Text

Proposed Tech Specs Re Containment Airlocks & Pressure,Air Temp,Containment Structural Integrity & Containment Ventilation Sys
	ML20012B151
Person / Time
Site:	Millstone
Issue date:	02/26/1990
From:	; NORTHEAST NUCLEAR ENERGY CO.
To:	;
Shared Package
ML20012B148	List: ML20012B147; ML20012B151;
References
	B13429, NUDOCS 9003130672
	Download: ML20012B151 (24)
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, Proposed Revision to Technical- Specifications L Containment Pressure t ,

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INDEX LIMITING CONDITIONS FOR OPERATION AND SVRVEILLANCE RE0VIREMENTS SECTION P_AE FIGURE 3.4-1 DOSE EQUIVALENT 1131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY

1*Ci/ gram ,

DOSE EQUIVALENT 1131.................................... 3/44-30 TABLE 4.4-4 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PR0 GRAM.................................................. 3/44-31 3/4.4.9 PRESSURE / TEMPERATURE LIMITS ;

Re actor Cool ant System. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-33 FIGURE'3.4-2 REACTOR COOLANT SYSTEM HEATVP LIMITATIONS ~-

P APPLICABLE UP TO 10 EFPY................................. 3/4 4 34 FIGURE 3.4-3 REACTOR COOLANT SYSTEM C00LDOWN LIMITATIONS -

APPLICABLE UP T0 10 EFPY................................. 3/4 4-35 TABLE 4.4-5 REACTOR VESSEL MATERIAL SVRVEILLANCE PROGRAM -

WITHDRAWAL SCHE 0VLE...................................... 3/4 4-36 Pressurizer.............................................. 3/4 4 Overpressure Protection Systems.......................... 3/4 4-38 l

L FIGURE 3.4-4a NOMINAL MAXIMUM ALLOWABLE PORV SETPOINT FOR THE COLD OVERPRESSURE SYSTEM (F0UR LOOP OPERATION)................ 3/4 4-40 FIGURE 3.4-4b NOMINAL MAXIMUM ALLOWABLE PORY SETPOINT FOR THE COLD l

OVERPRESSURE SYSTEM (THREE LOOP OPERATION)............... 3/4 4-41 3/4.4.10 STRUCTURAL INTEGRITY..................................... 3/4 4-42 L 3/4.4.11 REACTOR COOLANT SYSTEM VENTS............................. 3/4 4-43' l

H 3/4.5 ' EMERGENCY CORE COOLING SYSTEMS

'3/4.5.1 ACCUMULATORS............................................. 3/4 5-1 1 3/4.5.2- ECCS SUBSYSTEMS:- T avg GREATER THAN OR EQUAL TO 350.... 3/4 E-3 3/4. 5.~ 3 ECCS SUBSYSTEMS - T,yg LESS THAN 350................... 3/4 5-7 3/4.5.4 REFUELING WATER STORAGE TANK............................. 3/4 5-9 t 3/4.6 CONTAINMENT SYSTEMS L 3/4.6.1 PRIMARY CONTAINMENT.

L Containment Integrity.................................... 3/4 6-1 Containment Leakage...................................... 3/4 6-2 TABLE 3.6-1 ENCLOSURE BUILDING BYPASS LEAKAGE PATHS............... 3/4 6-4 Containment Air Locks.................................... 3/4 6-5 Containment Pressure..................................... 3/4 6-7 i

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, INDEX 1 i LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS l SECTION A E,A_[i[

Air Temperature................................... 3/4 6-9 ;

Containment Structural Integrity.................. 3/4 6 Containment Ventilation System.................... 3/4 6-11 1<

3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS

l. Containment Quench. Spray System.................... 3/4 6-12 1

!L Recircul ation Spray System. . . . . . . . . . . . . . . . . . . . . . . . . 3/4 6-13 Spray Additive System.............................. 3/4 6-14 l L 3/4.6.3' CONTAINMENT ISOLATION VALVES....................... 3/4 6-15 1 .

n f 3/4.6.4 COMBUSTIBLE GAS CONTROL

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Hydrogen Monitors..................................- 3/4 6-35

l. Electric Hydrogen Recombiners...................... 3/4 6-36 1-FIGURE 3.6-2 HYDR 0 GEN RECOMBINER ACCEPTANCE CRITERIA FLOW VS.

CONTAINMENT PRESSURE...............................

3/46-36a 3/4.6.5 SUBATMOSPHERIC PRESSURE CONTROL SYSTEM Steam Jet Air Ejector.............................. 3/4 6-37 3/4.6.6 SECONDARY CONTAINMENT l

L Supplementary Leak Collection and Release System... 3/4 6-38 L

Enclosure Building Integrity. . . . . . . . . . . . . . . . . . . . . . . 3/4 6-40 Enclosure Building Structural Integrity............ 3/4 6-41 3/4.7 PLANT SYSTEMS L '3/4.7.1 TURBINE CYCLE p Safety Valves........................................... 3/4 7-1 1

TABLE 3.7-1 MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH IN0PERABLE STEAM LINE SAFETY VALVES l-DURING FOUR LOOP 0PERATION......................... 3/4 7-2 TABLE 3.7-2 MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH IN0PERABLE STEAM LINE SAFETY VALVES THREE LOOP 0PERATION.............................. 3/4 7-2 MILLSTONE - UNIT 3 ix l

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3- - 3/4.'6 CONTAINMENT SYSTEMS ,

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, PRIMARY CONTAINMENT. j CONTAINMENT INTEGRITY

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' LIMITING CONDITION FOR OPERATION ,

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-3.6.1.1_ Primary CONTAINMENT INTEGRITY shall be maintained, o APPLICABILITY: MODES 1, 2, 3, and 4. -

ACTION:

Without primary CONTAINMENT. INTEGRITY, restore CONTAINMENT = INTEGRITY within-I 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 .

SURVEllLANCE RE0VIREMENTS 4.6.1.1 Primary CONTAINMENT INTEGRITY shall be demonstrated:

a. At least once per 31 days by verifying that all penetrations

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capable of being closed by OPERABLE containment automatic isolation ,

valves or operator action during periods when containment isolation valves are opened under administrative control,** and required to be.

closed during- accident conditions are closed by valves, blind flanges, or deactivated automatic valves secured in their positions.

b. By verifying that each containment air lock is in compliance with the requirements of Specification 3.6.1.3; and 1

c. After. each -closing- of -each penetration subject to Type .B testing, except the containment air locks, if opened following a Type A or B test, by leak rate testing the seal with gas- at a pressure not less than P 53.27 psia (38.57' psig), and verifying that when the measureh, leakage rate for these seals is added to the leakage rates determined pursuant to Specification 4.6.1.2d. for all other. Type B and C penetrations, the combined leakage rate is less than 0.60 L '

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Except valves, blind flanges, and deactivated automatic valves which are located inside the containment and are' locked, sealed, or otherwise secured in the closed - position. These penetrations shall be verified closed during each COLD SHUTDOWN except that such verification need not be performed more often than once per 92 days.

- The following manual valves may be opened on an intermittent basis under administrative- control. 3FPW-V661, 3FPW-666, 3 SSP-V13, 3 SSP-V14, 3HCS-V2, 3HCS-V3, 3HCS-V9, 3HCS-V10, 3HCS-V6, 3HCS-V13, 3SAS-V875, 3SAS-V50, 3CHS-V371, 3CCP-V886, 3CCP-V887, 3CVS-V13.

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) CONTAINMENT SYSTEMS

.v - CONTAINMENT LEAKAGE LIMITING CONDITION FOR OPERATION 3.6.1.2 ~ Containment leakage rates shall be limited to:

a. An overall integrated leakage rate of less than or equal to L a, 0.65% by weight of the containment. ai_r per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at P , '

53.27 psia (38.57 psig);

b. A' combined leakage rate of less than 0.60 L for all penetrations and valves subject to Type B and C tests, when pressurized to P i and- a- t s

c. A- combined leakage rate of less than or equal to 0.042 L for all penetrations identified in Table 3.6-1 as Enclosure Buildilig bypass -

leakage paths when pressurized to P

a APPLICABILITY: MODES 1, 2, 3, and 4. I ACTION: 1 With the measured overall integrated containment leakage rate exceeding 0.75-L , or_ the measured combined leakage rate for all penetrations and valves s0bject to Type B and C tests exceeding 0.60 L , or the combined bypass leakage rate exceeding 0.042 L , restore the overill integrated leakage rate -l to less than 0.75 L the combined leakage rate for all penetrations subject

.to Type B and C test.,s to less than 0.60 L , and the combined bypass leakage ;

rate to less than 0,042 L, prior to incdasing the Reactor Coolant System temperature above 200*F.

SURVEILLANCE RE0VIREMENTS 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 Part 50 using methods and provisions of ANSI N45.4-1972-(Total Time Method) and/or ANSI /ANS 56.8-1981 (Mass Point Method):

a. Three Type A tests (Overall Integrated Containment Leakage Rate) ,

shall be conducted at 40 10 month intervals during shutdown at a pressere not less than P , 53.27 psia (38.57 psig) during each 10-year service period, the third test of each set shall be conducted during the shutdown for the 10-year plant inservice inspection;

b. If any periodic Type A test fails to meet 0.75 L the test schedule for subsequent Type A tests shall be reviewed $n,d approved by the Commission. If two consecutive Type A tests fail to meet 0.75 L a Type A test shall be performed at least every 18 months until a,wot consecutive Type A tests meet 0.75 aL at which time the above test schedule may be resumed; MILLSTONE - UNIT 3 3/4 6-2

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l/L T CONTAINMENT SYSTEMS SURVEILLANCE REOUIREMENTS (Continued) s e c. The accuracy of each lype A test shall be verified by a supplemental test which:-

1) Confirms the accuracy of the test by verifying that-the supple-mental test results, L minus the sum ~of the Type A and the

, superimposed leak, oL , i,s equal _to or less than 0.25 L,; .

2) Has a duration sufficient to establish accurately the change in '

leakage rate between the Type A test and the; supplemental test; and 3)~ Requires that the rate 'at which gas is injected into the containment or- bled from- the- containment during the supplemental test is between 0.75 L, and 1.25 La '

d. Type B and C tests shall be conducted with gas at P , 53.27 psia (38.57 psig), at intervals no greater than 24 montf?s except ' for i tests involving: i

1) Air . locks

e. The combined bypass leakage rate shall be determined to be less than or equal _to 0.042 L by' applicable Type.B and C tests at least' once per 24 months excellt for penetrations which are not individually testable; penetrations not individually testable shall be determined to have no detectable leakage when tested with soap bubbles while the containment is - pressurized to P,, 53.27 .psig -(38.57 psig), during each Type A test;

f. Air locks shall be tested and demonstrated OPERABLE by the j requirements of Specification 4.6.1.3;

. g. Purge supply and exhaust isolation valves -shall be demonstrated j j

OPERABLE by the requirements of Specifications 4.6.3.2.c and 4.9.9.

h. ' The provisions of Specification 4.0.2 are not applicable.

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TABLE 3.6-1

[RC10SURE BUILDING BYPASS LEAKAGE PATHS PENETRATION DESCRIPTION RELEASE LOCATION ,

14 N2 to Safety Injection Tanks Ground Release c 15 Primary Water to Pressurizer Ground Release Relief Tanks 35 Vacuum Pump Suction Plant Vent 36 Vacuum Pump Suction Plant-Vent ,

37 Air Ejector Suction Plant Vent 38 Chilled Water Supply Plant Vent 45 Chilled Water Return Plant Vent 52 Service Air Turbine Building Roof-Exhaust 54 Instrument Air Turbine Building Roof Exhaust j 56 Fire Protection Ground Release 59 Fuel Pool Purification Ground ~ Release

60 Fuel Pool Purification Ground Release 70 Demineralized Water Ground Release

~72 Chilled Water Supply Plant Vent ,

85 Containment Purge Ground Release 86' Containment Purge Plant Vent 116 Chilled Water Return Plant Vent 124 _ Nitrogen to Containment Plant Vent L ,

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1 Y ' CONTAINMENT SYSTEMS CONTAINMENT AIR LOCKS LIMITING CONDITION FOR OPERATION 3.6.1.3 The containment air lock shall be OPERABLE with:

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

b. An overall air' lock leakage rate of less.than or equal to 0.05 La at ,

P,, 53.27 psia (38.57 psig).

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

1 ACTION:-

a.- With one containment air lock docr inoperable:

1. Maintain at-least the OPERABLE air lock door closed

and either restore the inoperable' air lock door to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or lock the OPERABLE air. lock door closed,

2. ' Operation may then continue until performance of the next

, required overall air lock leakage test provided that the OPERABLE air lock door is verified to be locked closed at least "

once per 31 days,

3. Otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 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 , and

4. The provisions of Specification 3.0.4 are'not applicable. ;

b. -With the containment air lock inoperable,-except as the result of an '

inoperable air lock door, maintain at 'least one air lock door closed; restore the inoperable air lock to. OPERABLE status within L 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 SHUT 00WN'within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

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*Except during entry to repair an inoperable inner door, for a cumulative I time not to exceed I hour per year.

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4.6.1.3 Each containment air lock shall be demonstrated OPERABLE: '

a. .1) Within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .following each closing, except when the air lock is being used for multiple entries, then at least once perc 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , by verifying no detectable seal leakage by. pressure decay when the volume between the door seals is pressurized to greater than or equal to P,, 53.27 psia- (38.57 psig), for at

least 15 minutes; or

2) Within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months following each closing, 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 that the seal leakage is less than 0.01 L as determined by precision flow measurements when -measured I f8r at least 30 seconds with the volume between the seals at a constant pressure of greater than or equal to Pa , 53.27 psia-

.(38.57 psig);

or

3) Within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months following each closing, except when the air lock is being used for multiple entries, then at least once per ;

72- hours, by completing an overall air lock leakage test per ,.

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By 53.27conducting psia (38.57overall psig),air andlock leakage verifying thetests at not overall less leakat air lock than P]e, rate is within its limit: i

1) At least once per 6 months,* and j

2) Prior to establishing- CONTAINMENT INTEGRITY when maintenance has been performed on the air lock that could affect the air. jl lock; sealing capability.**

c. At least once per 6 months- by verifying that only one door in each 1 air lock can be opened at a time. !

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The provisions of Specification 4.0.2 are not applicable. 3

- This represents an exemption to Appendix J, paragraph III.D.2.(b)(ii), of l 10 CFR Part 50.

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I, CONTAINMENT SYSTEMS:

. CONTAINMENT PRESSURE A

LIMITING CONDITION FOR OPERATION 3.6.1.4 Primary containment pressure shall be maintained greater than or equal to 10.6 psia and less than or equal to 14.0 psia.

APPLICABillTY: MODES 1, 2, 3, and 4.

ACTION:

With the containment pressure less than 10.6 psia or~ greater-than 14.0 psia, restore the containment pressure to within the limits 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 RE0VIREMENTS 4.6.1.4 The primary containment 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 . <

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3/4.6' CONTAINMENT SYSTEMS !

BASES 3/4.6.1 PRIMARY CONTAINMENT 3/4.6.1.1 CONTAINMENT INTEGRITY Primary CONTAINMENT INTEGRITY ensures that the release of radioactive materials from the containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the safety analyses. This restric- ,

tion, in conjunction with the leakage rate limitation, will limit the SITE BOUNDARY radiation doses to within the dose guidelines of 10 CFR Part 100 during accident conditions and the control room operators dose to.within the guidelines of GDC 19, 3/4.6.1.2 CONTAINMENT LEAKAGE The limitations on containment leakage rates ensure that the total containment leakage volume will not exceed the value assumed in the safety 7

analyses at the peak accident pressure, P,. As an added conservatism, the measured overall integrated . leakage rate is further limited to less than or equal to 0.75 L, during performance of the periodic test to account for ;

possible degradation of the containment leakage barriers between leakage tests.

The surveillance testing for measuring leakage rates are consistent with I

the requirements of Appendix J of 10 CFR Part 50.

l 3/4.6.1.3 CONTAINMENT AIR LOCKS The limitations on closure and leak rate for the containment air locks

(- are: required to meet the restrictions on CONTAINMENT INTEGRITY and containment

leak rate. Surveillance testing of the air lock seals provides assurance that l the overall air lock leakage will not become excessive due to seal damage l during the. intervals between air lock leakage tests.

3/4.6.1.4 and 3/4.6.1.5 AIR PRESSURE and AIR TEMPERATURE The limitations on containment pressure and average air temperature ensure that: (1) the containment structure is prevented from exceeding its design negative pressure of 8 psia, and (2) the containment peak pressure does not exceed the design pressure of 60 psia during LOCA conditions. Measure-ments shall be made at all listed locations, whether by fixed or portable

-instruments, prior to determining the average air temperature. The limits on the pressure and average air temperature are consistent with the assumptions of the safety analysis. The minimum total containment pressure of 10.6 psia is determined by summing the minimum permissible air partial pressure of 8.9 psia and the maximum expected vapor pressure of 1.7 psia (occurring at the maximum permissible containment initial temperature of 120*F).

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C CONTAINMENT SYSTEMS, l BASES I

3/4.6.1.6 CONTAINMENT STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the containment will be maintained comparable to the original design standards for the life of the facility. Structural integrity is required to ensure that the containment will withstand the maximum pressure of 60 psia in the event of a- LOCA. A visual inspection in conjunction with the Type A leakage tests is sufficient to demonstrate this capability.

3/4.6.1.7 CONTAINMENT VENTILATION SYSTEM The 42-inch containment purge supply and exhaust isolation valves are required to be locked closed during plant operation since these valves have I not been demonstrated capable of closing during a LOCA or steam line break i accident. Maintaining these valves closed during plant operations ensures that excessive quantities of radioactive materials will not be released via the Containment Purge System. To r,rovide assurance that these containment valves cannot be inadvertently opened, the valves are locked closed in accordance with Standard Review Plan 6.2.4 which includes mechanical devices to seal or lock the valve closed, or prevents power from being supplied to the valve operator.

The Type C testing frequency required by 4.6.1.2d is acceptable, provided that the resilient seats of these valves are replaced every other refueling outage. -

E 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS 3/4.6.2.1 and 3/4.6.2.2 CONTAINMENT 00ENCH SPRAY SYSTEM and RECIRCULATION SPRAY SYSTEM- .

The OPERABILITY of the Containment Spray Systems ensures that containment t 1

depressurization and iodine removal will occur in the event of a LOCA. The '

l- pressure reduction, iodine ~ removal capabilities and resultant containment l leakage are consistent with the assumptions used in the safety analyses.

3/4.6.2.3 SPRAY ADDITIVE SYSTEM L The OPERABILITY of the Spray Additive System ensures that sufficient Na0H 1:

is added to the containment spray in the event of a LOCA. The limits on Na0H volume and concentration ensure a pH value of between 7.0 and 7.35 for the solution recirculated within containment after a LOCA. This pH band minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components. The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical charac-teristics.

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i Attachment 3 :

' Millstone-Nuclear Power Station, Unit No. 3 j Description of the Proposed Technical Specification ~.

Changes- and Significant Hazards Considerations Discussion 1?

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.U.S. Nuclear Regulatory Commission 1 B13429/ Attachment 3/Page1 l February 26, 1990

. Millstone Nuclear Power Station, Unit No. 3 J Description of the Proposed Technical Specification ,

Chanaes and Sionificant Hazards Consideration Discussion Discription of the ProDosed Technical Specification Chanaes Technical ' Specification 3/4.6 (Containment Systems) and associated bases are being changed to allow the containment pressure to increase to 14.0 psia

during Modes 1 through 4. The purpose of the containment pressure increase is to reduce the potential for personnel injury when entering containment due to crossing the pressure boundary and due to oxygen deficiency. The proposed containment pressure change is based on the results of a recent containment analysis performed by Stone and Webster under the direction of Northeast Nuclear Energy Company (NNECO).

The proposed changes to Technical Specification 3/4.6 affects the following:

1. The peak calculated containment pressure (P ) is changed to 53.27 psia (38.57 psig) in Sections 4.6.1.1.c, 3.6.1.%.a, 4.6.1.2.a, 4.6.1.2.d, 4.6.1. 2.e, 3. 6.1.3. b, 4.6.1.3. a.1 and a.2, 4.6.1.3. b. This is based on the results of a revised containment analysis.

2.

The from integrated 0.9 weight leak rateper percent at.da P), containment to 0.65 weightleak percent rate (L pef) day is changed in Sec-

-tion 3.6.1.2.a.

3. The combined bypass leakage rate is changed from 0.01 L, to 0.042 L a I" Sections 3.6.1.2 ACTION and 4.6.1.2.e.

l' 4. The operating containment pressure of 14.0 psia is specified in Sec-tion 3.6.1.4. In addition, the maximum and minimum limit for the con-tainment pressure is specified as total containment pressure instead of l air partial pressure.

L 5. Figure 3.6.1 is deleted as the containment pressure will be read directly L from the main control board indicators.

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6. Bases for Sections 3/4.6.1.1, 3/4.6.1.4, 3/4.6.1.5, 3/4.6.2.1, and-3/4.6.2.2 are revised to reflect the above changes.

L 7. Index of Technical Specifications has been revised to reflect the above i

changes.

L Containment bypass penetrations are lines that come out of the primary con-

'tainment and run through the enclosure building to areas outside the plant.

Leakage through the containment isolation vaives (CIV) in these penetrations could bypass the secondary containment afforded by the enclosure building and l

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l U.S. Nuclear Regulatory Commission ]

B13429/ Attachment 3/Page 2 >

February 26, 1990 l

go into: the environment during a design basis accident (DBA). The following changes to Table 3.6-1 represent the results of refinements in previous analyses which identified bypass penetrations. It will improve containment integrity by deleting testing of-penetrations that are not potential bypass paths and refocusing. this testing on penetrations that really do have the potential for being bypass leakage paths.

'The proposed changes also correct the bypass penetration listing of Technical Specification Table 3.6.1 as follows: 1

1. Penetrations Z-28 and Z-29 (aerated drains and gaseous vents) are being deleted.

2. Penetrations Z-59, 2-60, and Z-124 (fuel pool purification and nitrogen supply to containment) are being added.

3. Table 3.6.1 has been revised to include description for each penetration.

Sianificant Hazards Consideration In accordance with 10CFR50.92, NNECO has reviewed the proposed Technical Specification changes and has concluded that they do not-involve a significant-hazards consideration. The basis for this conclusion is that the three

, criteria ' of 10CFR50.92(c) are not compromised. The proposed changes do not I-

' involve a significant hazards consideration because the changes would not:

1. Involve a significant increase in the probability of occurrence or consequences of an accident previously analyzed. :

L a. The increase in containment pressure affects the following:

l (1) The temperature and pressure in the containment due to a spectrum of postulated loss-of-coolant accidents (LOCA),

control rod ejection accidents (CREA), and secondary system steam and feedwater line breaks.

(2) The external pressure to which the containment is subjected.

(3) The range and accuracy of instrumentation that is provided to monitor and record containment conditions during and following an accident.

(4) Containment heat removal system.

L (5) Minimum containment pressure analysis for emergency core i cooling system performance capability studies (LOCA).

(6) Subcompartment analysis.

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U;S. Nuclear Regulatory Commission

B13429/ Attachment 3/Page 3 x< February. 26, 1990 l (7)- Mass and energy release analysis for postulated LOCAs and secondary system pipe ruptures.

(8) Combustible gas concentration.

(9) Containment leakage-testing.

_ (10) Determination of leakage paths.

b. The increase in containment pressure impacts the consequences of the DBAs listed above as follows:

'(1)- The pressures in the containment increaa. The containment

f. pressure / temperature response was evaluated (see Attachment 1)-

E assuming a maximum operating pressure of 14.2 psia using the same methods.and models described in Section 6.2.1 of the Final L' Safety Analysis Report (FSAR). The maximum peak- containment pressure was recalculated to be 38.57- psig, which shows an increase from the current containment pressure peak of 1 36.09 psig'(References 1 and 2).- (Note: The current Technical '

Specification does not reflect the current analysis.) The containment long-term -depressurization transient was also l l recalculated-(see Attachment 1). The containment pressure does H not return to subatmospheric pressure, and leakage is assumed i to continue:.for 30 days (see Attachment 1). The current analysis discussed in References 1 and 2 shows the containment

!. - pressure returns to atmospheric pressure within I hour post- 1 LOCA, at which time the containment _ leakage paths are assumed l .to stop. _ To help compensate - for the _ increased release of L radioactivity, the allowable Technical Specification leak rate, L L , is being reduced from 0.9 percent per. day to 0.65 percent .

dr day. In spite of the reduction in allowable leak rate, j L lowever, some of the calculated dose consequences of the - LOCA i l

and CREA increase (see Attachment 1). _The new P of 38.57 psig is well below the design containment pressure of 45 psig. The containment pressure reduces to less than 50 percent of the-peak containment pressure within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after the postulated t accident (Standard Review Plan 6.2.1' A) .

. The calculated ,

l radiation doses for the exclusion area boundary (EAB), low- 1 population ' zone (LPZ), and operating personnel remain well l

within the 10CFR100 limits, the GDC 19 limits, and the Standard ,

f Review Plan acceptance criteria (see Attachment 1). '

L (2) The external pressure to which the containment is subjected following, for example, inadvertent operation of the contain-ment heat removal system is unchanged or in some cases de-12 creased.

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U.S. Nuclear Regulatory Commission -

813429/ Attachment 3/Page4 ;

February 26, 1990:

l L (3) The range .and accuracy of instrumentation that is provided to monitor and record containment conditions during and following <

an accident is not changed. Currently transmitters 3LM3*PT934, 935, 936, and 937, which have a range O to 60 psia, are used to '

, perform High 1, 2, and- 3 containment isolation. The' range of ,

this transmitter is too large .for ' . operations to maintain containment pressure within the proposed Technical Specifica-l tions. The two narrow-range transmitters 3LMS*PT43A/43B (8.5

? to 14.5 psia) that provide indication on the main control board will be utilized to set and maintain containment pressure to-the proposed Technical Specifications. The total probable-error of the reading during normal plant condition was deter- '

mined to be i .167 psi when using the plant process computers. 1 This error is incorporated in the proposed Technical Specifica- ,

.tions. Use of other methods of reading pressure will be '

readjusted for total error. The electrical- equipment qualifi- ,

y cation for 10CFR50.49 is not impacted by the - increase . in -

I containment pressure (see Attachment 1). For normal environ-L ment conditions, the EEQ program is based on a normal contain~-c ment pressure range of 9.5 to 14.7 psia, which bounds the l

proposed contrinment pressure of 14.0 psia. For accident l environment conditions, the - EEQ program is based ~ on- the pres-sure and tempenture envelope (Millstone Unit No. 3- FSAR Section 3.11), which bounds - the. calculated new pressures -and temperatures indicated in Attachment 1. For post-DBA environ-ment conditions from I hour to 1 year, the containment pressure value of 1.75 psig, although not bounded by the existing I; envelope included in the Millstone Unit No. 3 FSAR Sec-L tion 3.11, has no impact on the EEQ qualification because the.

L pressure is not an aging parameter which causes degradation of -

material. . The proposed change will not impact the existing L accident radiation qualification of EEQ equipment. Although-the proposed increase in containment pressure results in some increase in the radiation consequences following a DBA (Attach-p ,

ment 1), the equipment qualification remains valid with ade- "

quate margin.

(4) The proposed change has no effect on the containment heat removal system effectiveness. The . containment heat removal systems have been shown to be capable of reducing rapidly the containment pressure and temperature following a LOCA (Attach-ment 1).

L (5) The results of the minimum containment pressure analysis are .

! favorably impacted since higher containment pressures yield higher core flooding rates during LOCA and subsequent lower fuel peak cladding temperatures (Standard Review Plan 6.2.1.5).

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U.S.; Nuclear Regulatory Commission B13429/ Attachment 3/Page5-February,26, 1990 (6) The results of subcompartment analysis are favorably impacted since higher containment )ressures minimizes the resultant differential pressure (Stancard Review Plan 6.2.1.2) -

(7) The FSAR LOCA long-term mass and energy analysis, Section 6.2, will remain valid. .The release rate of the flow into the containment will change due to -the initial containment pres-sure, but not the total effluent available.

(8) :The proposed change has no effect on the current evaluation of hydrogen generation and control (see Attachment 1). An Lin-crease in the containment operating pressure causes an increase in the mass of air in the containment.- Because the- rate of generation of hydrogen is unchanged, the concentration of-hydrogen is lower.

(9) The containment leak testing will reflect tighter containment-leakage limits due to the increase' in operating containment-pressure. The containment leak rate (L 0.9 percent per day to 0.65 percent p8r) will.The day. be secondary reduced from containment bypass leakage from the containment will be increased from 0.01 L a to 0.042 L this is actually an increase fr% With the reduction in ~ L ,m 0.009 percent ' p 0.028 percent per day. As explained earlier in this section, one of the consequences of this change is that the containment-pressure does not return to subatmospheric pressure following a LOCA (see Attachment 1). The current analysis' on the subat-mospheric design of- Millstone Unit No. 3, however shows that' all containment leakage- terminates within I hour. The Techni-cal Specification leak rate, L is being reduced from 0.9 per-cent per day ' to 0.65 percent,per day to compensate for the increased- time in leakage release. The proposed changes in containment leakage meet the requirements of 10CFR50, Appen-dix J. However, it requires administrative revision of con-

.tainment Type B and C' leak testing procedures.

(10) Currently, any preexisting bypass leak in the Millstone Unit No. 3 containment resulting from human error, such as valves left inadvertently open, can be detected shortly upon initial isolation. The proposed Technical Specification change to increase the operating containment pressure could cause a leakage path to go undetected for a longer period of time. It has been concluded that a 3/4-inch line is the smallest line that could be bypassed. NNEC0 has determined that at the maximum containment pressure of 14.2 psia, it could take 6.41 hours4.74537e-4 days 0.0114 hours 6.779101e-5 weeks 1.56005e-5 months to detect a .1 psi change without the containment vacuum pumps operating. Since the instrument error on contain-ment pressure measurement is .167 psi, the time to detect a

.267 psi change would be about 17.12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . Currently, l

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b .

r, U.S. Nuclear Regulatory Commission - l B13429/ Attachment 3/Page 6 o February 26, 1990 Millstone Unit No. 3 has a 10 4 chance of. early containment bypass. It has been estimated that when a leak from a 3/4-inch line is being left open and assumed to go undetected for ,

5 days, the probability of early containment bypass remains in .l the order of 10 '.

In summary, the increase in normal operating containment pressure' - ,

may increase the duration of containment radiation leakage following' a LOCA. In spite of a proposed reduction in allowable leak rate, some of-the calculated' doses following LOCA and CREA increase. The ,

10CFR100 limits, GDC 19 limits, and Standard Review Plan acceptance criteria,.however, are still satisfied.

c. The changes in Table 3.6.1 are limited to changes in the designa-tions of containment bypass penetration. Bypass penetrations. are piping systems that come -out of the primary containment and run through; the enclosure building to areas outside of the plant.

Leakage through the containment valves in these. penetrations after a DBA could bypass the secondary containment afforded by the enclosure building. However, a change.to the bypass penetration listing does- 1 not constitute an increase in potential ~.off-site consequences due to a DBA. . The leakage limit is applied to all bypass CIVs regardless of their number- (i.e., the total bypass leakage limit is shared by all the CIVs). In addition, the refinement of the listing of the-penetrations that are potential bypass paths does not affect the probability of occurrence of a DBA.

l

2. Create the possibility of a new.or different kind of accident from any previously analyzed. The proposed increase in nor_ mal operating pressure is within the~ existing design conditions of the equipment. The proposed b changes would not impact the plant response to_ the point where a new

- accident' is created. No new failure modes are introduced by these

- proposed Technical Specification changes-that would allow the containment .

to remain at 14.0 psia during Modes 1 through 4 and that would refine the l listing of bypass penetrations. !

3. Involve a significant reduction in a margin of safety.

a. The proposed increase in operating containment pressure to 14.0 psia l does not impact the safety limits for the protective boundaries.

The calculated P is well below the containment design pressure of 45 psig. The co8tainment pressure reduces to less than 50 percent of. the peak containment pressure within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after the postu-lated accident, thus satisfying the Standard Review Plan Sec-tion 6.2.1A. The calculated radiation dose in the EAB, LPZ, and for operating personnel remain well within 10CFR100 limits, the General Design Criterion 19 limits, and the Standard Review Plan acceptance criteria for the postulated LOCA and CREA (see Attachment 1). Since L

safety limits are not impacted, the margin of safety between the L

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U.S. Nuclear Regulatory Commission B13429/ Attachment 3/Page 7 February'26, 1990-

- safety limits and protective boundary failure is not impacted. I Therefore, the proposed changes do not result in a reduction of any i safety margin.

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b. . - The' proposed changes to Table 3.6.1 are the result of refinements in l L

previous' analyses which identified- bypass penetrations. .Since the i . proposed changes do not impact -the safety limits, the proposed >

L changes do not result in a reduction of any margin of safety.

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U.S.- Nuclear Regulatory Commission B13429/ Attachment 3/Page 8-February 26, 1990' Refereneti (1) 1M111 stone 3 Safety Evaluation Report, NUREG-1031, Supplement No. 5, January 1986.

(2). Millstone Unit No. 3 FSAR,. Amendment 17.

L-l

l References Reference 1 NE-89-L-816, R. M. Kacich to Distribution, " Reporting Require-ments Regarding Noncompliance with NRC Regulations" (December 13,1989).

t Reference 2 Letter, E. J. Mroczka (Northeast Utilities) to W. T. Russell j (NRC Region I), B12863, " Report of Substantial Safety Hazard" (March 25, 1988). :j l Reference 3 Letter, Rosemount, Inc. (Steve Wanck) to Northeast Utilities l (Director of QA), " Notification Under 10 C.F.R. 21" :l 4 (February 7, 1989). 1 Reference 4 REF 89-10, Completed Reportability Ev11uation for Millstone i j

Unit 3 (February 22,1989).

! Reference 5 REF 89-09, Completed Reportability Evaluation .for Millstone l Unit 2 (March 10, 1989). 'l Reference 6 REF 89-08, Completed Reportability Evaluation for Millstone Unit 1 (March 14, 1989).

Reference 7 Letter, U.S. Nuclear E. J. Mroczka (Northeast Utilities) to Transmitters,"-

Regulatory Commission, B13178, "Rosemount j l

i (April 13, 1989). l Reference 8 NRC-Information Notice 89-42, " Failure of Rosemount Models 1153 and 1154 Transmitters," (April 21,1989).

Reference 9 NRC Inspection Report 50 423/89-04, dated June 28, 1969.

Reference 10 GSP-89-299, " Operability Determination for Rosemount Pressure I

! and Differential Pressure Transmitters at Millstone Unit 3" (July 31, 1989).

Reference 11 Letter, E. J. Mroczka (Northeast Utilities) to U.S. Nuclear Regulatory Commission, A08132, " Response to Inspec-tion 50-423/89-04" (August 1, 1989).

Reference 12 Letter, E. J. Mroczka (Northeast Utilities). to U.S. Nuclear Regulatory Regulatory Commission, B13366, "Rosemount Transmit-tors" (October 31,1989).,

Reference 13 Letter, D. L. Fuller -(Westinghouse) to A. R. Roby (Northeast Utilities), NEU-90 518, "Rosemount Transmitter Review" (February 15,1990).

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1 References Reference 1 NE-89-L-816, R. M. Kacich to Distribution, " Reporting Require-ments Regarding Noncompliance with NRC Regulations' i (December 13,1989). j J

Reference 2 Letter, E. J. Mroczka (Northeast Utilities) to W. T. Russell (NRC Region I), B12863, " Report of Substantial Safety Hazard" (March 25, 1988).

Reference 3 Letter, Rosemount, Inc. (Steve Wanck) to Northeast Utilities (Director of QA), " Notification Under 10 C.F.R. 21" (February 7, 1989).

Reference 4 REF 89-10, Completed Reportability Evaluation for Millstone r Unit 3 (February 22,1989).

Reference 5 REF 89-09, Completed Reportability Evaluation for Millstone Unit 2 (March 10, 1989).

Reference 6 REF 89-08, Completed Reportability Evaluation for Millstone L

Unit 1 (March 14, 1989).-

[ Reference 7 Letter, U.S. Nuclear E. J. Mroczka (Northeast Utilities) to Transmitters,"-

Regulatory Commission, B13178, "Rosemount (April 13, 1989).

Reference 8 NRC Information Notice 89-42, " Failure of Rosemount Models 1153 and 1154 Transmitters," (April 21,1989).

Reference 9 NRC Inspection Report 50-423/89-04, dated June 28, 1969.

Y Reference'10 GSP-89-299, " Operability Determination for Rosemount Fressure and Differential Pressure Transmitters at Millstone Unit 3" (July 31,1989). ,

Reference 11 Letter, E.- J. Mroczka (Northeast Utilit'.es) to U.S. Nuclear Regulatory Commission, A08132, " Response to Inspec-tion 50-423/89-04" (August 1, 1989). ,

Reference 12 Letter, E. J. Mroczka (Northeast Utilities) to U.S. Nuclear Regulatory Regulatory Commission, B13366, "Rosemount Transmit-ters" (October 31,1989).

Reference 13 Letter, D. L. Fuller (Westinghouse) to A. R. Roby (Northeast Utilities), NEU-90-518, "Rosemount Transmitter Review" (February 15,1990).

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ML20012B151

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