ML20199L327
| ML20199L327 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 01/18/1999 |
| From: | NORTHEAST NUCLEAR ENERGY CO. |
| To: | |
| Shared Package | |
| ML20199L310 | List: |
| References | |
| NUDOCS 9901270198 | |
| Download: ML20199L327 (22) | |
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Docket No. 50-336 B17623 Millstone Nuclear Power Station, Unit No. 2 Proposed Revision to Technical Specifications Secondary Containment Bypass Leakage Marked Up Pages s
1 January 1999
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9901270198 990118 PDR ADOCK 05000336-P PDR ;
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, CONTAINMENT SYSTELi Sep;Echai 20, jusG--
CONTAINMENT LFAKAGE LIMITING CONDITION FOR OPERATION 3.6.1.2 Containment leakage rates shall be limited to:
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a.
An overall integrated leakage rate of <
i weight of the containment air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />., 0.50 percent by [
L at P., 54 psig.
b.
A corbined leakage rate of < 0.60 L, for all penetrations and valtes subject to Type B and C tests when pressurized to P
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c.
A combined leakage rate of < +Alf L for all penetrations
-ident4f4 edin-Table--3:64-asg%t-o P,.
leakage paths when pressurize econ ary containment bypass 1
1 APPtlCABILITY: MODES 1, 2, 3 and 4.Yn/ an ACTION:
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With either (a) the measured overall integrated contai exceeding 0.75 L., or (b nment leakage rate penetrations and valves ) subject to Types B and C testsfw (c) with the combined bypass leakage rate exceedin exceeding 0.60 L., or
) leakage rate (s) to within the limit (s) prior t g 9417 L., restore the System temperature above 200*F.
o increasing the Reactor Coolant
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SURVEILLANCE REQUIREMENTS z
4.6.1.2 with the Containment Leakage Rate Testing ProgramThe cordance l
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filtLSTONE-UNIT 2 3/4 6-2 6cndment Ho. J7;, JEp, pp/, (([
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September 20.1996
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i This Page Intentionally Deleted l
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l MILLSTONE - UNIT 2 cas 3/4 6-3 Amendment No. gy, ypy,203
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September 20.1996
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No Cbenge 5, I n' & ~ h ~
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MILLSTONE - UNIT 2-1/4 s.,
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_ TABLE 3.6-1 E
n SECON RY CONTAINMENT BYPASS LEAK 4GE PATH 1 PENETRATION NO.
SYSTEM a
RELEASE LOCATION 14 Normal Sump Unit Stack via Aerated Waste rain Tank Ve
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67
.s Refueling Water Pu frication Unit 2 Stadk via Auxiliary Building
'N Ventilation hy tem above Spent Fuel
- Poc1, 63 fueling Water Purificat qn Unit 2 Stack via Au,liiary Building N
y Ventilation System ab ve Spent Fuel Pool.
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= 3/4.6 CONTAINMENT SYSTEMS BASES 3/4.6.1 PRIMARY CONTAINMENT
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3/4.6.1.1 CONTAINMENT INTEGRITY Primary CONTAINMENT INTEGRITY ensures that the release of 1
radioactive materials from the containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the accident analyses. This restriction, in conjunc-i tion with the leakage rate limitation, will limit the site boundary radiation doses to within the limits of 10 CFR 100 during accident conditions.
i Primary CONTAINMENT INTEGRITY is required in MODES 1 through s.
This requires an OPERABLE containment automatic isolation valve system.
In MODES 1, 2, and 3 this is satisfied by the automatic containment isolation signals generated by low pressurizer pressure and high containment pressure.
In MODE 4 the automatic containment isolation
. signals generated by low pressurizer pressure and high containment pressure are not required to be OPERABLE. Automatic actuation of the containment isolation system in MODE 4 is not required because adequate time is available for plant operators to evaluate plant conditions and respond by manually operating engineered safety features components.
Since the manual actuation (trip pushbuttons) portion of the containment isolation system is required to be OPERABLE in MODE 4, the plant operators can use the manual pushbuttons to rapidly position all automatic containment isolation valves to the required accident position.
the requirement for an OPERABLE containment automatic isolation system in MODE 4.
t 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 accident analyses at the peak accident pressure of P,.
As an added conservatism, the measured overall integrated leakage rate is further limited to < 0.75 L during performance of the periodic tests to account for possible degradation of the containment leakage barriers between leakage tests.
The surveillance testing for measuring leakage rates is in acco{ancewiththeContainmentLeakageRateTestingProgram.
hEf r 3/4.6.1.3 CONTAINMENT AIR LOCKS A
The limitations on closure and leak rate for the containment n
j air locks are required to meet the restrictions on CONTAINMENT INTEGRITY and leak rate given in Specifications 3.6.1.1 and 3.6.1.2.
The limitations on the air locks allow entry and exit into and out of the containment during operation and ensure through the surveillance testing that air lock leakage will not become excessive through continuous usage.
MILLSTONE - UNIT 2 B'3/4 6-1 Amendment No. /79 797, //)
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l lNSERT A - Paae B 3/4 6-1 o
The Millstone Unit No. 2 FSAR contains a list of the containment penetrations that have been identified as secondary containment bypass leakage paths.
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dit/Lf' ADMINISTRATIVE CONTROLS September 20,1996 o, s_a E
S.19 CONTAINMENT RKif1GE ' RATE TESTING PR6dl0@. _..
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T' A program shall be established.to implement the leakage rate testing o primary containment as required by 10CFR50.54
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Option B as modified by approved exenptions. (o) and.10CFR50Appendii I
i accordance the guidelines contained in This p'atoryrogram shall -bei rin'
'Perfomance with Regul Guide Based containment Leak-Test Program," dated September.1995.1.163, M
The peak calculated primary Containment internal pressure for the loss of coolant accident is P.
The maximum allowable primary containment leakage rate, L, at P l
primary containment air-weight per day.
3
., is 0.5% of Leakage rate acceptance critaria are:
l Primary conta b..... c..:'
a.
laakage rate acceptance criterion is < 1.0 L,.
During the first unit startup following testing in accordance with this i
- program, the ' leakage rate acceptance criteria.are < 0.60 combined Type B and Type C tests, and < 0.75 L, for Type A tests; L, for the b.
Air lock testing acceptance criteria are:
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1.
Overall air lock leakage rate is s 0.05 L, when' tested at t P
'E 2.
For each door, t 25 psig for at least 15 minutes. pressure decay is s 0.1 psig when pressurized
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The provisions of SR 4.0.2 do not apply for test frequencies specified in l
l Primary Containment Leakage Rate Testing Program.
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The provisions of SR 4.0.3 are applicable to the Prinnry Containme Rate Testing Program.
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[l),1stoneUnit2 6-26 Amendment No. 203 l
l Docket No. 50-336 B17623 Millstone Nuclear Power Station, Unit No. 2 Proposed Revision to Technical Specifications Secondary Containment Bypass Leakage Retyped Pages January 1999
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CONTAINNENT SYSTEMS CONTAINNENT LEAKAGE LINITING CONDITION FOR OPERATION 3.6.1.2 C m tainment leakage rates shall be limited to:
a.
An overall integrated leakage rate of < L, 0.50 puunt by weight of the containment air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at P., 54 psig.
b.
A combined leakage rate of < 0.60 L, for all penetrations and valves subject to Type B and C tests when pressurized to p,.
A combined leakage rate of < 0.'J072 L, for all penetrations J
c.
that are secondary containment bypass leakage paths when pressurized to P,.
APPLICABILITY: N0 DES 1, 2, 3 and 4.
ACTION:
With either (a) the measured overall integrated containment leakage rate exceeding 0.75 L., or (b) with the measured combined leakage rate for all penetrations and valves subject to Types B and C tests exceeding 0.60 L., or (c) with the combined bypass leakage rate exceeding 0.0072 L, restore the
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leakage rate (s) to within the limit (s) prior to increasing the Reactor Coolant System temperature above 200'F.
SURVEILLANCE REQUIRENENTS 4.6.1.2 The containment leakage rates shall be demonstrated in accordance with the Containment Leakage Rate Testing Program.
NILLSTONE - UNIT 2 3/4 6-2 Amendment No. J7J, JS, J77, 797, o4ae
This Page Intentionally Deleted.
MILLSTONE - UNIT 2 3/4 6-5 Amendment No.
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jfi,6 CONTAINMENT SYSTEMS BASES a
9 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 accident analyses. This restriction, in conjunc-tion with the leakage rate limitation, will limit the site boundary radiation doses to within the limits of 10 CFR 100 during accident conditions.
Primary CONTAINMENT INTEGRITY is required in MODES 1 through 4.
This requires an OPERABLE containment automatic isolation valve system.
i In MODES 1, 2, and 3 thir, is satisfied by the automatic containment iMatian signals generated by low pressurizer pressure and high c;
,..ent pressure.
In MODE 4 the automatic containment isolation signals generated by low pressurizer pressure and high containment pressure are not required to be OPERABLE. Automatic actuation of the i
containment isolation system in MODE 4 is not required because adequate time is available for plant operators to evaluate plant conditions and respond by manually operating engineered safety features components.
Since the manual actuation (trip pushbuttons) portion of the containment isolation system is required to be OPERABLE in MODE 4, the plant operators can use the manual pushbuttons to rapidly position all automatic containment isolation valves to the required accident position. Therefore, the containment isolation trip pushbuttons satisfy the requirement for an OPERABLE containment automatic isolation valve system in MODE 4.
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 l
assumed in the accident analyses at the peak accident pressure of l
P,.
As an added conservatism, the measured overall integrated l
1eakage rate is further limited to < 0.75 L during performance of the periodic tests to account for possible degradation of the l
containment leakage barriers between leakage tests.
The surveillance testing for measuring leakage rates is in accordance with the Containment Leakage Rate Testing Program.
The Millstone Unit No. 2 FSAR contains a list of the containment penetrations that have been identified as secondary containment bypass leakage paths.
3/4.6.1.3 CONTAINMENT AIR LOCKS The limitations on closure and leak rate for the containment air Tocks are required to meet the rc:trictions on CONTAINMENT INTEGRITY and leak rate given in Specificctions 3.6.1.1 and MILLSTONE - UNIT 2 B 3/4 6-1 Amendment No. J7J, 79), 7JJi, 0440
3/4.6 CONTAINMENT SYSTEMS BASES 3.6.1.2.
The limitations an the air locks allow entry and exit into and out of the containment during operation and ensure through.the surveillance testing that air lock leakage will not become excessive through continuous usage.
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MILLSTONE - UNIT 2 B 3/4 6-la Amendment No.
o44o
Docket No. 50-336 B17623 1
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Millstone Nuclear Power Station, Unit No. 2
. Proposed Revision to FSAR Secondary Containment Bypass Leakage j
Marked Up Pages j
January 1999
94-(Yz-9 MNPS-2 FSAR gjo c jyg,fgf 5.3 ENCLOSURE BUILDING hf hMAsdhts OttiL f 5.3.1 General Description The enclosure building is a limited leakage st te' framed structure with uninsulated metal siding and an insulated roof deck. It also incluues those portions of the auxiliary building adjacent to the containment, as shown in Figures 5.3-1 and 5.3-2. The enclosure building completely surrounds the containment above grade and is designed and constructed to ensure that an acceptable upper limit of leakage of radioactive materials to the environment would not be exceeded in the unlikely event of a loss-of-coolant incident.
To assure the required degree of air-tightness and maintain partial vacuum within the ene,losure building, preformed neoprene gaskets are provided at all joints. Two cr.ntinuous lines of caulking are provided at alllap joints of both siding and decking.
Principal dimensions of the enclosure building are as follows:
Length (ft) 153.0 Width (ft) 147.0 Height (ft) 147.0 Decking (gauge) 20 Siding (gauge) 22 n 24 st W The enclosure building is supported partially on concrete grade beams and caissons, partially on the roof of the auxiliary and turbine buildings, and partially on the dome of the containment. The interior of the enclosure building contains permanent ladders, stairways and catwalks which provide access to the upper exterior regions of the cor'tainment and to equipment in this building. In addition, permanent work platforms are fumished for the periodic surveillance of the post-tensioned prestressing tendons.
Concrete floor slabs are provided at grade between the enclosure building and the contain-ment, and also at Elevations 36-6 and 38-6. A waterproofing mernbrane is provided under the slabs at grade and is extended down and around the containment below grade, as shown on Figure 5.3-3. Between the waterproofing membrane and the containment wall, corrugated asbestos-cement siding is installed as shown in Figure 5.3-4, to provide a passage of least resistance for possible leakage from the containment below grade to the enclosure building.
There are two stacks that exhaust radioactive effluents from the Millstone Unit 2 opera-tions. Radioactive effluenis are piped to the ventilation stack that was provided for the Millstone Nuclear Power Station, Unit 1. This stack provided for future expansion to accept effluent gases from the Unit 2 plant. The physical features of the stack are provided in Section 3.8 of the FSAR for the Millstone Nuclear Power Station, Unit 1. Gas volume increase is less than one percent, resulting in an exit gas velocity of 5,723 feet per minute. Section 3.8, of the Unit 1, FSAR lists the ventilation stack as a Class I structure.
This section outlines the design criteria for earthquake loading and the dynamic analysis applied to the structure. Stack failure would not directly impact any safety related equipment.
5$3.MP2 5.3-1 May 1998 l
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' ' ' o MNPS-2 FSAR NO CN4W6E
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Qul9 The only other stack that exhausts radioactive effluents to the atmosphere from Unit 2 is the stack located atop the enclosure building. The stack is, constructed of 1/4 inch steel l2hY1 plate *and standard structural shapes. Overall height is 13 feet. This is a seismic Class I stack, designed in accordance with the criteria contained in Gection 5.3.3 of the FSAR.
The stack has a constant rectangular cross section which has dimensions of 4'0" x 9'6".
Exit velocity of effluents is 1,684 feet per minute with two fans operating and 2,526 feet per minute with three fans operating. During normal plant operation, two fans are operating, 5.3.2 Construction Materials The following materials are used in the construction of the enclosure building:
- a. Structural steel i
ASTM A-3G
- b. Concrete (psi)
Grade beams and Caissons 4,000 Slabs at grade 3,000 Floor slabs 3,000 i
- c. Reinforcing steel ASTM A615, Grade 60
- d. Metal siding 22 gauge
')
- e. Metal roof decks 20 gauge 5.3.3 Design Bases The design of the enclosure building provides the required features as outlined in General
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Design Criteria 1,2,3,4,5, 60, Appendix A of 10 CFR Part 50.
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j 5.3.3.1 Gases for Design Loads 1
i The following loads are considered in the design of the enclosure building:
- a. Dead loads
- b. Live laads including extemal pressures
- c. Earthquake loads
- d. Wind and tornado loads 5.3.3.1.1 Dead Loads The dead loads consist of the weight of the steel frame, roof, metal siding, and access stairs arul ladders.
5.3.3.1.2 Live Loads The design live loads for the enclosure building are as follows:
mw2 5.3-2 July 1993 l c
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94-#2. -y MNPS-2 FSAR 4/o C//py N OfoSol AT2r Roof, snow loads (psf) 60 Our
. b. Slabs at grade Equipment hatch area AASHO H-20 truck load Other ansat (psf) 500 Extemal pressure i
(independent of wind and tornado loads) (psf) 10 Weights of equipment as indicated on drawings supplied by the manufacturer are included as live loads.
5.3.3.1.3 Earthquake Loads
%e earthquake loads are predicated on an operating basis earthquake (OBE) at the site having a horizontal ground surface acceleration of 0.09 g. In addition, a design basis i
I earthquake (DBE) having a horizontal ground surface acceleration of 0.17 g is used to check the design of the enclosure building to ensure no loss of structural function. The
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seismic design spectrum curves are given in Figure 5.8-1 and 5.8-2. A vertical component j
two-thirds of the magnitude of the horizontal ground surface component is applied at the br a simultaneously, i
A dynamic analysis includiag the effects of the attachments to the other structures is used l;
1 J-to arrive at the equivalent static loads for the design.
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5.3.3.1.4 Wind and Tornado Loads t
Winds loads for the design of the enclosure building are based on a wind velocity of 115 mph with gusts up to 140 mph. The ASCE Paper 3269 is used to determine the shape factors. However, the provisions in the paper for gust factors and variations of i
wind velocity with respect to height are not applied.
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The entire enclosure building is designed to resist the effects of the 140 mph wind.
I Tornado loads on the enclosure building are based on a tomado funnel having a peripheral tangential velocity of 300 mph and a translational velocity of 60 mph. These velocities are combined, resulting in a design basis tornado wind velocity of 360 mph. The enclosure building, adjacent to structures which house safety related equipment, is designed so that its structural framing will withstand tornado winds, but the siding will be blown away.
I The wind velocity is assumed to be uniformly distributed over the height of the struc.ture.
Probable missiles in the form of siding are less critical than the design missiles as spaified in Section 5.2.6.1.2 of the FSAR. The siding when blown off may induce superficial damage to the adjacent structures, but the structural integrities of the adjacent structures will be maintained. The design requirements for tornado loads for structures which house safety related equipment for shutdown are given in Section 5 of the FSAR.
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5.3-3 July 1993 l
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MNPS-2 FSAR h 2 h fof/I+7%cw i
Qn)C V The design of the enclosure building for tornado loads assumes that tornado wind is not coincident with a loss-of-coolant accident (LOCA) or earthquake.
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5.3.3.2 Design Load Combination and Structural Analysis The enclosure building is designed to meet the rerformance and strength requirements of the following loading combinations:
. a. At design loads
- b. At factored loads The design of structural steel is in accordance with the AISC Manual.of Steel Construction.
The design of concrete is in accordance with the ACI Code 318-63.
i 5.3.3.2.1 At Design Llads
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The enclosure building is analyzed for the following specific loading conditions:
- a. D + L Construction
- b. D + L + E Operating
- c. D + L + W Operating Where D = dead loads L = live loads E = operating basis earthyake loads (0.09 g)
W = wind or tornado loads 5.3.3.2.2 At Factored Loads The enclosure building is analyzed for the factored load combination to ensure tikat its structural integrity is not affected.
C = 11(1.0)(D) + E']
0 Where C = required capacity of the structure 0 = 0.90 for fabricated structural steel D = dead loads l
E' = design basis earthquake (0.17 g) sea.m 5.3-4 July 1993 l i/
9@M W p,
MNPS-2 FSAR l
The stresses of the members of the structure at factored loads are limited to the yield stres'ses of the structural steels.
5.3.3.2.3 Seismic Analysis The seismic analysis of the enclosure building is made on a mathematical model which consists of the lumped masses of the containment structure and the enclosure building.
The seismic response of the combined modelis obtained in accordance with the proce-dures outlined in Section 5.8.
I 5.3.4 Through-Line Leakage Evaluation l y@'8 To evaluate the through-line leakage that can bypass the enclosure building filtration region j
(EBFR), the fluid systems penetrating containment are categorized as follows:
- a. Piping System o the containment post-accident atmosphere.
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- b. Piping Systems which are closed and therefore not exposed to the containment l u,,aa post-accident atmosphere.
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The following assumptions are made to postulate the maximum hypothetical conditions:
- a. There is either a seismic occurrence and all Seismic Class 2 lines are broken, or l
there is no seismic occurrence and all Seismic Class 2 lines are intact.
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- b. The single failure criterion applies to Seismic Class 1 components only.
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The containme enetrations normall pen to the containmen nclude the following:
Pen.No.
System 4,5 Contain sent Sprays 4
1,13 Cont 'nment Sump Recirculati 4
N al Sump to Aerated Dr Tank 39,40 urge Air inlet and Purge r Discharge 9 N '3 42 Fuel Transfer Tube 61,86 Containment Air Moni oring l
62,87,88,89-Hydrogen Sample 67,68 Refueling Water P rification 82,83 Hydrogen Purge The condition of a seismic occurrence is not considered. Should the pipe break within the EBFR, all potential containment leakage would be processed by the enclosure building filtration system (EBFS) as per design. The EBFS has ample capacity for this event.
he contai ent spray system is failure c ierion is applicable. T e containment spraengineered safety eat e single system, except for thef re j
water torage tank (RWST),
contained within t EBFR. Therefore, t worst failure is
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as med to be a loss of e ergency power to o e (1) spray subsyste hich renders the mpinoperable and th otor-operated val in the open position igure 6.1-1) during ssa.ue2 5.3-5 May 1998 4
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e Q{~tW L ^1 MNPS-2 FSAR khe injection phase of oper tion. Containment leakage must pa s through two (2) s check valves and must greater than the minimum head of 0 ft. (25 psig) from the RWST. The model f ulated for valve leakage is that wit the smallleakage rate a the containment is ation valves lassumed conservativel as 2.0 cc/hr. per inch val e oss diameter at 54 ig) and the transient conditions of th containment pressure (de easing with time) the aximurn pressurization from leakage 'n a hypothetical penetratio at atmospheri onditions closed by valves at both e s is about 25% of the con inment pressure.
herefore, the head of water from th WST will prevent contain nt leakage.
The c ntainment sump recirculation piping i an engineered safety featur and contained wit n the EBFR. Assuming any failure, t containment leakage is eith r contained within t
EBFR or is prevented by the check Ive and elevation head from ue RWST.
The line from the normal sump to e aerated drain tank could pro de a potential path for bypass leakage. Assuming the I akage through the valve is pro. rtional to the square root of the pressure differential, th. maximum leakage through the o (2) series containment is 3 cc/hr. at containment st-accident design conditions.
fter approximately one (1) j hour the containment pre ure is less than 10 psig and the akage rate is less than j
1.0 cc/hr. This valve le age is diluted in the aerated w te system.
The containment a enclosure building purge syste is routed through the enclosure
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building into the xiliary building and is open to th atmosphere. However, the branch i
ductwork whic serves the enclosure building fro the containment purge system has been revised ch that the isolation dampers (
. 2.AC-3 and 2-AC-8 on Figure 9.9% fail l
in the open osition. This arrangement will v t the containment purge penetratioK l
leakage i o the enclosure building and thu eliminating potential through-line lea ' age from bypassi the EBFR.
l The ydrogen purge system is connec d to the suction of the enclosure b ding filtration l
sy em (EBFS). Therefore, all leaka through these penetrations is procpssed with the
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fluents from the EBFR. The hyd gen sample and containment air m itoring systems are closed loop systems with all omponents located within the EBF, Therefore, any leakage from the systems is c ntained within the EBFR and, thus is EBFS.
rocessed by,the The refueling water puri ation penetrations could provide a tential path of bypass leakage into the auxiii y building. The maximum leakage i approximately 4.0 cc/hr.
through each penett ion during the first hour. After this e maximum leakage is approxi mately 1.0 cc/hr.
ssuming normal system alignment ( gure 9.5-1), the leakage is dilut d j
and contained w' in the closed process piping.
l The fuel tran er tube, identified as penetration N,42 in FSAR Table 5.2-11, with etails provided in SAR Sections 5.2.7.1.4, provides connection between the contain ent and auxiliary uilding, thereby bypassing the EBFR Leakage from containment is mi imized by ]8'S l
the do le o-ring arrangement of the blind fl nge that seals the containment d of this pene ation. Any leakage is further restric d by the head of water exerted n the tube by the ormal water levelin the spent fuel of. No credit is taken for the fu transfer gate ve as a leakage barrier. Based on a/ouble o-ring seal arrangement o the simple blind v
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ange, the design leakage across th sf barrier is assumed to be O cc/hr.
riodic leakage j
1 ssa un 5.3-6 i
May 1998 I
e 4
C} 9 -MfL - g MNPS-2 FSAR t results for is leakage pa are included with er EBFR bypass age path
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From the basis formulated, systems which are not normally opened to the containment atmosphere, or normally closed systems, either do not leek (assuming no seismic event) or are vented to the EBFR (assuming a seismic event). Normally, closed systems which may be opened to the atmosphere during accident conditions, such as lines connected to the i
reactor coolant pressure boundary, are not considered. These systems are either operating at a higher pressure or form closed loops. Therefore, assuming a single failure, these lines i
either prevent leakage by the higher pressure or contain the leakage by the closed loop.
from the pr ding analysis, only the akage through the norma ump to aerated drain tank and fueling water purificati penetrations may be co dered as through line leaka. The maximum potent' leakage rate is 11.0 cc
. during the first hour an ess q
th 3.0 cc/hr. thereafter, other leakage is either ntained within the proep system j
within the EBFR.
The provisions for initial and periodic leak testing of containment penetrations and maximum allowable leakage are specified in Table 5.2-11 of the FSAR and Section 3.6.1.2.c of the Technical Specifications respectively.
97.i75 t/ifr Jos?Rr M
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$$3 MP2 5.3-7 May 1996 i
4 INSERT A - FSAR Paae 5.3-7 From this basis, an evaluation was performed to establish and document those containment penetrations that have the potential for providing leakage pathways from the reactor containment to areas beyond the EBFR. These leakage pathways could result in Post-Accident Containment Atmosphere bypassing the EBFR and discharging directly to the atmosphere thereby increasing onsite and off-site doses under radiological accident conditions. Leakage through these pathways is referred to as
" bypass leakage."
For a leakage pathway to viably result in bypass leakage, the pathway must be open to the containment atmosphere post-accident and provide a means of transporting the containment atmosphere beyond the EBFR as well as a means for the containmer,i attriosphere to escape the piping or duct.
For containment penetrations that are confirmed to contribute to bypass leakage, leakage rates may be based on measured values as opposed to the recommended or maximum allowable values used for testing.
In cases where measured values are used, steps are taken to ensure that degradation of the valve sealing capabilities are taken into account commensurate with the severity of service and the required time intervals between valve maintenance.
The evaluation concluded the following penetrations are considered to qualify as systems that are open both inside containment and outside containment, extend beyond the EBFR, and could contribute to bypass leakage:
Penetration System 14 Containment Sump Pump Discharge 37 Instrument Air System 38 Station Air System 42 Fuel Transfer Tube 61 Hydrogen Monitoring System 62 Hydrogen Monitoring System 67 Refueling Cavity Drain 60 Refueling Cavity Skimmer 85 Containment Pressure Test Connection 86 Hydrogen Monitoring System 87 Hydrogen Monitoring System The off-site and control room dose analyses are based on a calculated maximum bypass leakage (Refer to Section 14.8.4). For each verified bypass leakage pathway,
4 4'+
o a recommended leak rate is provided based on the limits used to satisfy the leakage limits established for the testing required by 10 CFR 50, Appendix J. Total leakage from all verified bypass leakage pathways will be summed and compared to the total limit.
The control room and off-site radiological dose calculations establish the maximum limit for total bypass leakage. In the event that total bypass leakaga exceeds this value, repairs will performed to reduce bypass leakage to an acceptable level.
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