WCAP-8472 (Non-Proprietary), April 1975.

6 _ _ _ _

.. o

10. '"Westingnouse ECCS Evaluation Macel - Octooer 1975 /ersion,*

WCAP-4622 (Proarietary), anc wCAP-6623 (Non Proprietary),

'iovemoer 197E.

11. '.etter .'IS-CE-924, dateo January 23, 1976, C. Eicneloinger (Westing-house) to 0. B. Vassallo (NRC).

12. 'Eicheidinger, C., " Westinghouse ECC5 Evtluation Model,, Feoruary 1978 Version," WCAP-9220 9-A (Proprietary Version), WCAP-9221-A-(Non-Proprietary Version), February 1978.

13. " Westinghouse ECCS Evaluation Model - 1981 Versio.n,"' WCAP-9220-P-A Revision I (Proprietary' Version), WCAP-9221-A~ Revision I (.Non-Proprietary Version), February 1982.

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TABLE 1 .

-LARGE BREAK TIME SEQUENCE OF EVENTS t

C -0.6 DECLG D '

(Sec)

START 0.0 S. l. Signal

• 0.68 -

S. I. Tank injection 15.4 i'

End of Blowdown 21.3 i .

Bottom of Core Recovery 34.0 S.1. Tank Empty 63.8 End of Bypass 21.3 .

i

• from containment pressure sensor i r ~ -

TABLE 2 .

LARGE BREAK ,

- 00-0.6 DECLG Pe,ak Clad Temp. *F 2055 Peak Clad Location,Ft. 7.0 Local Zrfil 02 Rxn(max)% 4.5 Local Zrlil 20 Location,Ft. .

7.0

<0.3 Total Zr/H 2O Rxn,1 llot Rod Burst Time,sec 28.6-llot Rod Burst Location,Ft. 5.7 Calculation Assumptions -

NSSS Power,Hwt,102% of 2700 Peak Core Linear Power, kw/ft 15.6 215 ,

S.I. Tank Actuation Pressuge, psia S.I. Tank Water Volume, f t 'per tank 1080 O

e e

e

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1. TABLE 3 Millstone Unit 2 Containment Physical Parameters 1.938 4 100 -t3 Nec ree '/olume

cntainment Initial Concitions: 99 5 Humioity Containment Temperature 60*F Enclosure Builoing Temperature 60*F 40,F Cround Tenperature 14.7 psia Initial Pressure Initial Time for: 26 seconos Spray Flow 0.0 seconos Fans (3) 14.0 seconos Additional Fan Containment Spray Water:

50*F Temperature 3300 gpm Flow Rate (Total, 2 pumps)

Fan Cooling Capacity (Per Fan)

Vapor Temperature (*F) Capacity (BTU /Sec) 60 0.0 145 3360.0 -

165 5280.0 .. ,

300 28800.0 350 32400.0 Containment Heat Absorbing Surfaces l

1. Surface Areas and Thicknesses

a. Shell and dome - 71,870 Ft2 (1) Paint - 0.003 In. (one side exposed to containment atmosphere)

(2) Caroon steel - 0.25 In.

(3) Concrete - 3.0 Ft. (one side exposeo to enclosure ouiloing atmosphere) .

b. Unlined Concrete - 62,800 Ft2 (1) Concrete - 2.0 Ft. (one side exposed to containment acnosphere, one side insulated)

c. Galvanized Steel - 120,000 Ft2 (1) Zinc - 0.0036 In. (one side exposeo to containment atmosphere)

(2) Caroon steel - 0.20 In. (one sice insulateo) 10

7ABL2 3 (Cant's.; .

Millston's Unit 2

• " - 2ntainment Physical Parameters

. 3 1intea hin Steel - 56,350 Ft2

'.li 3 11nt - 0.003 In. (one siae exposea to containment atmosonere)

(2) Caroon steel - 0.2 In. (one sice insulatec)

e. Painted Steel - 32,600 Ft2 (1) Paint - 0.003 In. (one side exposea to containment atmopshere)

(2) Caroon steel - 0.26 In. (one side 'insulatea)

f. Painted Steel - 22,425 Ft2 (1) Paint - 0.003 In. (one side exposea to containment atmosphere) ,

(2) Carbon steel - 0.86 In. (one siae 'insulatea) 9 Painted Thick Steel 4,230 Ft2 (1) Paint - 0.003 In. (one side exposea to containment atmosphere)

(2) Carbon steel - 2.94 In. (one side insulatea)

h. Containment Penetration Area - 3,000 Ft2 (1) Paint - 0.003 In. (o'ne s'ide dxpo' sed to containment atmosphere)

(2) Carbons steel - 0.75 In. f (3) Concrete - 3.75 Ft. (one siae exposea to enclos4.re .ouilaing atmosphere) .

i i 1. Stainless Steel Line Concrete - 8,340 Ft2 (1) Stainless steel - 0.25 In. (one sice exposed to containment atmosphere)

(2) Concrete - 2.0 Ft. (one side insulatea)

j. Base 51ab - 11,130 Ft2 (1) Concrete - 8.0 Ft. (one side exposea to con'tainment sump, one side exposed to ground)

k. Neutron Shield - 1400 Ft2 (1) Stainless steel - 0.024 Ft. (both siaes exposea to containment atmosphere) 2

1. CEDM Cable Support Structure - 1380 ft (1) Paint - 0.006 In.

(2) Stainless Steel.- 0.1094 ft. (both sides exoosed to containment atmosphere) t 11

TABLE 3 (Cant'c.; ,

Millstone Unit 2 Containment Physical Parameters

2. Thermai Procerties Conouctivity Heat Capacity ,

Material (BTU /hr-ft- F) (BTU /ft3 *F) 36

a. Concrete 2.0 35.0 55

b. Caroon Steel 62

c. Stainless Steel 10.0 '

1.5 32

d. Paint 45

e. Zinc 70.0 t

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6 Docket No. 50-336 Attachment 3 Millstone Nuclear Power Station, Unit No. 2 Small Break Loss-of-Coolant Accident Reevaluation l

October,1982 l

b.

4 Small Break Loss-of-Coolant Accident Analysis Reevaluation Northeast Nuclear Energy Company (NNECO) provided the NRC Staff with the results of the small break J.OCA analysis conducted for Cycle 3 operation in Reference (1). The results of this analysis have been demonstrated to be applicable and appropriate for cycles 4 and 5 pending the completion of small break LOCA model revisions by our fuel vendor. Small break LOCA model revisions are due to be submitted, in fulfillment of Item II.K.3.30 of NUREG-0737, in March,1983 (Reference (2)). A During the Cycle 4/5 refueling outage, unanticipated steam generator tube corrosion in the form of pitting necessitated the plugging of approximately seven hundred (700) tubes. NNECO evaluated the resulting elfects of reduced primary coolant flow and heat transfer area on the docketed small break LOCA analysis results and concluded in Reference (3) that continued plant operation was bounded by the assumptions utilized in the Reference (1) analysis.

I Anticipating the potential for additional steam generator tube plugging, NNECO has performed evaluations to dernonstrate the acceptability of the current small 1:reak LOCA analysis and results for up to an additional 9.2% reduction in heat transfer area (approximately 3150 plugged tubes, total) and a reactor coolant flow rate of 350,000 gpm, for Cycle 6 operation. This evaluation addresses the impact of these changes on the small break ECCS performance for the limiting small break, the 0.1 ft 2cold leg break. The reduction in heat transfer area is addressed first followed by a discussion of the effect of the reduction in primary coolant flow rate.

Reduced Steam Generator Heat Transfer Area In performing the evaluation, a total steam generator heat transfer area of 145,000 square feet, which corresponds to an additional reduction of 9.2% in heat transfer area, was chosen for conservatism. Using this conservatively reduced heat transfer area, the analysis determined the maximum increase in peak clad surface temperatures to be 13.20F for the limiting 0.1 square foot cold leg break at the pump discharge. This insensitivity is not surprising since for that portion of the transient during which steam generator heat transfer is important, the core power is characterized by fission product decay heat generation wherein the power levels are less than 5% since plant trip will have occurred.

Reference 4 presents an analysis applicable to the CE 2700 Mwt class of plants in which complete loss of one steam generator was assumed. The results of the Refarence 4 analysis (Case 17) demonstrated that the effect of a loss of one steam generator is insignificant for small breaks requiring steam generator heat transfer and includes break sizes 0.02 square feet and smaller. The 0.02 square foot break is well below the 0.1 square foot limiting small break for Millstone ,

Unit 2.

Since the peak clad temperature for the 0.1 square foot break is 19710F, in Reference 1, the expected peak clad temperature response due to the reduced steam generator area will be less than 1984.20F, which is still well below the 10 CFR50.46 limit of 22000F for acceptability.

4 o

Reduced Primary Flow As noted above, the reduced primary system flow of 350,000 gpm represents a 5.4% reduction relative to the 370,000 gpm flow utilized in the licensing analysis of Reference 1. The effect of the reduced flow will increase the initial fuel stored energy by a small amount. However, this small increase in initial fuel stored energy will have no effect on peak clad temperatures for the limiting 0.1 square foot break since all of the initial fuel stored energy is removed from the fuel rod prior to uncovering the core when the rod heatup begins. That is, uncovering of the core is not initiated until after 500 seconds following the opening of the break. Since reactor trip occurs within the first 15 seconds, the ensuing flow coastdown over the next 500 seconds is more than sufficient to remove the initial fuel stored energy. As illustrated in Reference (1) for the 0.1 square foot break, the clad temperature prior to uncovering of the core is less than 100F above the coolant temperature. At this time, the fuel pellet temperatures and hence clad temperatures are only a function of the decay heat generation rate which remains unchanged in the evaluations. Thus, increases in initial fuel stored energy resulting from the 5.4% reduction in flow rate, will not i aficct peak clad surface temperatures since the fuel temperature distribution will always subside to that distribution characterized by decay heat generation due to the large flow coastdown time prior to core uncovery.

SUMMARY

An evaluation has been performed to address the impact of a reduction in steam generator heat transfer area and a reduction in primary system flow rate on the small break LOCA ECCS performed for Millstone Unit 2. The results of the evaluation demonstrated that for the conservatively assumed reduction in area of 9.2%, an insignificant increase in peak clad temperature of only 13.20F resulted. This insensitivity is consistent with an evaluation applied to the 2700 MWT class plants reported in Reference 4 which demonstrated that a 50%

reduction in heat transfer area does not affect the small break LOCA performance.

The 5.4% reduction in flow also does not affect small break LOCA ECCS performance for the 0.1 square foot break due to the lengthy 500-second initial flow coastdown period during which all of the initial fuel stored energy is removed from the fuel rods. Since all of the initial fuel stored energy is removed from the rods prior to uncovering of the core, the small increase in initial fuel stored energy due to the lower flow will have no effect on peak clad temperature for the limiting 0.1 square foot break.

Based on the evaluation contained herein, the effect of reduction in steam generator heat transfer area and primary system flow rate will not impact the Millstone Unit 2 small break LOCA ECCS performance. It can therefore be concluded that operation of Millstone Unit 2 at a power level of 2754 MWT (102% of 2700 MWT) with a reduction in steam generation heat transfer area of 9.2% and a reduction in primary system flow rate of 5.4% is acceptable.

6

/

References:

(1) W. G. Counsil letter to R. Reid, dated March 22,1979.

(2) W. G. Counsil letter to R. A. Clark, dated September 2,1982.

(3) W. G. Counsil R. A. Clark, dated February 4,1982.

(4) CEN-ll4, Review of Small Break Transient in Combustion Engineering Nuclear Supply Systems, July 1979.

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