Text

Proposed Tech Specs Re Max River Water Temp Limit & Containment Pressure Analyses
	ML20245H393
Person / Time
Site:	Beaver Valley
Issue date:	06/22/1989
From:	; DUQUESNE LIGHT CO.
To:	;
Shared Package
ML20245H381	List: ML20245H378; ML20245H387; ML20245H393;
References
	NUDOCS 8906290473
	Download: ML20245H393 (123)
	v • d • e

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ATTACHMENT A Revise the Technical Specifications as follows:

Remove Pace Insert-Pagg.

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'3/4 7-14 3/4 7-14 B3/4 6-2 B3/4 6-2 8906290473 890622 DR ADOCK0500ggg4

REACTIVITY CONTROL SYSTEMS BORATED WATER.. SOURCES - SHUTDONN LIMITING CONDITION FOR OPERATION 3.1.2.7 'As a minimum, one of the following borated water sources shall be OPERABLE:

'a.

A boric acid storage system with:-

1. A minimum contained volume of 5000-gallons,

2. Between 7000 and 7700 ppm of boron, and

3. 'A minimum solution temperature of 65'F.

b. The refueling water storage tank with:

1. A minimum contained volume of 175,000 gallons,

2. A minimum boron concentration of 2000 ppm, and

3. A minimum solution temperature of 45'F.

l APPLICABILITY: MODES 5 and 6.

ACTION:

With no borated water source OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes until at least one borated water source is restored to OPERABLE status.

SURVEILLANCE REQUIRI'MENTS 4.1.2.7 The. above required borated water source shall be demonstrated OPERABLE:

a. At least once per 7 days by:

1. Verifying the boron concentration of the water,

2. Verifying the water level of the tank, and

3. Verifying the boric acid storage tank solution temperature when it is the source of borated water.

b. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWST temperature when it is the source of borated water and the outside ambient air temperature is <45'F.

l BEAVER VALLEY - UNIT 1 3/4 1-15 PROPOSED WORDING

l REACTIVITY CONTROL SYSTEMS l

BORATED WATER SOURfTS - OPERATING 1 l

LIKITING CONDITION FOR OPERATION' 3.1.2.8 As a ~ minimum, the following borated water source (s) shall be OPERABLE as required by Specification 3.1.2.2.

a. A' boric acid storage system with: ;

1. A minimum contained volume of 11,336 gallons,

2. Between 7000 and 7700 ppm of boron, and j

3. A minimum solution temperature of 65*F. I

b. The refueling water storage tank with:

1. A contained volume between 439,050 gallons.and 441,100 gallons of borated water,

2. A boron concentration between 2000 and 2100 ppm, and

3. A solution temperature of 145'F and $55'F.

APPLICABILITY: MODES 1, 2, 3 & 4.

ACTION:

a. With the boric acid storage system inoperable, restore the storage system to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to . at least 1% delta k/k at 200*F within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore the boric acid storage system to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With the refueling water storage tank inoperable, restore the tank to OPERABLE status within one hour or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.1.2.8 Each borated water source shall be demonstrated OPERABLE:

BEAVER VALLEY - UNIT 1 3/4 1-16 PROPOSED WORDING

t 4

REACTIVITY CONTROL SYSTEMS '

SURVEILLANCE REQUIREMENTS (Continued) {

l a.. l At least-once per 7 days by: ;

1. Verifying the boron concentration in each water .

4 source, ,

2. Verifying the water level in each water source, and {

3._ verifying the boric acid storage system solution temperature.

b. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by . verifying the RWST temperature when the RWST ambient air temperature is

<45'F or >55'F.

l BEAVER VALLEY - UNIT 1 3/4 1-17 PROPOSED WORDING l

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l TABLE 3.3-5 (Continued) l ENGINEERED SAFETY FEATURES RESPONSE TIMES 1

l INITIATING SIGNAL AND FUNCTION RESPONSE TIME.IN SECONDS l

4. Steam Line Pressure-Low

a. Safety Injection (ECCS) s.27.0#/37.0##

b. Reactor Trip (from SI) $ 3.0

c. Feedwater Isolation 5 75.0(1)

d. Containment Isolation-Phase "A" 5 22.0(3)/33.0(2)

e. Auxiliary Feedwater Pumps Not Applicable

f. Rx Plant River Water System 5 77.0(3)/110.0(2)

.g. Steam Line Isolation s 8.0 5.- Containment Pressure--Hich-Hiah

a. Containment Quench Spray 1 85.0(2)

b. Containment Isolation-Phase "B" Not Applicable

c. Control Room Ventilation Isolation 5 22.0(3)/77.0(2)

6. Steam Generator Water Level--Hiah-Hiah

a. . Turbine Trip-Reactor Trip 5 2.5

b. Feedwater Isolation i 13.0(1)

7. Containment Pressure--Intermediate Mich-Hiah

a. Steam Line Isolation s 8.0

8. Steamline Pressure Rate--Hich Necative

a. Steamline Isolation s 8.0

9. Loss of Power

a. 4.16kv Emergency Bus Undervoltage 1 1.3 (Loss of Voltage)

I

b. 4.16kV and 480- Emergency Bus s 95 Undervoltage (Degraded Voltage)

BEAVER VALLEY - UNIT 1 3/4 3-27 1 PROPOSED WORDING 2

.___-_________._.w CONTAINMENT SYSTEMS CONTAINMENT LEAKAGE LIMITING CONDITION FOR OPERATION 3

3.6.1.2 Containment leakage rates shall be limited to:

a. An overall integrated leakage rate of:

1. < L,a 0.10 percent by weight of the containment air per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at Pa , (40.0 psig), or l

b. A combined leakage rate of < 0.60 La for all penetrations and valves subject to Type B and C tests as identified in Table 3.6-1, when pressurized to P a*

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With either (a) the measured overall integrated containment leakage rate exceeding 0.75 L, r (b) with the measured a

combined leakage rate for all penetrations and valves subject to Types B and C tests exceeding 0.60 L, restore the leakage rate (s) to within the limit (s) prior to aincreasing the Reactor Coolant System temperature above 200*F.

SURVEILLANCE REQUIREMENTS 4.6.1.2 The containment leakage rates shall be demonstrated at the following test schedule and shall be determined in conformance with the criteria specified in Appendix J of 10 CFR 50* using the methods and provisions of ANSI N45.4-1972:

a. A Type-A test (Overall Integrated Containment Leakage Rate) shall be conducted at 40 1 10-month intervals during shutdown at P a (40.0 psig).

l

Exemption to Appendix J of 10 CFR 50, Section III.D.1(a),

granted on December 5, 1984.

BEAVER VALLEY - UNIT 1 3/4 6-2 PROPOSED WORDING

_ _ _ - - _ . _ a

L .g CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

-b. If .any. periodic. Type A test. fails to~ meet..75 L , the test ' schedule' for subsequent Type A tests ~shall.be reviewed and. approved by the Commission. .If two consecutive Type A tests fail to meet .75.L , a Type A test shall' be performed at least every $8 months until two consecutive Type A. tests , meet .75'La at which time the above test schedule may be resumed,

c. The . accuracy of each Type A test shall be verified by a supplemental test which:

1. Confirms the accuracy of the Type A test by.

verifying that the difference.between supplemental L and Type A test data is within 0.25 L,.

2. Has- a duration sufficient to accuratelyJestablish the change in leakage for between the Type A test and the supplemental test.

3. Requires the quantity of gas injected into the containment- or bled from the containment during the supplemental test to be equivalent to at least~

25 percent of the total. measured leakage rate at Pa (40.0 psig). l-

d. Type B and C tests shall be conducted with gas at Pa * (40.0 psig) at intervals no greater than 24 months j except for tests involving:

1. Air locks,

2. Penetrations using continuous leakage monitoring systems, and

3. Valves pressurized with fluid from a seal system,

e. Air locks shall be tested and demonstrated OPERABLE per Surveillance Requirement 4.6.1.3.

f. Leakage from isolation valves that are sealed with fluid from a seal system may be excluded, subject to the provisions of Appendix J,Section III.C.3, when determining the combined leakage rate provided the seal system and valves are pressurized to at least 1.10 P a

(44.0 psig) and the seal system capacity is adequate to l maintain system pressure for at least 30 days.

Applicable valves may be tested using water as the pressure fluid in accordance with the Inservice Testing Program.

BEAVER VALLEY - UNIT 1 3/4 6-3 PROPOSED WORDING

CONTAINMENT SYSTEMS CONTAINMENT AIR IOCKS 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 leakaga rate of less than or equal to 0.05 La at Pa (40.0 psig).

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

ACTION:

a. With one containment air lock door inoperable:

1. Maintain the associated OPERABLE air lock door closed and either restore the associated i inoperable air lock door to OPERABLE status Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or lock the associated OPERABLE air lock door closed.

2. Operation may then continue until performance of the next required overall air lock leakage test ;

provided that the associated OPERABLE air lock door is verified to be locked closed at least once per 31 dbys.

3. Otherwise, be in at least HOT STANDBY within the l next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

4. The provisions of Specification 3.0.4 are not applicable.

b. With a 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 OPERABLI status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

BEAVER VALLEY - UNIT 1 3/4 6-5 PROPOSED WORDING

l CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS 1

(

l 4.6.1.3 Each containment- air lock shall be demonstrated l OPERABLE: i

a. Within 72 hours following each containment entry, except when the air lock is being used for multiple entries, then at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , by verifying no detectable seal leakage when the gap between the door seals is pressurized for at least 2 minutes to:

1

1. Personnel airlock 1 40.0 psig

2. Emergency air lock 2 10.0 psig or, by quantifying the total air lock leakage to insure the requirements of 3.6.1.3.b are met.

b. By conducting overall air lock leakage tests, at not less than Pa (40.0 psig), and verifying the overall air lock leakage rate is within its limit: l

1. At least once per 6 months, # and

2. Upon completion- of maintenance which has been performed on the air lock that could affect the air lock sealing capability.*

c. At least once per 18 months during shutdown by verifying:

1. Only one door in each air lock can be' opened at a time, and

2. No detectable seal leakage when the volume between !

the emergency air lock shaft seals is pressurized to greater than or equal to 40.0 psig for at.least 2 minutes. l TheprovisionsofS[icification4.0.2arenotapplicable.

Exemption to Appendix J of 10 CFR 50, dated November 19, 1984.

BEAVER VALLEY - UNIT 1 3/4 6-Sa PROPOSED WORDING

CONTAINMENT SYSTEMS INTERNAL PRESSURE LIMITING CONDITION FOR OPERATION 3.6.1.4- Primary Containment internal air partial pressure shall be maintained 2 8.9 PSIA and within the acceptable operation range (below and to the left of the applicable containment temperature limit line) shown on Figure 3.6-1 as a function of river water temperature.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With the containment internal air partial pressure < 8.9 PSIA or above the applicable containment temperature limit line shown on Figure 3.6-1, restore the internal pressure to within the limits l within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.6.1.4 The primary containment internal pressure shall be determined to be within the limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

l BEAVER VALLEY - UNIT 1 3/4 6-6 PROPOSED WORDING i

u__-----_--.____-___--_-_____--_--___ _ . . _ _ __ _ _ - - _ _ - - - _ _ _ - _ . - _ _ _ _ _ - . _

I i

1 I i i l i i NCTES:

(

10.6 1. RWST TEMPERATURE BETWEEN -

45 0F AND 550F.

10.5 2. MAXIMUM ALLOWABLE CONTAINMENT -

TEMPERATURE 1050F.

10.4 , 3. MINIMUM ALLOWABLE CONTAINMENT -

AIR PARTIAL PRESSURE 8.9 PSIA.

10.3 4. MAXIMUM RIVER WATER -

TEMERATURE 900F 10.2 ,

CONT LINMEPlT 10.1

$ TEMP ERATUltE 2750F 7 o

gg 10.0 ,

o E; 9.9

\

w$ UN 4CCEP"ABLE 4 w 9.8 3 OPERATION 3 ACCEPTAB .E 3E s 9.7 ph OPERATIOPI 9.6 3

2 4 9.5 2

9.4 9.3 C 3NTAthMENT TEMPE RATURI:295 0F %~ -,

92 '

M wl 9.1 I

9.0 8.9 i

30 35 40 45 50 55 60 65 70 75 80 85 l 90 RIVEFf WATER TEMPERATURE (OF)

FIGURE 3.6-1 MAXIMUM ALLOWABLE PRIMARY CONTAINMENT AIR PRESSURE VERSUG RIVER WATER TEMPERATURE BEAVER VALLEY - UNIT 1 3/4 6-7

\

PROPOSED WORDING L_.-_m._._. _.__.- - - - - -----.___m. . _ - _ _ _ _ - - - - - -

I:

CONTAINMENT SYSTEMS AIR TEMPERATURE LIMITING CONDITION FOR OPERATION 3.6.1.5 Primary containment average- air temperature shall be maintained:

a. Greater than or equal to 75'F and less than or equal to 105'F, or

b. Greater than or equal to 95'F and less than or equal to 105'F in accordance with the requirements of Figure 3.6-1.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With the containment average air temperature > 105'F or less than the minimum containment temperature prescribed in Figure.

3.6-1 (75'F or 95'F) restore the average air temperature to-within~ the limit within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or be in at least HOT STANDBY within the .next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within .the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIRE)ENTS 4.6.1.5 The primary containment average maximum and minimum air temperatures shall be the arithmetical average of the temperatures at the following locations and shall be determined

.at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The nearest alternate thermocouple may be used for temperature determination up to a maximum of one per location.

Location

a. Reactor Head Storage Area - Elev. 802

b. Pressurizer Cubicle - Elev. 740

c. Annulus - Elev. 777

d. RHR Heat Exchanger - Elev. 730 <

e. Annulus - Elev. 701 l

BEAVER VALLEY - UNIT 1 3/4 6-8 PROPOSED WORDING j

CONTAINMENT SYSTEMS 3/4.6.2 DEPHRSURIZATION AND COOLING SYSTEMS CONTAINMENT OUENCH SPRAY SYSTEM LIMITING CONDITION FOR OPERATION 3.6.2.1 Two separate and independent containment quench spray subsystems shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION: 4 With' one containment quench spray subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or-be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.6.2.1 Each containment quench spray subsystem shall be demonstrated OPERABLE; '

a. At least once per 31 days by:

1. Verifying that each valve (manual, power-operated, or automatic) in the flow path not locked, sealed, or otherwise secured in position, is in its correct position; and

2. Verifying the temperature of the borated water in the refueling water storage tank is within the limite of Specification 3.1.2.8.b.3.

l

b. At least once per 31 days on a STAGGERED TEST BASIS, by verifying, that on recirculation flow, each pump develops a differential pressure of greater than or equal to 142 psid at a flow of 2 1600 gpm when tested pursuant to Specification 4.0.5.

l l

l BEAVER VALLEY - UNIT 1 3/4 6-11 PROPOSED WORDING

CONTAINMENT SYSTEMS L

} CONTAINMENT RECIRCULATION SPRAY SYSTEM LIMITING CONDITION FOR OPERATION 3.6.2.2 Four separate and independent containment recirculation spray subsystems, each composed of a spray pump, associated heat exchanger and flow path shall be OPERABLI.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

-With one containment recirculation spray subsystem inoperable, restore. the inoperable subsystem to OPERABLE status within 7

. days or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore the inoperable spray system to OPERABLE status.within the next 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.6.2.2 Each containment recirculation spray subsystem shall be demonstrated OPERABLE:

a. At least once. per 31 days by verifying that each accessible valve (manual, power-operated, or automatic) in the flow path not locked, sealed or otherwise secured in position, is in its correct position;

b. When tested pursuant to Specification 4.0.5, manually start each recirculation spray pump and verify the' pump shaft rotates;

c. At least once per 18 months by verifying that on a Containment Pressure-High-High signal, the recirculation spray pumps start automatically as follows:

RS-P-1A and RS-P-2B 210 1 5 second delay RS-P-2A and RS-P-1B 225 i 5 second delay

d. At least once per 18 months during shutdown, verify that on recirculation flow, each pump develops the required differential pressure and flow rate as shown below when tested pursuant to Specification 4.0.5:

RS-P-1A and RS-P-1B 2 127 psid at 2 2000 gpm RS-P-2A and RS-P-2B 2 132 psid at 2 2000 gpm BEAVER VALLEY - UNIT 1 3/4 6-13 PROPOSED WORDING

PLANT. SYSTEMS 3/4.7.5 ULTIMATE HEAT SINK - OHIO RIVER LIMITING CONDITION FOR OPERATION 3.7.5.1' The ultimate heat sink shall be OPERABLE with:

a. A minimum water level at or above elevation 654 Mean Sea Level, at the intake structure,_and

b. An average water temperature of $90'F.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:.

With the requirements of 1he above specification not satisfied, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

ST

_ . ' _ ' _ _NEILLANCE REQUIREMENTS 4.7.5.1 The ultimate heat sink shall be determined. OPERABLE at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the. average water temperature and water level to be within their limits.

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BEAVER VALLEY - UNIT 1 3/4 7-14 PROPOSED WORDING

g - - -

y e

CONTAINMENT SYSTEMS

. BASES' 4

3/4.6.1.4 and 3/4.6.1.5 INTERNAL PRESSURE AND AIR TEMPERATURE The. limitations on containment internal pressu're and' average air temperature as a function of river water temperature ensure that 1) .the containment l structure is prevented.from' exceeding l-

- its design negative pressure of. 8.0 psia, 2).the containment.

peak pressure does not' exceed the design pressure of 45-psig. l

- during LOCA conditions, and 3) the. containment pressure is returned to subatmospheric conditions following a LOCA.

The containment internal pressure. and temperature limits 1 shown as :a: function. of river water temperature describe the l operational . envelope- that will 1) limit the containment peak pressure to less than its' design value of 45 psig and 2) ensure

.the containment, internal pressure returns subatmospheric within ['

~ 60 minutes following a LOCA.

The~' limits on the parameters of Figure 3.6-1 are consistent j' with-the assumptions of the accident analyses.

3/4.6.1.6- CONTAINMENT STRUCTURAL INTEGIT This limitation ensures.that the structural' integrity of the containment. vessel will be maintained comparable to the original design standards for the life. of the facility.. Structural integrity .is required to ensure that the vessel will withstand

.the maximum pressure of 40.0 psig.in the event of a LOCA. The l visual and Type A leakage tests are sufficient to demonstrate this capability.

3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS 3/4.6.2.1 and 3/4.6.2.2 CONTAINMENT OUENCH AND RECIRCULATION SPRAY SYSTEMS The OPERABILITY of the containment spray systems ensures that cont inment depressurization and subsequent return to subatmospheric pressure will occur in the event of a LOCA. The pressure reduction and resultant termination of containment leakage are consistent with the assumptions used in the accident analyses.

BEAVER VALLEY - UNIT 1 B 3/4 6-2 i: PROPOSED WORDING l

ATTACHMENT B Safety Analysis Beaver Valley Power Station, Unit No. 1 Proposed Technical Specification Change No. 157 Description of Amendment Request: The proposed amendment would revise the maximum river water temperature limit and revise several other Technical Specifications based on the revised containment pressure analyses performed to support increasing the river water temperature limit. More specifically, these changes would include:

1. The minimum RWST temperature in Specifications 3.1.2.7 and 3.1.2.8 would be increased from 43*F to 45'F and a maximum allowable RWST temperature of 55'F is provided in Specification 3.1.2.8. Surveillance requirement 4.6.2.1.a.2 is revised to refer to the RWST temperature limits given in Specification 3.1.2.8.

2. The Quench Spray System response time listed in Technical Specification Table 3.3-5 would be increased from 77 sec. to 85 sec.

3. The Peak Accident Pressure (Pa) in Specifications 3.6.1.2, 4.6.1.2, 3.6.1.3 and 4.6.1.3 would be increased from 38.3 to 40.0 psig.

4. The containment air partial pressure specification 3.6.1.4, associated Figure 3.6-1 and the containment air temperature specification 3.6.1.5 would be revised to be consistent with the assumptions of the revised containment pressure analyses.

5. Pump surveillance requirements for the quench spray and recirculation spray pumps listed in Specifications 4.6.2.1.b and 4.6.2.2.d, respectively, would be revised to reflect allowable margins for pump degradation assumed in the revised analyses.

6. The ultimate heat sink temperature in specification 3.7.5.1 would be revised from 5 86*F to 5 90*F.

7. The RWST level -

auto QS flow reduction trip setpoint in Technical Specification Table 3.3-4 is revised from 11'0" i 3" to 8'6". The allowable value for this setpoint is revised from 11' 0" 1 6" to 2 8' 0" and 5 9' 0".

Discussion: The ultimate heat sink provides a source of cooling water for normal operation and to dissipate the heat of an accident and to achieve and maintain the unit in a safe shutdown condition. The design river water temperature for Unit 1 is 86*F. The impact of increasing the river water temperature limit ,

from 86*F to 90*F was evaluated for its effect on safety related equipment for which the river water system supplies cooling water to and its impact on the plant containment depressurization analysis.

Propo2nd Technical Sp cific tion Changn No. 157 Page 2 l

l Containment Depressurization Section 6.4 of the Updated Final Safety Analysis Report for Beaver Valley Unit 1, describes the containment depressurization system i and lists the design basis. The system is designed to cool and '

depressurize the containment to subatmospheric pressure in less than 60 minutes following the Design Basis Accident and maintain the pressure subatmospheric for an extended period. This is accomplished assuming the operation of at least minimum engineered safety features. The system is composed of two (2) subsystems, the Quench Spray System and the Recirculation Spray System. The q recirculation spray heat exchangers provide cooling of the l' containment spray water and long term heat removal during the recirculation mode following a LOCA. The river water system provides the cooling medium for these heat exchangers. An increase in the river water temperature would affect the performance- of the depressurization system by reducing the heat removal capacity of the recirculation spray heat exchangers.

An analysis was performed, using the LOCTIC computer code, for river water temperatures up to 90*F. This analysis included limiting case' analyses for containment depressurization time, subatmospheric peak pressure, peak containment pressure and NPSH analyses for the Low Head Safety Injection Pumps and the Recirculation Spray Pumps. In addition to increasing the assumed river water temperature, the following changes were made in the analyses in order to utilize more current methodology and to address observations from a recent Safety Systems Functional Evaluation (SSFE) review of our Quench Spray System:

1. All analyses were based on a single maximum RWST temperature.

The current containment maximum allowable air partial pressure l curve (Technical Specification Figure 3.6-1) has limits i corresponding to two RWST maximum temperatures. This approach l 1s no longer required and elimination of the second line alleviates the potential for misinterpretation of the limits. l

2. This analysis incorporates the previously analyzed changes in RWST level setpoints for transfer to safety injection recirculation mode and Quench Spray cutback. These changes were made to incorporate the Westinghouse WCAP 11419 methodology for calculating setpe.nt uncertainty levels.

These level setpoint changes were submitted to the NRC in March, 1989 as part of proposed Technical Specification Change Request No. 156 and is currently under review. It should be i noted that the RWST level-auto QS flow reduction setpoint l proposed in this Technical Specification change supercedes the l

1 setpoint provided in the previously submitted change request.

n. _.

i l

Propon;d Technical Sp cification Chang 2 No. 157 Page 3

3. A change was made to the containment net free volume input for the LOCTIC Program. It was determined that the nominal containment net free volume had been used in the current analyses. New values, which are based on the nominal value with uncertainty levels added, were used to provide a more conservative approach. Minimum volumes were used for peak pressure and depressurization calculations, and maximum volumes were used for NPSH calculations.

4. New system performance calculations were performed for the Quench Spray and Recirculation Spray Systems assuming degraded pump performance. These calculations were done to develop flow inputs for the LOCTIC containment analyses. These changes were made to establish degradation allowances for the pumps for consistency with the ASME Section XI Testing program and address SSFE concerns pertaining to Quench Spray pump margin. A degradation of 10% of rated head at the design point was used.

5. Spray thermal efficiency inputs for the LOCTIC Program were modified. The current licensing calculations assume a thermal i efficiency of 95% for the Quench Spray and 90% for the !

Recirculation Spray. In the past, the NRC limited the assumed thermal efficiency to 95% unless otherwise justified. In addition, the original system designs included Model 1HH30100 nozzles manufactured by Spraying Systems company. The nozzles produced a large average droplet diameter, which tended to degrade the spray systems thermal efficiencies. Replacement of these nozzles with SPRACO Model 1713A nozzles in the late 70's, resulted in increased system thermal efficiencies since the Model 1713A nozzle sprays a relatively small average droplet diameter. Thermal efficiency of the Quench Spray and Recirculation Spray Systems have been recalculated. The Quench Spray thermal efficiency has been revised to 99%. The Recirculation Spray thermal efficiency has been revised to vary with time according to Attachment B-1.

6. The ANS 5.1-1979 model for decay heat has been employed for this analysis. The original depressurization analysis utilized decay heat values based on ANS 5.1-1971. Current analyses of subatmospheric plants utilize decay heat values based on ANS 5.1-1979. This later version of ANS 5.1 does not include the large margins that are included in the earlier version of the standard. A statistical uncertainty is used

( instead.

7. The input to the LOCTIC Program for reactor thermal power has been reduced from the ESF rating (2766 MWth) to the licensed rating plus 2% uncertainty (2713 MWth). This change provides necessary margins for relaxation of other limits.

1

Propozad Tschnical' Specification Chengo:No. 157

~

'Page 4

8. 1An internal; option.within the LOCTIC Program'was'used to.take

~ ~

credit'.for steam condensation by safety injection. During the period ;in..which Unit :1; was -licensed, credit for . steam condensation. by safety ' injection during reflooding of the reactor core was not permitted by the NRC. Westinghouse has

'now received- credit for this effort in their latestL ,

containment mass and energy-release model.. For this effort, I this effect has also been included in the LOCTIC evaluation of mass. and energy release. Therefore, . steam. which l passes through the downcomer region. of the reactor vessel is condensed- by mixing with the entering SI water; this reduces the mass of steam released to containment during core reflooding.

i

'The- results. of the . analyses. show that the Containment-L Depressurization : System is capable of depressurizing lthe containment within one (1) hour -and maintaining the. pressure subatmospheric .for.-river. water temperatures up to 88'F.with a minimum, containment temperature of 75'F. Additional analyses was-also performed . assuming. an -initial containment average . air-

. temperature- of 95'F. This increased containment air temperature ~ R reduces the initial. air mass- in the containment and therefore reduces- thef requirements of the containment depressurization-system.. . The analysis results showed that with an initial minimum.

' containment' air ' temperature of 95*T the.depressurization systems were. capableL of:-meeting their design basis -for . river water temperatures up- to 90*F. The containment' peak pressure increases slightly from 38.3 to 40.0 psig and is still well below the design pressure of 45 psig. NPSH calculations for the low head safety injection and. recirculation spray pumps- also show' acceptable results. The minimum available NPSH for the low head safety injection pumps was 12.4 feet which is above the minimum required NPSH of 10.6 feet. The minimum available NPSH for the inside and outside Recirculation Spray pumps was 13.7 feet and 12.8 feet respectively. The minimum NPSH requirement for these pumps is 9.8 feet. A summary of the results and initial conditions for the limiting cases is shown in Attachment B-2.

River Water System Components The effect of increasing the river water temperature limit from 86*F to 90*F on components, other than the recirculation spray heat exchangers, that are required to operate during a post l DBA lineup of the river water system was also reviewed. This

-review included the following components:

Emergency Diesel Generators Cooling System Charging Pump Lube Oil Coolers Control Room Air Conditioning Condenser All components were found to be capable of accepting the increased river water temperature while continuing to perform their intended design function.

i

PropoCd T chnical Sp:cificction Chnnga No. 157 Page 5 i

Plant Cooldown Cacabiljtv i

)

Section 9.4 of the UFSAR lists the functions and design basis of the Primary Plant Component Cooling Water System. For normal operation, the system is designed to supply cooling water to the s components listed at 100*F or less, with river water temperature !

at a maximum of 86*F. Based on a normal heat load of 55 x 10 i BTU /HR with maximum design heat exchanger fouling, an increase in I the river water temperature to 90*F will result in a maximum cooling water supply temperature of 104*F.

The system is also designed to support cooldown of the Reactor Coolant System from 350*F to 140*F in 16 HR, using the RHR System. Based on an increase in the river water temperature of 4*F, the time required to cool down from 350*F to 140*F will increase by approximately four (4) hours.

The Component Cooling System is not used for accident purposes and is not considered part of the engineered safety features. These changes in the design basis of the component Cooling System caused by an increase in the river water maximum temperature will not affect any accident analysis.

The small increase in component cooling water temperature and time required to cool down the plant will not significantly impact the plant's cooldown capability.

Technical Specification Chances The following is a description of the proposed changes to the Technical Specifications and the basis for each change:

1. The minimum RWST temperature in Technical Specifications 3.1.2.7 and 3.1.2.8 is changed from 43*F to 45'F. This change !

is required to make the limit consistent with the assumptions used in the analysis. A minimum RWST temperature of 45'F is !

used for analysic of inadvertent operation of the Quench Spray System. This analysis establishes the minimum allowable containment air partial pressure. In addition, a maximum allowable RWST temperature of 55'F is provided in Specification 3.1.2.8. This limit is consistent with the maximum RWST temperature assumed in the revised containment depressurization analysis. Surveillance requirement 4.6.2.1.a.2 is revised to refer to the allowable RWST temperature range given in Specification 3.1.2.8.

2. The Quench Spray response time listed in Technical Specification Table 3.3-5 is revised to $85 seconds. This change is required to maintain consistency between the overall system response time and the individual component requirements. The 85 second limit is the sum of 10 seconds for diesel generator startup and loading and 75 seconds for stroke time of the Quench Spray containment isolation valves.

These values have been assumed in calculating the Quench Spray delay time for the containment analyses.

1 Propo:: d T chnicnl Sp cific tion Changi No. 157 Page 6 W

3. The Peak Accident Pressure (Pa) as used in Technical ]

Specifications 3.6.1.2, 4.6.1.2, 3.6.1.3, and 4.6.1.3, should i be increased from 38.3 psig to 40.0 psig. This change is the j result of using a more conservative value for containment free 1 volume in the containment analyses as previously discussed.

4. Technical Specification Figure 3.6-1 has been revised to reflect the revised containment depressurization analysis based on a single RWST temperature limit and the increased river water temperature limit. The revised figure includes additional containment average air temperature restrictions when operating with river water temperature above 88'F as required to support the assumptions of the revised analysis. ;

Technical Specification 3.6.1.4 and 3.6.1.5 for containment i maximum air partial pressure and containment minimum temperature are revised to be consistent with the new Figure 3.6-1.

I surveillance requirements for the Quench Spray and

5. Pump Recirculation Spray pumps as listed in Technical Specifications 4.6.2.1.b and 4.6.2.2.d, respectively, are revised to reflect the allowable margins for pump degradation included in the revised containment depressurization analysis. In addition, pump performance requirements are expressed in differential pressure versus discharge pressure.

6. Technical Specification 3.7.5.1 for ultimate heat sink temperature is revised to $90*F. This change allows operation with increased river water temperatures and is consistent with the maximum value assumed in the revised analyses.

7. The RWST level-auto QS flow reduction setpoint in Table 3.3-4 is revised from 11'0" to 8'6". The Allowable value is also reduced by 2 1/2'. This setpoint change was required to include revised instrument channel uncertainties into the analysis. New Reactor Trip and ESF instrument channel inaccuracies were recently calculated for Beaver Valley Unit 1 using an updated methodology described in Westinghouse WCAP 11419. Changes to the Reactor Trip and ESF setpoints and Allowable Values based on this revised methodology was submitted to the NRC in Technical Specification Change Request No. 156. Included in that change request was a revised RWST level-auto QS flow reduction setpoint which was incorrectly calculated. This proposed setpoint change supersedes the setpoint provided in Technical Specification Change Request No. 156. This proposed RWST level setpoint change was used in the revised containment depressurization analysis.

8. The bases for Technical Specification 3.6.1.4, 3.6.1.5 and 3.6.1.6, are revised to maintain consistency with the revised Technical Specifications and analysis results.

PrcpDard'T3chnical Sp;cification Changa No. 157

'Page 7

' Conclusion The containment'. depressurization systems and the components supplied by .the river water system :have been analyzed for and determined- to be capable of continuing to' perform their design safety _ functions' at river water temperatures up to 90*F. This change is-therefore considered safe.

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[. ,1 ATTACHMENT B-1 BVPS-11

'; RECIRCULATION SPRAY SYSTEM THERMAL EFFICIENCY-TIME (SECOND). -EFFICIENCY 0.0- 0.989 400- .0.989 950' O.985' 1700. Oa978 2450 0.969-3200. 0.961-3950: 0.957 4700 'O.958:

5450- -0.956.

6200 0.954 6900 0.952 7700 -0.958 8450 0.958 11200 0.956

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ATTACEMENT C No Significant Hazard Evaluation proposed Technical Specification Chance No. 157 Basis for Proposed No Significant Hazards Consideration Determination: The Commission has provided standards for determining whether a significant hazards consideration exists (10 CFR 50.92(c)). A proposed amendment to an operating license for a facility involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated; (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety.

The proposed change does not involve a significant hazards consideration because:

1) Increasing the river water temperature limit will not increase the probability of an accident previously evaluated. To ensure that the consequences of any accident previously evaluated will not be significantly increased the effected plant safety analysis have been reevaluated. In addition the effect of the increased river water temperature on safety related equipment assumed to operate during an accident which require river water have been reviewed. These evaluations have determined that the design basis requirements of the containment depressurization systems will continue to be met and that the safety related equipment which require river water cooling will be capable of performing their design function at the increased river water limit. The proposed Technical Specification changes will ensure that the assumptions of the containment depressurization safety analysis will remain valid. Therefore, the proposed changes do not involve a significant increase in the consequences of an accident previously evaluated.

2) The proposed changes do not involve any plant equipment or operating configuration changes within the plant. Therefore the probability of an accident or a malfunction of a different type than previously evaluated would not be created.

3) The proposed river water temperature limit is based on a J re-analysis of the containment depressurization safety analysis.

This analysis has determined that with the increased river water temperature limit the containment depressurization systems will be capable of depressurizing the containment within one hour and ;

maintaining the containment at subatmospheric pressure. The i proposed Technical Specification changes will ensure that the assumptions of this re-analysis will remain valid. This change will therefore not involve a significant reduction in a margin of l

safety. l Based on the above considerations, it is proposed to characterize the change as involving no significant hazards consideration.

- _ - - - _ _ l

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ATTACHMENT D FSAR Changes Beaver Valley Power Station, Unit No.'l Proposed-Technical Specification Change No.-157 I

BVPS-1-UPDATED FSAR Rev. 0 (1/8 2 )

177.81. 10 US(B), 256 i s 5.- Safeguards area

6. Main steam valve' area.

The 's-intenance of integrity of the Seismic class-I piping systems =

after differential. settlement has. occurred and while experiencing-leading due to differential -movement. of equipment and structures 1 during seismic _ disturbance is ensured by designing the piping, hangers: and equipment- supports _to accept these loadings while experiencing stress values that are below those allowed by ANSI .)

B31.1. 4 Table 5.2-17 provides the structures, systems and components'within-the containment that are not designated Seismic class I.

Nonseismic Class I structures, systems and components within the '

containment' whose- potential failure could compromise'the functional ;

capability of surrounding class .I components are designed to q withstand . -the combined effects. of norral operating loads and "

earthquake loads. By the static analysis methods outlined in'Section

' B.2.2 structural - elements,. anchorages and restraints are designed-to preclude the possibility of seismic Class .I components becoming-endangered. ' Analysis of nonseismic components is not intended to i' assure their functional capability. .

- 5.2.2 Design Basis and Leading criteria The structural design of the' containment.is based upon:

1. Shielding requirements

2. The pressure and temperature generated by the DBA (see Section 14)

3. The operational and Design' Basis Earthquake (see Section 2.5 and Appendix B)

4. e maximum calculated core thermal power level of 1[Il4 The temperature and pressure resulting from the DBA are selected as the containment design basis, since the normal operating conditions would result in lower design temperatures and ' pressures. The containment structure is also designed for subatmospheric operation and for a maximum leakage of less than 0.1 percent per day of the weight of containment air at the calculated peak containment pressure of psig.

During normal operation, he containment air partial pressure is maintained between 9 and psia and the containment air temperature ranges between 75 F mi um and 105 F maximum, depending upon the temperature of the available river water. The minimum selected operating pressure corresponds to the average 5.2-3 S I- i _ _ _ _ _ _ - _ . . _ _ _ . _ - _ _ . _ , . - _ _ _ __.__m.____m_ _ . _ _ _ . _ _ . - _ _ _

BVPS-1-UPDATED FSAR Rev. 0 (1/82)

.. 17 7 bl .10 US W 256 slightly (less than 3 percent). This reduction is within the capacity reduction factors listed below for variation in material strengths and workmanship. In addition, the DBA pressure will be reduced significantly before the rebar is affected by the temperature ris,e. ,

i The effect of temperature, shrinkage and creep on the concrete i modulus does not affect the analysis of the containment shell and {

dome since the concrete is assumed to be cracked during the DBA

, loading. The containment mat, 10 ft thick, is insulated by l; I

approximately 2 ft of overlying concrete. The DBA produces only i short term loading on the mat and creep is not significant.

Concrete is considered to offer resistance only near the base of the containment. The temperature gradient through the concrete .

shell thickness of 48 inches is shown in Figure 5.2-28. Since j the maximum design pressure only exists for the first 20 seconds i g s shown un rigures 14.4-s2, acA and 568, the temperature effects

, are negligible.

The load capacity of cencrete members that are subject to tension is based on the guaranteed minimum yield strength of the reinforcement steel.

Load capacities of flexural and compression members' are determined in accordance with ACI 318-63. Load capacities so determined are reduced by a reduction factor multiplier, $, to compensate for small adverse variations in material and .

1 workmanship. The reduction factors are provided in Table 5.2-18. '

Stress and strain limits resulting from the loading criteria conform to the requirements of ACI 318-63, Part IV-B. Principle reinforcing steel used in the construction of the containment structure has a minimum yield strength of 50,000 psi and a minimum ultimate strength of 70,000 psi. Concrete has a specified 28 day compressive strength of 3,000 pai. The analyses of the structure for static loading and for dynamic loading are covered in Section 5.2.2.5. Design load values relative to the site are summarized in Section 2.7.

Structural steel sections of the containment interior are designed in accordance with the American Institute of Steel Construction, AISC-63 I381, except that the seismic loading referred to in AISC-63, Part 1, Paragraph 1.56 is considered as the OBE. Allowable stresses are factored as follows:

1. OBE: 33 percent increase

2. DBE: 90 percent of the specified minimum yield strength for structural steel.

The designed ultimate load capacity of the containment structure, ;

as modified by the safety provisions of ACI 318-63, Section 1504, l l

5.2-5 l 1

i

i-i BVPS-1-UPDATED FSAR' ' Rev. O . (1/82)

(

contain static and dynamic loads for some of the components.

These load paths are shown in Figure 5.2-5. Thermal gradients produced in the supports by pipe rupture jet impingement were.

also investigated. .A. horizontal longitudinal split .in. the reactor ' coolant hot leg, for example, produces an average temperature- in ;the flange of the nearest support member of approximately 250 F. .This and other areas have been examined for the effects of jet impingement thermal gradient and forces.

Reactor Vessel Supports, Neutron Shield Tank and Pressurizer Supports For the determination of dynamic loads on the reactor vessel supports, neutron - shield tank and the pressurizer supports, an analysis technique was applied which in similar to that used fon .

the steam generator and reactor coolaat pump supports.- However for the reactor vessel supports and neutron shield tank the STRUDL program was employed. ~

The dynamic model.and.the stress model . used in the analysis of reactor vessel supports are shown in Figures 5.2-10 and 5.2-27,

-respectively. The dynamic model employed for the pressurizer support is shown in Figure 5.2-9. For this system no stress model . was required since the dynamic model was sufficiently detailed to'directly compute stresses and loads.

5.2.2.4. Bases for Containment Analyses To ensure that the containment performs the desired function of protecting the public against gross equipment failures,' analyses have been performed assuming the. double-ended displacement

, ruptures of(a not leg or_ of a cold leb in the reactor coolant

{f 8] - system. These ruptures nave been assumed to c.: cur in conjunction with loss of all off-site power. They are defined in the Design Basis Accident (DBA) description in Section 14.3.

The design of an atmospheric containment is concerned with the WP dhdsry p.f..

peak containment pressure following the double-ended ruoture of al3 ot leg which is higher than the peak containment pressur

-following the double-ended rupture of a cold leg. For peak i

pressure, winter conditions are more severe than summer b knditionsMince, for a subatmospneric containment, le ge ceases when the containment pressure is below atmospheric, two

. additional points in the containment pressure transient are important.

The first is the time required to depressurize the containment, which together with a safety factor is the basis (time) for post DBA site dose calculation (Section 14,3.5). This E-pressurization time is longer following the double-ended rupture of a cold leg than it is followina the double-ende upture of a -hot leo.Mhe second is the maximum -s 4 n t pressure following depressurization. Crhis pressure is highe A 5.2-14

BVPS-1-UPDATED FSAR Rev. 0 (1/82) following the double-ended rupture of a cold leg than it 17 o lowing the double-ended ruoture of a hot _ leg / This value must be such tnat tne containment remains depressurized u n d e r __a l l credible atmospheric pressura- griationsf For cepressurization L time and for pressure following depressurization, summe ditions are more severe than the winter. [ These are the criteria by which the containment depressurization system and the containment are designed.

The BVPS-1 containment volume, heat sink capacity, and engineered safety features are designed to:

1. Yield a maximum peak calculated containment pressure of g .8. sig. In order to meet this condition, the con ainment is designed to withstand, without loss of.

integrity, a maximum pressure of 45 psig and a maximum containment atmosphere temperature of 280 F ,

2. Depressurize in 60 minutes or less, thus terminating all outleakage

3. Remain depressurized following the double-ended displacement rupture of a reactor hot leg, cold leg or reactor coolant pump suction pipe.

hdesign, order to study the effect of these criteria on containment initial containment conditions consistent with the o expected operating pressures are established. The two conditions -k which are used are given in Table 5.2-5 and labeled " Summer" and

" Winter."

Depressurization after the peak pressure and temperature are reached is dependent upon the action of the engineered safety features. For this study (i.e., a DBA) minimum engineered safety features give the longest depressurization time. The peak pressure reached upon a DBA for the hot leg rupture is not a function of whether " normal engineered safety features" or

" minimum engineered safety features" are in operation, since all active engineered safety features become effective after the peak conditions are reached after the DBA.

Table 14.3-4 presents the station input data, with the exception of containment conditions, used to achieve the peak pressures presented in Table 5. 2-5. The containment conditions are given in Table 5.2-5. These data are chosen for illustrative purposes only.

Secfa,n 14.3 The results presented in (Figure 14. 3-26f and analyses of other less serious accidents (Section 14.2) indicate that the bases for design are more than adequate, and that sufficient margin is allowed for uncertainties in design. l l

5.2-15

P "l '

-BVPS-1-UPDATED FSAR Rev. 0 (1/82)

O. .

7 42. psig. : The test. pressure. used during the tests to determine integrity of the containment structure is 15 percent greater than the design pressure of 45 psig.

Temperature During normal operation, the ' temperature . inside the containment

. ranges from- 75 F ' . minimum to 105 F . maximum.. The maximum temperature difference through the containment wall is 85 F.

Wind The containment site is' located in the 80 mph wind velocity : area

'as determined from Figure 1-(b) for 100 year period of recurrence in ASCE Paper No. 3269, (Section 2.7) . Wind design ' pressure, when substituted for the earthquake loading in the Containment-Shell Structural Loading Criteria (Section 5.2.2.1), does not cause maximum stress conditions.

, Tornado A tornado is assumed to produce a reduction in the atmospheric pressure outside the containment plus a wind with transnational and. rotational velocities. The description of the hypothetical-tornado is given in Section 2.7. The wind velocity is converted

[ into an equivalent pressure as recommended in Reference 1.

Missiles The design of the containment structure is checked for . potential external missile penetration during a hypothetical' tornado. The impact loading is concurrent with the tornado ' wind loading criterion stated in Section 5.2.2.1. The hypothetical tornado missile is described in Section 2.7.

Missile impinging upon the containment are discussed in Section 5.2.6.

5.2.2.5.2 Foundation Mat Analysis The circular containment mat is analyzed to determine the effects of loads imposed by the DBA conditions. Analysis is accomplished by means of a digital computer program which has the capability to calculate bending moments, shears and soil pressures for a symmetrically loaded circular plate on an elastic foundation.

The general method used is described in " Practical Methods for Analysis of Beams and Plates on Elastic Foundations," (in Russian) by Boris N. Zhomochkin, which is for a plate on a semi-infinite elastic half space. This method is an adaptation j of the Boussinesq approach. Zhomochkin's method is modified to !

account for a finite depth of elastic foundation , i.e., the distance between mat and underlying rock. In addition, the 5.2-17 1 l

17761. 10 UMW) D b.

BVPS-1-UPDATED FSAR Rsv/0. (1/82)' q

. _ '. .' r' f TABLE 5.2-5' /

.f DETERMINATION OF PEAK PRESSURE DESIG BASES- i

FOR A - HOT LEG RUPTURE LOCA,-~ I

,r .

i.

Initial Conditions Summer Winter

[

Containment Temperature, F 105 80

/

Containment Dew Point, F , 80 70 t .\

Containment Total Pressure, psia,/ 9.5 10.8 '

l Water Vapor Partial Pressure af Dew Point,. psi ,/ 0.5 0.36 Air Partial Pressure, psi !

f 9.0 10.44 i

/

. Peak Conditions /

Containment Temperature', F 269.3 268.4 f

Containment Total P ssure, psia 53.0 54.9 Containment Total ressure, psig 38.7 40.6 Steam Partial Pr ssure at Containment Temperature, p la 41.4 40.8 Air Partial P ssure, psia 11.6 14.1 River Wa'er emperature, F 86 45 Not . Atmospheric pressure at the BVPS-1 site is taken to be 14.3 psia.

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BVPS-1-UPDATED FSAR R ;v. 1 (1/83) ;

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- 17781. 10 UstBJ 2

[ 5.5 DESIGN EVALUATION The reactor containment concept is based upon the use of a dry containment maintained at a subatmospheric pressure of between 9.5 and 11.5 psia during normal operation. This total pressure allows restricted personnel access to the containment. Following a Design Basis Accident, the containment pressu rises above atmospheric to a maximum possible peak of about 38.3 psig, with subsequent outleakage. ga Through the use of the containment depressurization system (Section 6.4), the containment returns to subatmospheric pressure l within 60 minutes after initiation of a LOCA, thus terminating outleakage from the containment. The amount of activity released to the environment as a result of a DBA is much less than would be released from an atmospheric containment, thereby reducing the size of the required exclusion area and the low population zone as defined and determined by 10CFR100. The population center distance requirement is correspondingly reduced below that

. required for an atmospheric containment (Section 2.1).

The containment depressurization system is censidered to be an engineered safety features system. The containment vacuum system (Section 5.4.2) is not considered to be an engineered safety i features system.

Containment isolation features, such as penetrations, access hatches, and isolation valves, meet the requirements of 10CFR50 Appendix A (Section 1.3 and Appendix 1A).

The containment structural design is in accordance with the best current design practices for steel lined reinforced concrete reactor containment structures. The design procedures incorporate accepted analytical methods. Rigid controls were maintained for all materials and construction practices as indicated in Section 5.2 and 10CFR50 Appendix A. The proposed subatmospheric pressure operation results in no significant effect on the structural design.

5.5-1 I

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BVPS-1-UPDATED FSAR Rev. 3 (1/85)

Containment Leakage Rate Tests The containment leakage rate tests are performed in accordance with the guidelines of Appendix J of 10CFR50, " Primary Reactor Containment Leakage Testing for Water Cooled Power Reactors".

]

The containment leakage testing program includes the performance l l

l of Type A tests, to measure the containment overall integrated leakage rate, )

Type B tests, to measure leakage of certain containment components, and Type C tests, to measure containment i isolation valve leakage rate.

j j

The preoperational Type A tests was conducted according to the rules of Section'III.A of Appendix J. ]

Periodic of testsJare(with Appendix conducted in accordance with the exemption noted in Section III .D.1 the Technical Specifications).

These tests are performed using the leakage monitoring system (described in Section 5.4. 2. 2) .

The measured leakage rate does not exceed the design basis accident leakage rate (La) of 0.1 percent per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of the weight of containment air at the' calculated peak containment pressure of Q psig. The remaining leakage characteristics are determined in accordance with Appendix J, Sections III.A.4.a and III.A.S.a. M.0 Type B tests are carried out to monitor the principal sources of leak development in accordance with Appendix J, Section III.B.1 Test Methods. These tests are performed to measure leakage originating at containment penetrations, air lock door seals, equipment and personnel access hatches, and all other components which may develop leaks and require repairs to meet the acceptance criterion of the Type A test.

The preoperational and periodic Type B tests are conducted according to the rules of Appendix J, Section III.B.2 by local pneumatic pressurization of the containment components at a pressure not less than Pa. The acceptance criterion for Type B l tests is given in Appendix J,Section III.B.3.

l The periodic Type B tests are scheduled according to the guidelines of Appendix J,Section III.D.2.

The their Type C tests are performed on the isolation valves to verify sealing capability and leaktightness according to Appendix J,Section III.C.l.

leakage tests. The test includes valve closure and A valve closure test is conducted prior to a valve leakage. test to demonstrate the proper sealing capability of a valve upon receipt of an isolation signal. Those isolation 4 valves which are normally closed are exercised to verify closure and sealing capabilities. Those containment isolation valves which are in a system that is expected to be filled with water for 30 daya following a LOCA and therefore do not represent a 5.6-4

p.

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BVPS-1-UPDATED FSAR Rev. 3 (1/85) be accomplished by aligning the. discharge of the low head safety

. injection pumps. pumps with the suction of the safety injection charging I

The reciundant features of the ECCS recirculation loop include one pump in each of two separable and redundant trains'with crossover

. capability.

suction at the through discharge of each pump. . Each pump takes-separate cross-connected lines from containment sump. the The design of the containment sump and piping configuration from 1the illustrated containment in Figure sump to the low head safety injection pumps is 6.1-1. .

'After the one day, the spray water collected is cold enough to reduce temperature of the combined mass sufficiently. for recirculation without flashing. The basis for this determination is from the plot of sump temperature versus time after the accident which is shown in Figure C6. 3-33 All heat removal is through the recirculation spray subsy(stem. There are no heat exchangers in the ECCS.

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.Those portions of the ECCS located outside of the containment which are designed to circulate, under post accident conditions, radioactively contaminated water collected - in the containment meet the following requirements:

1. Shielding to maintain radiation levels within 'the guidelines set forth in 10CFR100

2. Collection of discharges from pressure relieving devices into closed systems

3. Means to detect and control radioactivity leakage into.

the environs, to the limits consistent with guidelines set forth in 10CFR100.

This criterion is met by minimizing leakage from the system.

Recirculation loop leakage is discussed in Section 6.3.3.

Chanceover from Injection Phase to Recirculation Phase During level set point at which the injection phase, the RWST level decreases to the low time sufficient water is delivered to the containment sump via the containment depressurization system, the ECCS and from the RCS through the break to provide the q required recirculation.

NPSH of the low head safety injection pumps to change to g j

Section 6.3.3.9. The automatic transfer sequence is provided in i

)

6.3-8

BVPS-1-UPDATED FSAR Rr.v . 3 (1/85)

I 17781. 10 ljS(B) 256-During recirculation, a significant margin exists between the design and operating conditions of the low head safety injection system components. In view of these margins, it is considered that the leakage rates tabulated in Table 6.3-9 are conservative.

Leakage detection exterior to the containment is achieved through the use of sump level detection (Section 9.7). The auxiliary building sump pumps start automatically in the event that liquid accumulates in the sump and an alarm in the main control rocm indicates that water has accumulated in the sump. Valving is provided to permit the operator to isolate individually the low head safety injection pumps.

6.3.3.9 Pump NPSH Requirements Low Head Safety Injection Pumps The NPSH of the low head safety injection pumps is evaluated fer g both the injection and recirculation phase orera + d - ^*

fesign basis gr8(limiting accident. I Recirculauon operation gives " a,1 the p wSH requirement and the NPSH available is determined

'by using the pressure flash method. This method assumes that M hg liquid being expelled from the break flashes at the saturation I temperature corresponding to the containment total pressure.. At initiation of recirculation, the available NPSH for the LFSI>

Lcumes was calculated to be 11.7 feet s 2ne minimum NPSH required is 10.6 feet for the LHS1 pumps.

.The LHSI pumps will be automatically throttled during all operating modes (initial injection, cold leg recirculation and hot leg recirculation) using cavitating venturies. The cavitating venturi is designed to reduce the pressure below the fluid vapor pressure and thus choke the flow. Pressure recovers downstream of the venturi near its initial value.

Safety Injection Charcing Pumps The NPSH for the safety injection charging pumps is calculated for both the injection and recirculation phase operation of the design basis accident.

The end of injection phase operation gives the limiting NPSH requirement and the NPSH available is determined from the elevation head and vapor pressure of the water in the refueling water storage tank, and the pressure drop in the suction piping from the tank to the pumps. The Cealculatedj equired NPSH for these pumps is 22 ft and the minimum available NPSH is 30 ft.

6.3-30 u_____ ___

17781. 10 US(B) .256 INSERT A The limiting available NPSH occurs at - the initiation ~ of the. recirculation -

phase following a LOCA. 'A transient calculation of the available NPSH 'is perfcrmed with the LOCTIC computer code described ~ in Section 14.3 for - the '

containment evaluation. However, the available NPSH transient analysis maximizes the energy addition to the containment sump instead of the atmosphere as was done in the Section 14.3 analysis. The major differences are:

1. The' pressure flash model is used for break fluid flashing.

2. A spray thermal efficiency of 100% is assumed.

As shown in Figure 6.3-10, the minimum available NPSH is 12.4 ft and occurs at the initiation of LHSI recirculation mode following a pump suction double-ended rupture.

The limiting single failure is the loss of one emergency diesel. This results in the loss of two recirculation spray heat exchangers and therefore minimizes heat removal from the sump water before the LHSI pump suctions are switched to the sump.

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i i 6.4 CONTAINMENT DEPRESSURIZATION SYSTEM 6.4 l. Design Bases i

The' containment depressurization system is composed of two groups of subsystems: 'the quench spray. subsystems and the recirculation spray subsystems. These systems are designed to provide the

' necessary . cooling and depressurization of the containment after any LOCA.

The subsystems are designed in 'accordance with 1971 General Design' Criteria 38 through 43 of Appendix A to I' 10CFR50.

I operating together, these subsystems cool and depressurize the containment to subatmospheric pressure'in less than 60 minutes following the Design Basis Accident (DBA), assuming the operation of . at least minimum engi,neered safety features as defined in Section 14.3 In addition, the recirculation spray subsystems are capable of maintaining the subatmospheric pressure in the containment for an extended period following the DBA.

Equipment in the containment depressurization system are designed according to the code criteria and earthquake criteria specified in Section 6.2 and 2.5, respectively. The safety and code classes of piping and valves are indicated on_the flow diagrams.

6.4.2 Description The quench and recirculation spray subsystems are sized to satisfy the following design bases:

1. The containment shall be depressurized (returned to below 1 atmosphere pressure) in less than 60 minutes following a LOCA.

2. Once depressurized, it shall remain depressurized after a LOCA.

3. The LOCA to be considered is a double-ended rupture of a reactor coolant n ce- m,in leas.,$wg . l(og

4. b"'T The temperature of the RWST is in the range of 45 F to 55 F.

5. The inlet temperature of the cooling (river) ter for the recirculation cooler is at a maximum of F.

6. The minimum engineered safeguards configu tion (one 360 degree quench spray header and two 180 degree i recirculation headers) shall be used. '

6.4-1 k

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17781.IU u:nu) 256 ) i BVPS-1-UPDATED FSAR Rsv. 0 (1/82) q i

7. Initiation of a recirculation spray subsystem shall be delayed 300 seconds to ensure adequate water in the ,

containment sumps and to provide high containment back i pressures to enhance ECCS effectiveness during reflood.

The function of the sprays is to remove heat from the containment 1 atmosphere by convection and condensation. Heat is transferred d out of the containment bv means of the recirculation coolereA A fvA proguct ot 0.75 x 10" Btu /hr/F and a total duty of 61 x 10 Btu /hr per cooler were found m eatisfv the anei d ,

< constraints listed above J Table 6.4-2 lists the design I parameters ror components of the heat removal system.

All parts in contact with the pumped fluids are made of austenitic stainles 3 steel. The tube side fouling factor is assumedt to be 0.00 . The shell side fouling factor is assumed O '#

to be W , as the shell will be laid up dry. Any shell side !

fouling occurring during service is more than compensated for since the heat duty required of the cooler decreases rapidly with time.

The fluid flow paths through a recirculation cooler is shown in Figure 6.4-4.

The tube side and shell side temperatures and flow rates for the recirculation coolers are determined as follows:

1. Recirculationn $ffdwater flow rates, river water flow rates, quench spra flow rates and refueling water storage tank capacity are assumed

2. An overall heat removal capability, UA (Btu /hr/F) is assumed

3. A parsmeter study is conducted for the minimum safeguards case using varying assumptions for the variables in 1 and 2 above. Initial containment conditions and maximum river water temperatures are selected to maximize the heat load requirements on the recirculation coolers. For each set of assumptions the shell and tube side temperatures and heat removal rates resulting from these assumptions are available as output from a computer analysis.

4. As a, result of the parameter study, a set of recirculation flow rates, quench spray flow rates and the recirculation heat exchanger UA is selected, consistent with emergency diesel loading requirements and depressurization of the containment, as prescribed by the designed bases (Section 6.4$1).

6.4-2 i

_17781. 10- us m m BVPS-1-UPDATED FSAR Rav. 1 (1/83) type of particle could conceivably pass lengthwise through the-screen. and cause clogging of a spray nozzle.

However, since the final' screen opening is smaller than - the smallest spray cozzle

-size, which'is 0.360 inch, such an occurrence is considered to be highly improbable. .For the recirculation spray subsystems, the screen assemblies in the containment sumps are arranged so. ' that no single failure could result in the clogging of all suction points (Section 6. 4. 2) to.the recirculation spray subsystems. A main screen assembly failure, coupled with the plugging or-failure of the suction point caps, must occur for any one of the suction points to be lost. Sufficient area has been provided to ensure that system operation during accident conditions is not t

impaired and entrance flow velocities are< low enough to prevent-entrainment of most small particles. System overdesign allcws for some plugging or loss of function, in spite of the foregoing.

Consideration has been given to the possibility of a reaction between sodium hydroxide and atmospheric carbon dioxide forming a precipitate in the chemical addition tank. If the temperature of the gas space in the tank varies over a range of 60 F each day and all the carbon dioxide which enters due to this breathing reacts, 90 years would be required to react with 1 percent of the stored caustic. The sodium carbonate formed by this reaction could remain soluble at 45-F.

During nom.al unit operation, the recirculation spray coolers are dry on the shell side and Con the tube sid]p r filled wit river water with corrosion preventativesfas described in Section v.9.2.

For .ong term operation, on the order of weeks, there may be some fouling of the tubes on'the river water side, with resultant loss in heat transfer capability. (A fouling f actor of zero is assumeJ--Q -

(becausd) j;,he loss of heat transrer capability will be more than offsetE b7 the decrease in necessary neat load due to decreasing decay heat production. One day after a loss-of-coolant accident, the drop in decay heat is such that one pump and heat exchanger has sufficient heat-removal capacity to hold the containment 4prea urized. With an expected maximum river water temperature 90'of F, the recirculation spray subsystem design is l

conservative, with a minimum 100 percent backup capacity at the onset of an accident. Within one day after the LOCA, the backup capacity exceeds 400 percent.

The recirculation spray coolers have welded construction at all points where there is a potential for leakage of radioactiw recirculation water into the river water. The maximum pressure differential which can occur between the river water and the recirculation water is 150 psi; under these conditions, leakage flow from the recirculation spray subsystems is toward the river water system. The river water is monitored for leakage by means of radiation monitors. The defective subsystems are shut down if leakage above the allowable values (within the limits of 10CFR20) is detected. As a result of the above pressure difference, inleakage of nonborated water into the containment, causing dilution of the borated water in the containment, is not l possible.

6.4-11

17781. 10 (JS(B) 256 BVPS-1-UPDATED FSAR Ray. 1 (1/83)

J Valve operators are located at El. 747 ft in the safeguard area structure. Valve extensions are totally enclosed and water j tight.

The valve pit structure is attached directly to the reactor j i containment mat and lower portion of the shell. Porous concrete {

is placed below the pit and against its three sides. The pit is designed for seismic and hydrostatic loading in addition to dead ",

and live loads. The pit pump casings are completely enclosed by the butyl waterproof membrane surrounding the reactor containment, up to El. 730 ft. where it terminates between the j bottom mat of the safeguard building and the containment shell. I The concrete which surrounds the safeguards area is sealed to prevent entry of ground water into the safeguards area. The seal !

membrane also serves to prevent leakage of any recirculation !

vater from the safeguards area into the earth backfill between the cofferdam and the containment.

rg The inside and outside recirculation spray pumps are cacable of !

bga meeting NPSH requirements under LOCA conditions.[ Minimum NPS Nccurs wnen the containment returned subatmospheric g)$ '

pressure.

is to Analyses show that the water on the containment floor 3 is subcooled with respect to the containment temperature. The water in the sumps provides a net static head of water, after allowance is made for the suction piping losses of about 3.5 ft to 7.0 ft. A minimum of 12.7 ft and 12.2 ft of head is available for the inside and outside recirculation sprav pumps ,

respectively.f The deep well pumps require approximately 9.8 tt NPSH.

The containment depressurization system provides the capability '

to thoroughly mix the containment atmosphere and minimize the possibility of hydrogen accumulating in the containment subcompartments and escaping the mixing action of the spray water. The internal design of the containment structures allows air to circulate freely. All cubicles and compartments within the containment are open at the top and allow air circulation.

Convective mixing in conjunction with containment spray assures a uniform mixture of hydrogen in the containment.

Provisions for sampling the containment atmosphere following a LOCA are provided.

With regard to mixing of post LOCA hydrogen in the subcompartments, four different aspects have been considered:

1. Mixing in the bulk containment above the operating floor elevation

2. Mixing in the subcompartments in which there is no DBA

3. Mixing in the subcompartments in which the DBA occurred 6.4-13

?* 17781. 10 US(BT -256 ]

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A transient , analysis of ' the available. NPSH is performed .with .the. LOCTIC -

.. computer code' described in Section 14.3. Some modifications are made to the.

assumptions ~ in the Section 14 analysis to provide conservatism in the NPSH '

analysis. ' Generally, these assumptions tend to maximize.' sump water temperature and minimize containment pressure. -The major.model-Differences are as follows:

'1; B'reak fluid flashing is modeled by a pressure flash.

2. Spray thermal efficiency is assumed to be 100%.

3. Maximum engineered safety features is assumed operable.

is I2,2 As shown in Figure 6.4-3, the minimum available NPSH h ft for the inside IL/

recirculation spray pumps and Q ft. for the outside recirculation spray.

pumps. The minimum available NPSH occurs following a hot leg. double-ended rupture at approximately 300 seconds after the pumps start.

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BVPS-1-UPDATED FSAR R v. 0 (1/82) designed to reduce the temperature of the reactor coolant to 140 F within 20 hr after a reactor shutdown.

Each primary plant component cooling water. heat exchanger is designed to remove half of the normal cooldown heat removal load occurring 4 hr after a shutdown of the station.

E This system with all components operating as designed will su Tits teria when the river water is at y 100 F water minimum design maximUjp temperatureor meetof 70 F the temperature 1 above with the andcr{Q86 F a a produce water river water at 34 F.

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[ The primary plant component cooling water heat exchangers and pumps, seismic and the component cooling surge tank are designed as Class I components (Appendix B). The valves and interconnecting piping designed as seismic Class I.

between the above components are also seal coolers, The RHR heat exchangers, RHR pump fuel pool heat exchangers, and the component cooling piping connecting these components with the component cooling pumps and heat exchangers are designed as Seismic Class I.

The component cooling water subsystem normally supplies water to some safety-related items (RHR heat exchangers and fuel pool htat exchangers).

However, the component cooling water subsystem is not used for accident engineered safety features.

purposes and is not considered part of t he 9.4.1.2 Chilled Water Subsystem The chilled water subsystem is designed to perform the followf ng functions:

1. Cool the water in the refueling water storage tank to a temperature of 45 F I

l

2. Supply cooling water for containment air recirculation coolers

3. Supply cooling water to the containment air compressor aftercoolers and water jacke,ts

4. Supply cooling water to condition the air in various equipment areas such as service building, safeguards area, auxiliary building, pipe tunnel, turbine building basement, and those areas in the station which are normally occupied.

9.4.1.3 Neutron Shield Tank Cooling Water Subsystems Subsystem is designed to circulate and cool the water in the neutron shield tank.

9.4-2

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INSERT 1 i

Additional analysis was performed for a maximum river. water temperature of 90*F. At this temperature,.the component cooling water system will supply a maximum of 104*F water with all components operating as designed. The time to reduce the temperature of the Reactor Coolant to 140*F after a reactor shutdown would increase by approximately 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

l l

BVPS-1-UPDATED FSAR Rev. 0 (1/82) l of the spent fuel pool for short periods of time. Various i

loss-of-cooling in Section 9.5.

conditions for the spent fuel pool are analyzed i water may be at The fuel pool cooling assumes component cooling elevated temperatures, or unavailable, i periods of time. for Excessive temperatures do not result from temporary inadequacy of the component cooling water system. In addition, heat is removed by evaporation of water from the spent fuel pool with intermittent make-up. No hazardous conditions can develop.

For all component cooling load conditions, one primary plant component cooling water pump provides the flow required to maintain the component cooling water supply temperature below 100 F for conditions where the river water temperature is less than 78 F.

Underx normal station operation component cooling load conditions (55 10' Btu per hr) , when the river water temperature exceeds 78 F, one river water pump, two primary plant component cooling water pumps, and two primary plant component cooling water heat exchangers are required to maintain the component cooling water supply at less than 100 F. For extreme i 4

component cooling water loads which exceed 55 x 10' Btu per hr and river water temperatures in excess of 78 F, the third heat exchanger and/or pump provided in this system can be used to maintain 100 F.

the component cooling water supply temperature below 90 During normal cooldown, with river water temperature at @ F, the {

component cooling water system requires use of two primary plant component !

cooling water pumps, cooling water heat exchangers, three primary plant component two residual heat removal exchangers, and two river water pumps to cool the reactor coolant from 350 F to 140 F in 20 hr. A slower, but acceptable cooldown may be accomplished under all conditions when any single pump or heat exchanger is not available for service in either component cooling or RHR system.

Air-operated trip valves are installed in the .,utlet cooling water lines from the reactor coolant pumps' thermal barriera. A check valve is installed in each inlet cooling water line to the thermal barriers. In the event that a leak occurs in the thermal barrier cooling coil, an alarm annunciated in the main control room and the automatic high pressure reactor coolant is safely contained by closure of the appropriate isolation valve. The air-operated stop valves in the cooling water lines leaving the reactor containment, and in the reactor containment air recirculation cooler lines leaving the reactor containment close on containment isolation phase B signal (Section 5.3).

The parts.

neutron shield tank cooling water subsystem has no moving Malfunction can occur only by loss of water through

.eakage, loss of natural convection circulation through blockage, or low heat transfer due to fouling. Malfunction is indicated by a low expansion tank level alarm or by high shield water temperature indication. If no operator action is taken, neutron 9.4-14

r BVPS-1-UPDATED FSAR Rev. 2 (1/84)

3. One control room air conditioning condenser or one control room river water cooling coil (Section 9.13)

4. At least one emergency diesel generator cooling system heat exchanger (Section 8.5) .

The river water system is designed handle the above loads at a maximum water level river water temperature of 86 F at either the low river of El. 648.6 or the Probable Maximum El. 730.0. Flood of The river water system is designed as a Seismic Class I system (Appendix B) and is tornado and missile protected (Section 2. 7) until the discharge from the various heat exchangers enters the turbine building. As discussed in Appendix D, there are no high energy lines located within the intake structure, this minimizes ,

the potential effects of pipe breaks. The closed fire doors connecting the adjacent compartments minimize the amount of water entering an adjacent compartment. Whatever water enters the compartments sump pumps. by passing around the fire doors will be removed by The system is engineered and designed so that all components, pumps, and heat exchangers can be individually isolated thus providing maintenance.

for continual operation during equipment repair and 9.9.2 Description i

The river water system is shown in Figure 9.9-1A & 1B.

The cooling requirements are achieved by the provision of three river water pumps. Each is able to deliver approximately 9,000 gpm and is designed to supply the quantity of water needed for the essential safety-related cooling requirements for all unit operating conditions.

The river water pumps are a-c motor driven vertical wet pit-type units. They are mounted above and take suction from three i separate sections of the screenwell. The pump motors are in cubicles which are protected from flooding by the Maximum Flood. Probable The intakes of the pumps are at El. 640 ft I 7 inches which is below the low river water level. Figures 9.9-2 and 9.9-3 show the arrangement of the screenwell in which the river water pumps are located. The three pump motors are powered from the emergency buses. One pump receives power from one of the emergency buses, the second pump from the other emergency bus. The third pump is not normally connected to either bus, but can be manually connected to either.

both emergency buses at the same time. It cannot be connected to 9.9-2

I' BVPS-1-UPDATED FSAR Rev. 3 (1/85)

At the minimum possible river elevation (648.6 f t) , the river flow assumes open channel flow characteristics at the rate of 4,000 cfs (1.8 million gpm). River water system ' low requirement for safe shutdown is approximately 20 cfs(9 thousand gpm) or j

O.5 percent of flow available.

At minimum flow conditions and at minimum river level, the Ohio River can fully meet the cooling water requirements of BVPS-1.

Further, assuming that the Shippingport Power Station and BVPS-2 require the same amount of cooling water, only 1.5 percent of available river flow would be required for all three nuclear power stations.

The Ohio River, the Ultimate Heat Sink for the BVPS-1, meets the criteria of Safety Guide 27 as discussed in Section 1.3.3.27 The minimum River Water System flow of 8,660 gpm shown for a DBA in Table 9.9-3 is calculated on the basis of an extreme 1 w-pool level of El. 648.6, a maximum river water temperature of 86 F and a pipe flow friction factor applicable to the end o life condition after forty years of operation. 90 The entire system is designed as a seismic Class I system (refer to Appendix B) and is tornado and missile protected up to the entrance into the turbine building. If this piping in the turbine building were lost, the river water system would continue to operate. The turbine building basement would be filled with river water up to El. 708 ft. Water would then spill out the tube withdrawal opening in the north wall of the building and flow into the north yard and downhill to the river.

River water to all equipment is supplied via one 24-inch river water header. A completely redundant 24-inch river water header is provided to ensure a water supply in the event of a pipe rupture in the operating header. Duplication in the river water system of items such as pumps, the screenwell structure cubicles, the main pipe headers from the screenwell, and essential valves assures that the system will meet the single failore criterion (Section 1.3.1). All components, pumps and heat exchangers can l be individually isolated, thus providing for continual operation during equipment repair and maintenance. Because of the duplication in the system, any component can be serviced while the system maintains sufficient cooling capacity to keep the unit in a safe condition. The alternate intake structure (Section 9.16) also ensures a continuous source of cooling capability in addition to the above system. Any single failure of the system can be detected by various instrumentation, such as temperature sensors, flow meters, etc., located throughout the system. Any failure can be isolated by isolating that portion of the failed system or by isolating the antire header and operating on the redundant header.

9.9-7

17 7 01. EU G@45U 63W 1

! BVPS-1-UPDATED FSAR Rav. 3 (1/85) i l

The maximum calculat'ed pressure plus thermal stress in the safety injection nozzle during the safety injection transient was calculated to be approximately 50,900 psi. This value compares favorably with the code allowable stress of 80,000 psi.

Safety injection transients are considered along with all the other design transients for the vessel in the fatigue analyses of the nozzles. The analysis shows the estimated usage factor for the safety injection nozzles to be 0.47 which is well below the Code allowable value of 1.0.

The safety injection nozzles are not in the highly irradiated region of the vessel and thus they are considered ductile during the safety injection transient.

The effect of the safety injection water on the fuel assembly grid springs has been evaluated, and due to the fact that the ,

springs have a large surface area to volume ratio, being in the form of thin strips, they are expected to follow the coolant temperature transient with very little lag; hence, no thermal shock is expected and the core cooling is not comprised. .

Evaluations of the core barrel and thermal shield have also shown that core cooling is not jeopardized under the postulated accident conditions.

14.3.4 Containment Evaluation 14.3.4.1 Design Ba Geciden d her the purposes m evaAuating the effects of a LOCA, the DBA is defined as follows:

1. The reactor is assumed to be operating at maximum calculated core thermal power (2,766 MWt) and to have been operating at this power long enough to have reached its equilibrium concentration of fission products.

2. IFor the reactor evaluation, an instantaneous double-ended displacement rupture is assumed to occur in the cold leg since this is more severe than a break ,

in the hot leg.

For the containment evaluation, the instantaneous double-ended displacement is assumed to occur in the hot leg, the cold lee or the reactor ecolant pump suction line.

A AIb1F $ 14.2-22

h ,

17781.30 US(BT 256 i'1 .. j

't INSERT C The design of _ the subatmospheric containment structure is based on . the following criteria:

1. The peak calculated containment. . atmosphere pressure shall not exceed the design pressure of 45 psig.

2. The containment shall be - depressurized following a design basis accident to below 1 atm absolute pressure in less than 60 minutes.

3. Once depressurized, the containment shall be maintained at a.

pressure less than 1 atm absolute for the duration of the accident.

The peak containment pressure due to a postulated loss-of-coolant accident (LOCA) occurs after a double-ended rupture (DER) of a reactor coolant pump discharge line and is a function of the initial f,otal pressure and average temperature of the containment atmosphere, the containment free volume, the passive heat sink inventory in the containment, and the rates of mass and energy released to the containment. The passive heat sinks in the contain-ment are considered to be at the same initial temperature as the initial average containment atmosphere temperature, Maximizing the initial contain-ment total pressure and average atmospheric temperature maximizes the calculated peak pressure.

20732c-1778110-s4 1

' 17 7/ B l . d U US ( 0)- ~256 ANSERT C (Cone)

The- time ' required. to depressurize the containment and the ' capability to maintain it depressurized below 1 atm after a pump suction double-ended rupture (PSDER) depends on the mass of air in the containment, on the design-of the containment. depressurization system. (both quench spray and h /c.r recirculation spray subsystems, see Section 6.4), and on the servtee water Yt v t.t temperature. When ' the ,c rv k e water temperature is high, it is more difficult- to depressurize the containment in 60 min (criterion 2). There-fore, during plant operation the maximum permissible containment air partial pressure is specified . in -the Technical Specifications ' as a . function of ri v t i" m vice water temperature.

rnftr In summary, the containment structure is sized for the cold ---"'- water conditions (criterion 1) and the containment depressurization system is riv er sized in accordance with criteria 2 and 3 for the warm ---"'-- water conditions.

The reactor is assumed to be operating at the maximum core thermal power of 2713 Mwt and to have been operating - at this power long enough to have reached its equilibrium concentration of fission products. Coincident with the LOC A , a complete loss of all offsite electric power is assumed. One emergency diesel generator starts and operates to supply emergency power.

This results in activation of the minimum engineered safety features.

The minimum engineered safety features that are assumed to be activated to limit the consequences of the LOCA are as follows:

1 l

i 20732C-1778110-B4 2 L_______.-__ _ -- _ --

'17 7 81.17.1 nsERT cVMAl? . 256 p

L1. . . Emergency core cooling-by.

.a. All of three nitrogen-pressurized accumulators-

, c;.c b .' 'One out of three charging pumps

c. 'One out of two low-head safet'y injection (LHSI) pumps-

2. Containment depressurization by

a. Or.a .'out of two trains of the quench ~ spray - subsystem and'

b. One out of two trains of the recirculation spray subsystem (i.e., one inside recirculation spray pump and one outside recirculation spray pump).

The emergency diesel generator provides the power to operate the pumps. The accumulators are passive and discharge into the RCS when the RCS pressure drops to below the accumulator pressure.

I 20732C-1778110-B4 3 l

l l'

- _ = _ _ _ _ _ _ _ _ _ _ _ _ _ _

us(BF 17~ 7 81.10BVPS-1-UPDATED'FSAR 256 Rov. 3 (1/85) 3.. An overall loss of electric power on the site occurs, (out of two) emergercy diesel generators.has

.and oneand started is' operating to supply emergency power.

4. Minimum engineered safety features' are 'then activated

-(Section 6) to limituthe consequences of the accident by providing the following:

a. Three gas. pressurized accumulators discharge into the reactor coolant system when. the reactor vessel pressure falls below 600 psig.

b. Safety injection by: -

'One of'two charging pumps and one of two low head safety injection pumps.

c. Containment depressurization by:

One of two quench spray subsystems and two of.

four recirculation spray subsystems.

The emergency diesel generator operating as described in item 3 provides the necessary power to operate the pumpe in Items 4b and 4c; 4a does not require power of any kind to operate.

The design containment inleakage rate is assumed to be the same as design outleakage rate: 0.1 percent of the. containment volume per day. This amount of inleakage has a negligible e!.fect on the containment pressure transient during the term of the analysis; accordingly, it is neglected.

Compliance with Appendix J, 10CFR50, will ensure that the design leak rate will not be exceeded (Section 5.6). Following a loss-of-coolant accident, after containment depressurization, the containment pressure can be monitored to verify that the pressure remains subatmospheric.

Subsections 14.3.4.2, 14.3.4.3 and 14.3.4.4 discuss and snalyze the effects on the containment of a LOCA involving a double-ended displacement rupture of the reactor coolant piping. Other smaller LOCA's are also discussed.

The effects of the DBA and the RCS are discussed and analyzed in i Subsections 14.3.1, 14.3.2 and 14.3.3. Subsection 14.3.6 summarizes the consequences of a DBA on both the RCS and the 4 containment. ];

14.3.4.2 Method of Analysis Effects on Containment j Enalyses of the effects on the containment of LOCA are made using

\ s

g(k 14.3-23

17781T10 US(B) 256 INSERT D Analyses of the effects on the containment of high energy line breaks are l made with the LOCTIC digital computer program.15 This program calculates the temperature and pressure of the containment atmosphere as a function of i time following a LOCA or a main steam line break (MSLB) inside containment.

The program considers the various heat sources and sinks as a function of time to calculate the temperature and pressure transients of the containment atmosphere. The calculational process is a digital integration of the changes taking place. The program assumes that only small changes take place during each time step and that calculations during each time step are 1 on a steady-state basis. At the end of each time step, the heat inflows and outflows are summed and new containtnent and primary and secondary system conditions established.

For a LOCA, the program reads into the computer the input information i

required to specify the containment and the RCS. Table 14.3-5 lists LOCTIC input data to describe the RCS, containment, and depressurization systems.

Also read in are the physical constants for the containment and the RCS materials, as given in Table 14.3-6.

20732B-1778110-B4 1 I

i

~

17 ? bl . 9.Rstar oTeknT . IN The' following. is a description = of the major points considered by' the LOCTIC computer program:

Fission Product-Decay Heat The LOCTIC program interpolates a curve representing the decay heat generat-ed as a . fraction of full operating power versus time after the accident.

The curve is calculated as prescribed in American Nuclear Society Standard

' ANSI / ANS-5.1 1979" .

This ' curve is based on an equivalent of 1095 full-power days _(3 years) of '

operation prior to shutdown and . includes the contribution .of heat released by the heavy elements. For conservatism, and in order to allow for uncer-tainties, the nominal values for fission product decay heat are increased by

.two standard deviations (2 sigma).

Power Coastdown Heat

' After receipt of a reactor trip signal caused by the LOCA, the reactor power decreases - to fission product decay levels over a - finite period of time, depending upon the time it takes for the control rods to drop, the rate of boron injection, the half-life of the longest-lived delayed neutron precur-l sor, moderator and fuel temperatures, and moderator levels in the reactor vessel during blowdown. The power coastdown curve is shown on Figure 14.3-54.

I i

207329-1778110-B4 2 l

_ _ _ _ _ _ _ _ _ - - - _ - _ - . _ - _ - - .- l

E 2^$

1 7 7 0 1 . 2 MSeaT J 80W /

c The heat from fission product decay and power coastdown during the time interval under consideration is computed and added as sensible heat to the reactor core.

Core Sensible Heat The reactor thermal power is 2713 MWt, and the heat in the core is 15.68 x 10' Btu above then average coolant temperature.

Since the core contains considerable heat at a temperature above the average reactor coolant temperature prior to the accident, the program computes the heat transfer from the core to the reactor coolant as a function of the amount of water in the core and determines whether nucleate boiling or film boiling heat transfer is occurring. This relationship is conservative because more heat is transferred to the reactor coolant during blowdown and, hence, to the containment than would be expected to occur. The program adds fission product decay heat and power coastdown heat to the core sensible heat, which is transferred to the coolant for transport to the containment through the break. Heat from the zirconium-water reaction is not included in containment analysis since cladding temperature is much less than that which would result in zirconium-water reaction.

Reactor Coolant System Hot Metal Sensible heat is transferred from the RCS hot metal to the reactor coolant.

This is a transient heat transfer calculation, and the Dusinberre numerical 20732B-1778110-B4 3

r I'E T U E . *SkNSERT N kd8b[) 8d )D '

1 method ! ' is used for solution of the problem. The RCS hot metal is divided-

[< into the following cases'for' analysis':

1. Ruptured loop piping..

2. Ruptured loop metal, two cases.

3. 0ther loop piping.

-4. Other loop metal, two cases.

5. Reactor vessel head.

6. Reactor vessel shell and thermal shield.

7 Reactor vessel bottom.

8. Pressurizer and surge line.

- The metal of the valves, pumps, and steam generator heads is included in the two thick metal cases for.the ruptured loop and other loops. Table 14.3-7 presents the equivalent thicknesses and amounts of RCS hot metal. When a metal, is covered with water, the surface temperature is ' conservatively set equal to the water temperature. When the metal is uncovered, heat outflow is controlled by an input heat transfer coefficient.

' Accumulators The water flow from the accumulators to the RCS is calculated on the basis of differential pressure and an input flow coefficient. A differential mass transfer is calculated and added to the reactor coolant inventory. The loss of the accumulator water results in an increase in the accumulator gas volume. The pressurized nitrogen gas is assumed to be a perfect gas and is expanded to the new volume, with a resulting change in the accumulator 207328-1778110-B4 4

1 N U I

S kNSERT D C )

l R .

. pressure. When the accumulator water is completely discharged, the pressur-ized gas exhausts to the containment. .

'i High-Head Safety Injection .

High-head safety injection is accomplished by the charging pumps and becomes-effective 30 sec'after a'LOCA. Initially water is transferred from the RWST I to the.RCS as a function of RCS pressure. Differential mass and heat flows are. calculated and added to the coolant inventory. The effective delay time for high-head safety injection is input to the LOCTIC program' .

1 Low-Head Safety Injection The treatment ~of LHSI is similar to high-head safety injection, with a system flow curve used to calculated differential mass flow . to the RCS.

This mass and associated heat are- added to the reactor coolant inventory.

When the appropriate level is reached in the RWST, water-is drawn from the containment' sump instead of the RWST.

Reactor Coolant Blowdown A summation is made at each time interval of all heat and mass inputs to, and outputs from, the RCS, and new system conditions are established. A j differential discharge of mass and heat from the RCS to the containment is calculated. The flow rate is based on Henry-Fauske subcooled and friction-less Moody saturated critical flows at high BCS pressures. It is assumed j- that this discharge reaches equilibrium with the containment steam-air 20732B-1778110-B4 5

)

I

17:7 01 4M NWU CO "'

. INSERT D (Cont) l.

atmosphere, ' and that any water remaining as liquid. after flashing takes

place falls to the containment ~ floor where it mixes with water on the containment floor. j Core Reflooding 1

i l

ll At. the end of blowdown, the water level is assumed to be at the bottom of the active core. The water in the lower plenum is saturated liquid at the reactor vessel pressure. At this time, both the core and downcomer.begin to .;

fill. The water from the downcomer mixes with the water in the lower plenum.

The downcomer driving head, average core heat transfer coefficient, flow split between . broken and intact loops, and core flooding rate are. all controlled by transient data specified as input to the program. These data are supplied as a function of time after reflooding begins.

If the safety injection pumps come on before the accumulators empty, mixing is assumed to occur in the unbroken cold legs and/or the inlet plenum so that the liquid entering the downcomer is at a mixed mean enthalpy.

The heat stored in the fuel rods and generated by fission product and heavy-element decay is transferred to the coolant passing through the core during reflooding and is calculated with the input-specified core heat transfer coefficient data.

207328-1778110-B4 6

m------ _-__ __ _

V w

D e 17 7 81. llbRr M bN 256 l

=The heat addec to the two-p s' ' ketween the core and the steam genera-tor comes J from two metal' sources: the vessel head and other vessel parts i

above the nozzles; and thin metal, thick metal, and piping-in the ruptured _

loops. .-

In. the steam generator, heat stored in the secondary liquid is transferred to the two-phase coolant coming from the core. Whatever liquid is' presentf

~

in the' primary flow is vaporized, and superheated steam exits from the steam generator and flows to the break.

Superheated steam from the steam generators in unbroken loops must pass through cold legs and the reactor vessel inlet plenum in order to reach the rupture in the broken cold leg. In so doing, it passes.the points at which high- and low-head safety injection water is introduced into the reactor coolant system.

Reflooding ends when the water level in the core reaches the 10-ft elevation (active core height = 12 ft). Use of the 10-ft elevation to terminate reflooding presumes the core axial power shape to have been a cosine distri-bution prior to the accident. PWR FLECHT tests, in which a cosine distribu-tion is assumed, have shown the entire core to be cooled down, i.e., all of the stored heat to be removed by the time the water level reas he< 10 ft."

Post-Reflood During the post-reflood period, the mass and energy release rates provided by Westinghouse for a larger plant are modified by LOCTIC as described 20732B-1778110-B4 7

L' I F 7 d l . J. U - UMt91 2% W INSERT D (Cont)z m.

[

k below'.

L The steam' generator. energy is released to'the containment atmosphere'in two-stages . referred to as the " equilibration stage" '. and the' "depressurization stage." In the former, the energy sources above the reference pressure used

. in calculating the mass and energy.' releases are brought into equilibrium with the containment' pressure. The rate for this phase is set by the Westinghouse froth calculation model. In the latter stage, the sources give up additional energy as . the containment ' pressure decreases. The rate for this' stage is set by the containment depressurization rate.

The intact loop steam generators and metal energies are lumped together for this calculation. ,

After the post-reflood froth period, LOCTIC computes the mass and energy release (boil-off) to the containment atmosphere, which is governed by decay heat.

I' Broken Loop Steam Generator-Equilibration Stage The mass and energy release rates into the containment during the time prior to broken loop steam generator equilibration (t'eq(b1)) are presented in Table 14.3-15.

At the time (t'eq(b1)) when the broken loop equilibrates with the reference pressure (P'), the LOCTIC calculated containment pressure may be less than the reference pressure used by Westinghouse for generating the mass and I

207328-1778110-B4 8 ,

l

I~

, 17 781l.14Sen EM 2.56 u, energy.' releases. If so,.t extension of the broken loop equilibration stage-is required. The tota 1' additional energy which must be transferred from the

broken loop at any time during the' extension of-the equilibration stage is

Y O

eq(b1)

ay(bi) f jp where:

i:

- AE eq(b1)

= . Broken ' loop, steam generator secondary energy to be removed L during the extension of the broken loop equilibration stage.

E ay(b1)

= The total energy (Btu) available in the broken loop steam generator secondary at the reference pressure relative to 212F and.14.7 psia.

P* =' The containment reference pressure _ used in the NSSS vendor mass and energy release analysis (psia).

P,g

= Assumed minimum LOCA containment pressure (11.5 psia)

APg is defined as AP g = P* - P g ; i=1 AP g =

Pg ,g - Pg ; i>1 and "i" defines the i th time interval following the reference broken loop equilibration time (t'eq(b1))'

207328-1778110-B4 9

' CO .

!NSERT D (Cont) 17781. 10

~

us(s) 256 At the'. end of each. interval, a new equilibration time is calculated by'--

t eq(b1)'= t' q(b1) e + AEea(b1) 4*(b1) t,q(37) = Broken loop equilibration time based on calculated containment pressure response

= Reference equilibration time in the broken loop t'eq(b1) q'(b1) = Sec ndary-to-primary heat transfer rate in the broken loop at t'eq(b1)

When t eq(b1) is calculated to be equal to the current time after accident initiation, the broken loop is assumed to be equilibrated.

The energy rate to the containment from the broken loop between. t' eq(b1) and t eq(b1) is

! assumed to remain constant at q'(b1)*

Broken Loop Steam Generator - Depressurization Stage l

During the depressurization stage, the steam generator secondary is brought to the ambient conditions in the containment. The secondary energy which remains after equilibration is:

E dep(b1)

ay(b1) ~ 0 eq(b1) 20732B-1778110-B4 10

(.

^

c/ .

'TNSERT D (Cont)--

@l" where:

.17781. 10 US(B) 256L E

dep(b1)

nergy which must be transferred from the broken loop steam' -

generator secondary during the depressurization stage.

E ay(bl)

= The total energy available in the broken loop steam generator secondary at : the reference. pressure relative . to 212F and

'4 14.7 psia.

6E eq(bi) : Energy transferred from the -~ broken loop ' steam generator se ondary - during the extension of the ' equilibration stage.

The energy release rates during this period are those basad on the reference pressure (P*) plus an additional energy increment due to.depressurization of the secondary system as'follows:

4E E dep(bl) : ay(bl)fp,,0p lo where:

aE dep(bl) = The energy transferred from the steam generator secondary during a time increment, and AP = The change in containment pressure during the previous time increment.

207328-1778110 B4 11

F:- - M o' INSERT D (Cont) 17781. 10- Us(B) 2 %' ~

The additional mass increment is then calculated by:'

!!. ag = 8P01) fg -

where h pg is the latent; heat of -vaporization at the current. containment conditions.

' Intact Loop Steam Generators - Equilibration Stage The- same procedure is used as for the broken loop. However, metal and core energy . are lumped with the steam generator . energy for this calculation, t

The equations are the same as those for the broken loop except. that the;

- subscript "(11)" replaces the subscript'"(bl)".

- Intact' Loop Steam Generators - Depressurization Stage Again the procedure used is the same as for the broken loop case except that

}. decay heat, qdcy, is added to the heat adcition rate.

The additional mass rate to be added is then dcy 6

dcy h fg 207328-1778110-B4 12 I

p

' INSERT D (Cont) 17781. 10 Us(By 256 Thermodynamic State i

.- 1 p'

l j LOCTIC calculates the temperature and pressure of the containment atmosphere i

'. as a function of time following a high energy pipe break accident. This calculation assumes the containment atmosphere to be a homogenous mixture of steam and air in thermal equilibrium.

Condensation Heat and Mass Transfer No distinction is made between the condensing coefficients for water vapor on steel and painted concrete surfaces. Although Ref. 76 states that the experimentally measured heat transfer coefficient on concrete surfaces 'was found to be 40 percent of the value measured on steel surfaces, the same

~

reference also states that "a painted concrete surface can allow dropwise condensation and, therefore, ' have a heat transfer ' coefficient comparable to the value for steel." All concrete surfaces which are exposed to the containment atmosphere are painted.

A thermal conductance based upon the paint flim thickness is also used in the heat transfer model. The thermal resistance of the paint film is in series with that of the condensing film. The values of thermal conductance utilized in the containment analysis are given in Table 14.3-6.

Loss of Coolant Accident l~

The Tagami heat transfer correlation is used in the containment LOCA analy-p sis. The use of the Tagami coefficient as described below differs slightly l l

207328-1778110-B4 13 1.

1

TNSERT D'(Cont) 17781110

.us(B) 256

.from' Reference 76 in that the peak condensing coefficient is taken.to occur at the. time of the first peak pressure rather than at the end of blowdown.

~

e This 'is. logical since the steam / air ' ratio is greater- at. the time of peak pressure- than at the end of blo6.down. The maximum heat transfer coeffi-cient, which occurs .at the first containment peak pressure, is given by:

3 0.6-h

[E P l max =75(Vt-) p where:

h,,, = Tagami heat transfer coefficient, Btu /hr-fta_p Ep : Energy released to the containment atmosphere at the time of the first peak pressure, Btu V= Containment free volume, ft 3 t = Time of first peak pressure, sec p

75 = Empirical constant (see Reference 76)

Before the first peak pressure is reached, the heat transfer coefficient is calculated as:

h=h ,(t/t p) where t is the time in seconds after the accident.

20732B-1778110-B4 14

mm . ,. -

c :p f' g INSERT D (Cont) 17781. 10 Us (B)' 256

.After the . first : peak . pressure is . reached, the heat transfer' coefficient 3

. equation takes the following form (Ref. 76):

hah stag + (h max -h stag}'

E where:

h Stagnati n heat transfer coefficient = 2+ 50x atag and

-Steam / air ttssa ratio.

~

x =

When the temperature of the containment atmosphere is less.than or equal to the heat sink surface temperature, a natural -convection coefficient of 1.8 Btu /hr-ft -F is used. The program checks the surface temperature 'of each heat sink slab and uses the condensing coefficient or natural convection coefficient as applicable. The coefficients used in the containment LOCA analysis are shown in Figures 14.3-45 through 14.3-47.

Static Heat Sinks The static heat sinks include the containment structure, interior concrete, and miscellaneous metal in the containment structure. The Dusinberre method is used to calculate the transient heat flow into these sinks. The containment structure and its internal concrete are divided into the follow-ing cases according to thicknesses:

207328-1778110-B4 15

__. _ -_ _-__--____________-_--_____O

(C INSERT D (Cont) 17781. 10'. US(B)' 256

1. Containment structure shell below grade.

L.- -

2.- Containment structure shell above grade.

3. Containment structure dome.

4 Containment structure floor above floor liner.

5. Containment structure mat below floor liner.

6. Internal concrete slabs, five cases.

7. -Several classes cf miscellaneous metal inside the containment structure.

! The analytical model used considers transient heat transfer to the contain-ment structure through the overall thermal resistance made up of the paint film on'the steel liner, the liner itself, and the concrete. The Dusinberre method allows the face temperature to lag containment atmosphere temperature as .would be expected under actual conditions. Heat conducted through the containment cylindrical wall and dome to _ the outside air is considered, but is minor. Tables 14.3-8 and 14.3-18 present the information used by LOCTIC for interior concrete sinks, the containment, and miscellaneous metal.

Resistance to heat transfer at the liner-concrete interface is considered by

-the use of an interface thermal conductance coefficient of 10 Btu /hr/ft jp, 2 This is conservative since the steel liner is used as a form for the pouring of the concrete, so the non-matting surface irregularities and gaps that characterize thermal contact resistance are absent.

In calculating the heat absorbed by the static heat sinks, the LOCTIC computer program uses the explicit method of Dusinberre!8 Since the LOCTIC program allows input specification of the time step size to be used in the l 207328-1778110-B4 16

--- --___---------___j

[

'INSEM D (Cont)

V:

17.781. 10 US(B) .256

.c calculation, the maximum heat sink slice thickness consistent with this o method must be calculated as follows:

DX = (AMot) I#

where:

l DX = slice thickness A- = ' thermal diffusivity, K/pc p .= material density K = material thermal conductivity c a material heat capacity M = dimensionless modulus, (DX)2/A/ot at = time interval In order that the method be numerically stable, .M must be equal. to or greater than 2. A value of 4 is used.

Sufficient mesh points are used to include initially the entire steel liner as well as 5.5 to 6 in, of the concrete immediately behind the liner or the first 5.5 to 6 in. of concrete where no liner exists on interior- concrete walls. As the temperature increase penetrates the heat sink, the mesh points are extended farther into the material. Approximately 100 points are-established at all times, and 6 in. of material is added at a time (if possible). Upon each addition, the mesh point spacing is reestablished, always within the requirements of the previously stated method.

l l

l i 20732B-1778110-B4 17 l.

1

17781. 10 US(B) 256 BVPS-1-UPDATED FSAR R3v. 3 (1/85) the LOCTIC digital computer' program described in SWND-1 l13I.

This is the same version of LOCTIC that was used to analyze Surry Power Station Units 1 and 2 and North Anna Power. Station Units 1 and 2, and that was described in reports transmitted by f Stone r. Webster on December 6, 1971 to the AEC/DRL. The program calculates the temperature and pressure of the containment as a-function of time following a LOCA. The effect.s of various heat sources and sinks as a function of time are considered in calculating temperature and pressure transients for the containment. The method is a digital integration of the changes .

taking place. The program assumes that heat and mass flow rates I are constant during each small time step. At the end of each time step, the heat inflow. and outflow are summed and new containment and RCS conditions established. The distribution of energy in the containment following a DBA is given in Table 14.3-4.

Conser'vatism in the mass and energy releases to the containment during blowdown is ensured by margins in the .following input variables:

1. Liquid level in the pressurizer is taken to be at the upper limit.

2. The sensible heat stored in the core contains a safety ~ J

-margin.- '

3. A double-ended rupture of the largest primary coolant pipe is assumed.

4. The time at which DNB occurs is not considered. In fact, no reduction in heat transfer coefficient from the core to the coolant is made to account for DNB.

The heat transfer 'ecef ficient from the core to the coolant is based on the assumption that:

a. The fuel-clad thermal contact resistance, the' fuel pellet thermal resistance and the clad thermal resistance remain unchanged.

b. The convective heat transfer coefficient from clad to coolant is given by the maximum of: (1) two phase convective boiling, or (2) pool boiling heat transfer coefficients. The equation for heat transfer coefficients appears on page 3.1.5-2 of Reference 37.

Figure 14.3-70 shows the :cre-to-coolant heat transfer coefficient as a function of t .. .e calculated during the blowdown through a cold leg pump suct.cn DER. These heat transfer coefficients result in the rencval of 70 percent of the available core sensible heat by the time reactor vessel blowdown ends (3ei, 14.3-24

BVPS-1-UPDATED FSAR Rev. 3 : (1/ 8 5) r .

The program reads into the computer the information required to describe . the specific . containment and~ - RCS under investigation.

Table ,14. 3-5 gives input data for BVPS-1. . Also . read are the physical co_nstants for the containment .and coolant system materials as given in Table 14.3-6.

Containment Free Volume The containment free volume is an input to-the LOCTIC program.

The containment free volume is determined by first computing the gross volume of the containment and then computing and subtracting the volumes of all' components-and structures within the containment from the gross volume. The volumes are computed from the dimensions obtained from the containment layout drawings. It is estimated that the maximum uncertainty 'in free volume is plus or minus 5 percent. . The following are the results of a sensitivity study of the effects of uncertainty in the containment free volume on peak containment pressure:

Free Volume, Peak Pressure, Run ft3 esig

-5 percent 1.71 x 106 42.79 Best estimate 1.80 x 106 4o,79 j

+5 percent 1.89.x 106 38.94 The sensitivity study was run for the cold leg DER under winter conditions.

The following is a description of the major features of the LOCTIC computer program. !

I Fission Product Decay Heat and Power Decay The LOCTIC program interpolates a curve representing the decay heat generated as a fraction of full operating power. For conservatism, it is assumed that during the early stages of the LOCA '(up to 1 hr) , the decay heat generation' rate is equivalent to-that resulting from fission products in the core as a result of infinite core exposure. After approximately 20 hr, the fission product concentration is assumed to be that equivalent to 3 yr of_ core exposure. Figure 14.3-53 presents the decay curve used by LOCTIC for BVPS-1. The curve presented is in good agreement with the proposed ANS standard of subcommittee No. 53 published June 11, 1968 and revised January 1969.

After receipt of a reactor trip signal caused by the LOCA, the reactor power decreases to fission product decay levels over a finite period of time, depending upon the time it takes for control rods to drop, the rate of boron injaction, the half-life of the longest-lived delayed neutron precursor, moderator and ~

fuel temperatures, and moderator levels in the pressure vessel 14.3-25 i

17781.BVPS-1-UPDATED 10 US(BT FSAR 256Rav. 3 (1/85)

M during blowdown. Power coastdown curves have been generated for the, case of a 29 inch double-ended rupture and the 11.2 inch pressurizer surge line break. Figure 14.3-54 presents these curves used by LOCTIC for BVPS-1.

The heat from fission product decay and power coastdown during the time interval under consideration ia computed and added as sensible heat to the reactor core. Ganca heating frem fission product coastdown and power decay is deposited in the core metals.

Core Sensible Heat

Since the core contains considerable heat at a temperature above the average coolant temperature, the program computes the transfer of heat from the core to the coolant as a function of the amount of water in the core and whether boiling or convection heat transfer is occurring. This relationship is conservative frcm the standpoint of transferring more heat to the containment faster than would be expected *to occur; however, the relationship is invalid for the study of core thermal effects. The codes discussed in Section 14.3.1 are utilized for the latter analysis.

The LOCTIC program adds all' fission product decay and power coastdown heat and all the metal-water reaction heat to sensible heat in the core. This sensible heat is transferred frem the core to the coolant for transport to the containment. -

Reactor Coolant System Hot Metal 1

Sensible heat is transferred from the RCS hot metal to the coolant. This is a transient heat transfer calculation, and the Dusinberre numerical method, discussed below, is used for solution of the problem. The RCS metal is divided into the following ten cases:

Ruptured loop piping Ruptured loop metal, two cases other loop piping other loop metal, two cases Reactor vessel head Reactor vessel shell and thermal shield Reactor vessel bottom l Pressurizer and surge line.

The metal of the valves, pumps and steam generator heads is included in the cases for two different metal thicknesses for the ruptured loop and other loops. Table 14.3-7 presents the

14.3 to v

1/ 7 61.1 U US(B) 256 BVPS-1-UPDATED FSAR Rev. 3 (1/55) equivalent thicknesses and amounts of metal to be treated by Dusinberre method (161 for RCS hot metal. When the metal, in any of the above cases, is covered with hot water, the surface temperature is set equal to the water temperature. When the metal is. uncovered, heat cutflow is controlled by an input heat transfer coefficient.

f Accumulators The water transfer from the accumulators to the RCS is calculated on the basis of differential pressure and an input flow coefficient. A differential mass transfer is calculated and added to the reactor coolant inventory. The loss of water frem an accumulator results in an increase in gas volume. The driving gas is assumed to be a perfect gas and is expanded to the new volume, with a resulting change in accumulator pressure. When the accumulators are completely discharged, the pressurizing gas exhausts to.the containment. l High Head Safety Injection Safety injection is as sumed . to be effective at an input time, after which water is transferred to the coolant system as a function of RCS pressure (See Table 14.3-5). Differential mass and heat flow are calculated and added to the coolant inventory.

High head injection is accomplished by the charging pumps.

Appropriate delay times for receipt of a safety injection signal, valve ope' ration, and pump start are input to the LOCTIC program.

Low Head Safety Injection ]

The treatment of low head safety injection is similar to high head safety injection, with a system discharge curve used to calculate differential mass flow to the coolant system (See Table 14.3-5). This mass and associated heat are added to the coolant in.ventory.

l. Coolant Blowdown A summatien is made at each time interval of all heat and mass input to, and output from, the RCS, and new system conditions are established. A differential discharge of mass and heat from the coolant system to the containment is calculated. It is assumed that this discharge reaches equilibrium with the containment steam-air atmosphere, and that any water remaining as liquid after flashing takes place falls to the containment floor where it mixes with any water on the containment floor.

Condensing Film Coefficient The heat transfer coefficient for condensation at surfaces inside j the containment structure used in LOCTIC is the Tagami condensing coefficient (391, and is presented in Table 14.3-31.

14.3-27 ,

I l

l l

l L____._______ _ _ _ _ _

m = _-

1778L 10 US(B) 256 BVPS-1-UPDATED FSAR Rav. 3 (1/85) p f The. use .of the 'Tagami' condensing . coefficient is very conservative, as is vividly exhibited in Figure 14.3-71. Th measured . pressure transient was obtained frem Test. 3 of e Simulated Design Basis Accident tests of the Carolinas-Virgd la Tube Reactor ^ Containment I40i.

~

The CONTEMPT best est'.. ate calculation using the .Uchida condensing coefficient' gener es a ,

peak pressure 45 percent greater than the measured peak. he use L-- of the Tagami condensing coefficient would yield a similar result.

Better agreement with the data is achieved using the AEH average coefficient to calculate the condensing heat transf . Figure 38 of Reference 40 indicates that the TAEH average oefficient is

'four times the Uchida value at the time of peak p ssure.

.If the heat sink surface temperature is gre er than the dew point temperature, convective heat transfer o the containment atmosphere is calculated using a conservat' ely high constant heat transfer coefficient value of 1.8 u/hr-ft 2 -F. It is conservative to overestimate heat transf to the containment atmosphere. This reversal . occurs as he containment total pressure is decreasing and approaching a .ospheric pressure. The heat' transfer is considered natural conv ction to air because the mass and energy releases to the contai ment are relatively small at that time (decay heat boil-off).

Kreith(571 recommends, for ccnvecti from vertical plates in the tu'rbulent region (Grasshoff number reater than 10 10):

h =

f0.021 (Gr x P 0.4 (14.3-10)

Where: L = height of he sink h = convective eat transfer coefficient k = thermal, nductivity_

, Gr = Grassho . number Pr = Prand number McAdams tssi reco nds free convection from vertical plates for Gr greater t n 10 5,for the correlation Nu = 0.13 Gr x Pr)o.33 or h = b 0.13 (Gr Pr)o.33 L

Where: Nu Nussult number j

! The pro rties are evaluated at the mean film temperature. The i Grassh f number is proportional to the temperature difference betw n the heat sink and the containment atmosphere, and also is pro rtional to the height of the sink cubed.

14.3-28

-w-_-_.-__ _ . _ _ , . - -

7' j 177.81. 10 usj BVPS-1-UPDAipB)FSAR 2 56sGV+3 <uesi I Figures 14.3-72 and 14.3-74 present the sensitivity of .e !

I convective heat transfer coefficient to the heat sink contair ent '

atmosphere temperature difference and the heat sink. h ght, respectively. The curves demonstrate that, for very conse ative bounding values, the calculated heat transfer coeffi ent is about one-half of the value used in LOCTIC (1.8 Btu /hr # 2 7} ,

Static Heat Sinks The static heat sinks include the containment struc ure, internal

. concrete and miscellaneous metal in the containment. The Dusinberre method is used to calculate the transi4nt flew of heac into these sinks. The containment and interd.al concrete and metal are divided into the following eleven cas'es:

Containment shell below grade

/

Containment shell above grade l Containment dome Containment floor above floor 1 ear

', Containment mat below floor 1 . ear Internal concrete slabs, fi cases Miscel.laneous metal in t containment.

To establish a surface tem rature for each of the above cases, an overall heat transfer coefficient is . calculated considering the condensing coefficpnt, paint film conductivity, concrete conductivity and line@f conductivity. The model used considers transient heat flow :o the containment structure through the composite thermal re stance made up of the paint film, the steel liner and the coner te. The Dusinberre method allows the surface temperature to la containment temperature as would be expected under actual con tions. The concrete is assumed to be exposed on one or two s es to the containment atmosphere as appropriate.

Heat conducted hrough the shell to the air is considered, but is minor. Table 4.3-8 presents the information used by LOCTIC for interior con rete sinks, the containment and miscellaneous metal.

l Table 14. 8 identifies the concrete structures that are doubly exposed. Table 14.3-18 provides a tabulation of the ma]or compone s that make up the miscellaneous metals in containment used as heat sinks.

The ysical properties of structural materials, such as carbon l ste and concrete, are factored into the modeling of the st ctural heat sinks. Table 14.3-6 includes a material d ignation for each structural material used and includes an fective thermal conductivity for a paint layer 0.006 inch hick. These material designation numbers are added to 14 3-29 E-_m._m_ _ _ _ _ _ . . . _ - _ . _ _ _

17 78L 10 US(B) 256 M BVPS-1-UPDATED FSAR Rsv. 3 (1/85)

Table 14.3-8 and Table 14.3-18 to relate the material properties to the structural heat sinks modeled for program LOCTIC.

Quench and Recirculation Soray The containment depressurization system discharges water into the containment via the quench spray and recirculation spray headers.

Heat transfer between both the quench and recirculation sprays and the containment atmosphere is computed in each time interval.

4. f 2 4f The recirculation spray subsyste reject heat through coolers to river water. This he t trans r is calculated by the standard log mean delta tempe ture d ference (LMTD) calculation method using an input UA o about 2.7 x 106 Btu /hr/F for each cooler.

Table 14.3-5 gives x 10 Btu /hr/F for two coolers. An exit temperature for the recirculation water leaving the cooler is calculated. The recirculation spray headers are located approximately 80 ft above the main operating floor.

The quench spray subsystems spray chilled water from the refueling water storage tank into the containment via the quench spray headers located approximately 96 ft above the main operating floor.

Table 14.3-5 indicates that the quench spray subsystem flow rate is a function of the containment pressure. Quench spray flow is actually -determined as a function ~of the difference between contab m t total cressure and (a cifference in elevatio A UY(converted to psi) betwee3 the RWST water levelcmo ssay heac #

This.is based upon the pump head versus capacity curve supplied by the pump manuf acturer and pressure losses in lines, headers and nozzles. ,, ,, a m pegg, gf r y gp l Thequenchsprayheadersare[locatedat El. 862 ft-11 3/16 inch l and 864 f t- 3 ___ ine_h . The (MITIEEh water level in the RWST is at El. 787 ftF6 inc@ and the elevation of the (RWST outleD is 73fft. dnese elevations are shown in Figure 14.3-73'5'I M he height or tne column oT waLei for a iuAl tank is an input parameter to the LOCTIC computer program. The height of the column of water is computed throughout the run based on the summation of flows out of the tank.

A plot of quench spray flow rate worst case depressurization vs '

time from the accident is shown as Figure 14.3-76 f531 Quench and Recirculation Sorav Effectiveness The effectiveness of the quench and recirculation sprays to cool the containment atmosphere following a LOCA is greater than 99.9 percent. This is deduced from analytical results contained in Reference 41, a copy of which was transmitted to the USAEC by Reference 42.

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+ '17.7 81. 10 US(B) 256 E BVPS-1-UPDATED FSAR . Rsv. '3 ' (1/ 85)

'2. .The. boundary. layer.on.the drops is-assumed to be fully e

developed.- In reality'there is no boundary layer when the-. drop .is generated and, _therefore, .the heat-transfer..is much higher during the development.of.the boundary layer.

^

.3.- The drop'11 assumed to act as a rigid body,.so that the only heat transport mechanism inside the drop is conduction. Any internal circulation would speed up-L heat transfer inside the drop.

4. The temperature .of the containment ~ atmosphere- is assumed constant during a drop lifetime. :This is very

.nearly so.

5.- The containment atmosphere is assumed to be saturated.

This assumption _.is conservative since any cooling effect by spray evaporation is eliminated.

c. Dusinberre Numerical Method ~

The Dusinberre method fl6I is used for transient heat flow calculations for both '. the RCS hot metal and . the static . sj nks. -

The' method calculates heat flow as a function of . material density, specific heat, conductivity, . slab thickness and time nterval.

Effects on' Compartments-The computer' code CUPAT is used for determining the design pressures for the . interior compartments or rooms such as the reactor cavity, steam generator, and pressurizer cubicles. This q code is derived from the computer code LOCTICrisi used for the primary containment analyses. In order to calculate the pressure transient within a compartment, ~it numerically solves finite difference equations defining heat and mass flows into and out of the compartme..t. The code is a mathematical description of the compartment and calculates 'the effects of reactor coolant discharging into the compartment and of flow from the compartment to the. main-volume of the containment atmosphere. The mass and enthalpy flow rates from a double-ended rupture of the primary coolant pipe into the compartment are obtained from the LOCTIC code. The LOCTVS vent flow model I171 is used to calculate flow out of the compartment.

The LOCTVS vent flow model assumes homogeneous flow (no slip between the gaseous and liquid phases) and is based upon compressible gas theory net extended to include the effect of the noncompressible (liquid) phase. The behavior of the mixture can be conveniently described by considering ic as an ideal gas wien appropriately calculated pseudo properties /191 Blowdown Model Comparison The computer code CUPAT was used to calculate the peak pressure 14.3-32 1

l

I l

_ _ _ _ _ _ - . _ _ __ _ _ _ _ _ _ __ _ _ - - - - - - - - -- _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _. I

'17 7 01.1 U- unu) 2%

~

1 Pav. 3'(1/55)

BVPS-1-UPDATED FSAR .;

resulting frem a LOCA in the containment interior compartments, j and that the mass and energy release rates used in CUPAT were 1 obtained from the computer code LOCTIC. To assure the l applicability of the LOCTIC blowdown model, the CUPAT results are compared to the SATAN VI blowdown code for the~ double-ended het leg break.- The mass and energy. release rates from a double ended i hot- leg break as computed by LOCTIC and SATAN VI are shown in Figures 14.3-77A and B and 14.3-78A and B.

In 'the analysis of- containment- subcompartments, localized pressure buildup effects are considered. All closed or restricted compartments subjected to these localized pressure buildups are designed with sufficient vent openings to limit the pressure differentials between adjacent compartments. All structural components, walls, floors and beams in these compartments are designed to withstand the pressure differential as part of their design loading to protect the interior of the

. primary containment. -

f( 14.3.4.3 Pressure Transient Results i'

Figure 14.3-55 illustrates the containment pressure transients for a 29 inch double-ended rupture (DER) with normal engineered safety features in operation as follows:

1. Two charging pumps become effective 25.0 seconds after i

- the start of the incident.

2. Two low head safety injection pumps become effective 25.0 seconds after the start of the incident.

3. Three gas pressurized accumulators begin to inject water into the RCS when the reactor vessel pressure A falls below 600 psia.

4. Two quench spray pumps become effective 60 seconds after receipt of the containment isolation phase B signal. The containment isolation phase B signal is initiated when the containment pressure reaches 10 psig.

5. Four recirculation pumps become effective 5 minutes after receipt of the containment isolation phase B signal. This delay time assures sufficient water inventory in the sump for proper pump operation.

For the 29 inch double-ended hot leg rupture with normal engineered safety features, RCS blowdown is essentially completed in 16 seconds, at which time the containment pressure has reached an initial peak of about 38 psig at 268 F. For break sizes other than the 29 inch hot leg DER, blowdown is extended, with more heat absorbed by the steel and concrete in the containment, or by the engineered safety features. The DER results in a containment 14.3-33 M_ m.___.__2_2 .___ _-_ __.__._m2.

sy 17781. 10 e US(Bf 256 p..

INSERT E L- Containment The reactor containment is maintained at a subatmospheric pressure during reactor. operation, during which time the air partial pressure varies between I

approximately 9 and 10.5 psia. The allowable variation is based upon the cooldown capability of the containment depressurization system considering seasonal temperature changes in the ser.vice water. The analysis-indicates that the containment pressure will not rise above the design pressure of 45 psig following a LOCA inside containment. The containment returns to subatmospheric pressure within 60 minutes from the occurrence of the accident, thus terminating any outleakage from the containment.

A spectrum of RCS break sizes and locations is evaluated in the containment LOCA analysis which is based on an average reactor coolant temperature of 565.2*F and a core power of 2713 MWt.

A summary of the .results of the peak pressure analysis is presented in Table 14.3-16. A full ' double-ended rupture (DER) of a hot-leg, pump suction, and pump discharge, and smaller breaks at the pump suction are analyzed.

The calculated peak pressure following a LOCA is maximized by the maximum initial containment total pressure and temperature The initial conditions for the peak prer.1 r3 analysis are listed in Table 14.3-17.

20732E-1778110-B4 1 l

< ' [ 7)b 1' U hSERT ECo 'd)-

I'. b Pump-suction . breaks yield the highest . erergy flow to the containment during.'

the ' post-bicwdown period. For. the pump-discharge break, all the ' fluid j

' leaving the top of the core passes tnrough the steam generators and may become.superheated. However, the core flooding rate (and.therefore the rate of fluid leaving the core) is limited to a relatively ' low value by. the resistance of the pump 'in the broken.' loop. For a hot-leg. break, the core flooding rate is not so restricted and the majority.of the fluid leaving the top of the core bypasses the steam generators and is not superheated. Thus,

' the steam generators add- little or no energy to - the containment for . a hot-leg break. The pump-suction break, on the other hand, has a relatively _l high core flooding ' rate combined with all the ' fluid passing through the

, primary side of the steam generators.

The containment pressure transients ' for a DER at the. three break : locations are shown in Figure 14.3-55. ' Failure of one diesel generator is assumed for:

- this an'alysis. A single failure analysis is not required since the peak pressure for all breaks occurs prior to the initiation of the quench spray and recirculation spray. Pressure transients are plotted for the. pump suction, hot-leg and pump-discharge DERs.

Yiver The peak containment pressure does not depend directly on the seev+ee water temperature since the peak pressure occurs prior to the time the recircula-i tion spray subsystem becomes effective. However, the peak containment pressare is maximized by a high initial containment pressure, which .is riVtf .

permissible only at low serv +ee water temperatures. l 20732E-1778110-B4 2 u __ - __-__

7778L !10TNSERT Us 81 E (Co(nt' l%d) 'I Pressure transients for three different size pump-suction. ruptures are plotted in Figure 14.3-56.

As noted in Table 14.3-16, the break location and size giving the highest containment pressure is a pump discharge double-ended ~. rupture (PDDER).

The initial conditions for the containment depressurization analysis are also listed in Table 14.3-17. As discussed earlier, during operation the maximum permissible containment air partial pressure varies as a thw function of the n.-; ice water temperature according to the technical Specifications. This is required to ensure that the containment de-pressurizes in less than 60 minutes and maintains subatmospneric pressure thereafter. The pump suction DER is the. limiting break for depressurization.

The limiting single failure is the failure of one emergency diesel generator to start. This minimizes the energy removal from the con-tainment required for depressurization of the containment. The

-)'

containment pressure transients for this failure are shown in Figure 14.3-56 for the three pump suction break sizes analyzed. The limiting depressurization time is 3520 sec for the pump suction DER.

A chronology of events for this accident is presented in Table 14.3-11.

The containment atmosphere and sump temperature transients are shown in Figure 14.3-58. i l

)

20732E-1778110-B4 3 l

l

~

J. / 7 81. 1 U US(B) 256-BVPS-1-UPDATED FSAR Rev. 3 (1/85) {

i pressure, which is less than containment design pressure of 45 psig by an adequate margin. j r Figure 14.3-56 shows the containment pressure transients resulting from a 29 inch DER or a surge line rupture with the operation of minimum engineered safety features. Minimum j engineered safety features are defined as those engineered safety l features which operate during loss of all outside electrical power and with only one of the two emergency diesel generators in operation.

These minimum engineered safety features are as follows:

3

1. One charging pump tecoming effective 25 seconds after the start of the incident.

2. One low head safety injection pump bccoming effective 25 seconds after the etart of the incident.

3. Three gas pressurized accumulators discharging into the RCS when the reactor vessel pressure falls below 600 i A ;

psia. l

4. One quench spray pump becoming effective 60 seconds )

after receipt of the containment isolation phase B- {

signal. The containment isolation phase B signal !

occurs when the containment pressure reaches 10 psig.

5. Two recirculation pumps becoming effective 5 minutes after receipt of the containment isolation phase B signal.

At 5 minutes after receipt of the containment isolation phase B signal, there is adequate water on the floor for the recirculation spray pumps to operate. The sump water level for an 11.2 inch surge line breek and a 29 inch DER of the hot leg are shown in Table 14.3-10 for various times following a DBA. It should be noted that the recirculation pumps start later for a surge line break than for a 29 inch DER of a hot leg.

An accident chronology, which shows the approximate times of initiation of the various engineered safety features following a DBA, is given in Table 14.3-11. This table covers both minimum and normal engineered safety features during summer conditions.

The pressure transient for a period of one day following a 29 inch hot leg DER is shown in Figure 14.3-57 for minimum engineered safety features during summer conditions.

The corresponding sump water and containment atmosphere temperature transients are shown in Figure 14.3-58.

14.3-34 i

g.

US(B) p, 177 81

I QPS-1-UPDATED B FSAR Rev.

256 3 (1/85) q l The recirculation spray heat exchanger heat removal rate f . ene day following a ' hot leg DER is shown in Figure 14.3 9 for minimum engineered safety features during summer conditi .s.

The heat inventory in the co'ntainment prior to a DBA s shown in Table '14.3-4. .. This table also lists the heat .ventory at 17 seconds, which is,the approximate time of peak p ssure and at 88,860 seconds.

Pressure transients for DER's of the 27.5 ir h cold leg and 31 inch reactor coolant pump suction line w re performed for winter and summer conditions and for n mal and minimum engineered safeguards. Figure 14.3-55A ows a containment pressure transient resulting from a 27.5 inch DER for winter operating conditions with normal engineere safeguards except for failure of one quench spray pump. Th containment pressure transient for the cold leg or reactor c lant pump suction line DER differs from that of the hot leg DE in the following ways:

1. The peak. containment press e is slightly lower for a cold leg or pump suction R.

2. The peak pressure occurs approximately 2 seconds later for a cold leg or pump uction DER.

3. The containment pre ure transient shows a second pressure peak for a old leg or pump suction DER.

-4. The containment de pressurization time is longer for a cold leg or pump uction DER than for a hot leg DER.

The reasons for these dif rences are'as follows:

1. For a cold eg or pump suction DER, some of the accumulator ater is lost through the break resulting in less qu ching action in the reactor core.

2. For a e d leg or pump suction DER, the effective dischar e area of the break is smaller, resulting in a longer reactor vessel blowdown time.

3. Foll ing a cold leg or pump suction DER, but not fo owing a hot leg DER or surge line DER, a second p ssure break occurs at the end of core reflooding cause the core effluent during reflooding must pass hrough the steam generators to reach the break. The .!

steam generator tubes have sufficient surface area to )

boil the liquid pcrtion of the core effluent and to superheat all of the effluent to the temperature of the steam generator secondary water. This results in steam being injected into the containment atmosphere rather than hot water spilling onto the containment

' floor as for a hot leg DER or surge line DER.

14.3-35 1

1

1 7 7 0 1'. 1U. . US (ti) 256 l: BVPS-1-UPDATED FSAR Rav. 3 (1/85) 1 1

4. Depressurization time for a cold leg or pump suction DER is longer than for a hot leg DER since there is more heat to be removed from the containment atmosphere for the former.

Containment pressure transients for a DER of the reactor coolant < d pump suction line are shown -in Figures 14.3-80 and 14.3-81.

Figure 14.3-80 represents the case for a DER during winter operation with normal safeguards, but with failure of one quench spray pump. This case produces the highest peak pressure (33 psig) for this break location. Figure 14.3-81 represents the case for a DER during summer operation with maximum ECCS, but with minimum quench and recirculation (failure of one CIB This case produces the longest depressurization time

( signal).

(approximately 55 m4 m :::

l l 14.3.4.4 Post-DBA Hydrogen Generation The post-DBA hydrogen control system, described in Section 6.5, consists of redundant hydrogen reccmbiners and a ' backup purge system and is designed to maintain the hydrogen concentration in the containment atmosphere below the 4 percent lower flammability' limit, following a LCCA. The main control of hydrogen during the first 20 to 30 days following a LOCA can be accomplished with one j of the two recombiners or with the backup purge system. After about 30 days, control of hydrogen concentration can be effectively accomplished with either the recombiner or the purge ,

system operating at a reduced rate of approximately 16 scfm.

The five major sources of hydrogen generation accounted for after a DBA are as follows: ,

1. Zirconium - water reaction !

l

2. Pressurizer gas space and RCS water

3. Radiolytic generation in the containment sump

4. Radiolytic generation in the reactor core

5. Corrosion.

Table 14.3-9 gives the parameters used in calculating the above hydrogen sources.

Figure 14.3-60 gives the hydrogen concentration following a DBA as a function of time with no recombiners, with one recombiner j (minimum engineered safety features) and with two recombiners.

No purging is assumed. The recombiners are actuated two days after a DBA.

Convective mixing associated with containment spray system operation ensures a uniform mixture of hydrogen within the containment. The BVPS-1 containment is similar to the Surry 14.3-36

= ---

1 j .

l Rnferences to Section 14.3 (Cont'd)

74. Doses to the BV-1 Control Room Due to a LOCA at BV-1, Calculation Identification Number 14110.39, UR(B)-457 dated 5-11-87.

References do not appear in text. These references are a part of the BVPS-1 ECCS Reanalysis, Westinghouse Letter DLW-79-530, dated April 25. 1979, and are provided for information purposes.

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BVPS-1-UPDATED FSAR Rav. 0 (1/82) j 1

ll TABLE 14.3-10 ../

/

SUMP WATER LEVELS FOLLOWING A DBA /

/

Level (ft) f Time (seconds) Surge Line Break fli 29 inch DER [of Hot Leg (2) t 180 10.0 ,7 10.1

./

360 10.3 jf 10.3 560 10.6 j 10.6

/

750 10.8 [ 10.9

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780 10.9 jf' 10.9 1,100 11.3 / 11.4

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12. ]p/ 12.8 6,610 14/2 14.3 1

9,130 I4.5 14.6

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l' (11 Reci ulation pumps start at 322 seconds (2) Rec culation pumps start at 303 seconds 1 of 1

17701. S-1-UPD US FSAR 256 R av. 0 (1/82)

[Acc,ja,t O rweleg y, L fa g [ucf. k Dt'A - b '"' f ' ABLE 14,3-11 (s; ..e Derm u< <>4 ~

Tzr-e s-str F SEQUENCE OF EVENT 5')

L The following sequence of events occurs following a LOCA (DBA):

pproximateTTime Jee Event 0 DBA occurs 1Q Containment high-pressure setpoint reached Containment isolation Phase A actuation signal to low-head safety injection, high head safety injection 3(econd) Containment high-high pressure setpoint reached Containment isolation Phase B actuation signals to quench and recirculation sprays 1 Csecond@ Gas accumulators begin to inject to core 1$ (secondD Containment peak pressure reached' 25 (secondB Core reflooding begins 30 Q iecond_s) Low-head safety injection and high-head injection (seconds] Accumulators empty (fdc"cn'db eo Quench spray initiation reStoo,C.,. is co m le-Je

@p,g(secondg c- Core is filled'; containment second]

essure peak -

C 30@ Ge'i: ends) Recirculation spray initiation 20 minutes OperatorTna.niaces steam generator dump 7:o atmosphere _

no ucn Qc ,,.

QO -i-"t9 (Eutomatic) initiation of recirculation mode low-head safety injection 24 to (TOO minuteg End of quench spray

[ hours nd of safety injection from refuelin water storage tank

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_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ J

BVPS-1-UPDATED FSAR Rsv. 0 (1/82)

TABLE 14.3-11 (CONT'D)

STEP-BY-STEP SEQUENCE EVENTS Approximate Time

/ Event 8 days operatr8r starts recombiners and initiates ther' hydrogen control system d

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