Text

Proposed Tech Specs,Implementing 10CFR50,App J Option B for Containment Leak Rate Testing
	ML20095E413
Person / Time
Site:	Browns Ferry
Issue date:	12/08/1995
From:	; TENNESSEE VALLEY AUTHORITY
To:	;
Shared Package
ML20095E406	List: ML20095E405; ML20095E413;
References
	NUDOCS 9512150069
	Download: ML20095E413 (58)
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ENCLOSURE 2 TENNESSEE VALLEY AUTHORITY EROWN8 FERRY NUCLEAR PLANT (BFN)

UNITS 1, 2, AND 3 PROPOSED TECHNICAL SPECIFICATION (TS) CHANGE TS-364 NARKED PAGES I.

AFFECTED PAGE LIST Unit 1 Unit 2 Unit 3 3.7/4.7-3 3.7/4.7-3 3.7/4.7-3 3.7/4.7-4 3.7/4.7-4 3.7/4.7-4 3.7/4.7-5 3.7/4.7-5 3.7/4.7-5 3.7/4.7-6 3.7/4.7-6 3.7/4.7-6 3.7/4.7-7 3.7/4.7-7 3.7/4.7-7 3.7/4.7-25 3.7/4.7-25 3.7/4.7-24 6.0-24 6.0-23c 6.0-23c II.

MARKED PAGES See attached.

l

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9512150069 951200 PDR ADOCK 05000259 P

PDR

3.7/4.7 CONTAINMENT SYSTEMS DEC 0 71994 SURVEILIANCE REOUIREMENTS LIMITING CONDITIONS FOR OPERATION 4.7.A.

Primary Containment 3.7.A.

Primary Conta4nment

2. Interrated Leak Rate Testing 2.a.

Primary containment integrity shall be maintained at all times Primary containment nitrogen when the reactor is critical consumption shall be monitored to determine the or when the reactor water temperature is above 212*F average daily nitrogen and fuel is in the reactor consumption for the last 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Excessive leakage vessel except while is indicated by a N2 performing "open vessel" consumption rate of > 2% of physics tests at power the primary containment free levels not to exceed volume per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 5 MW(t).

(corrected for drywell

b. Primary containment temperature, pressure, and integrity is confirmed if venting operations) at the maximum allowable 49.6 psig. Corrected to norinal drywell operating integrated leakage rate, L,, does not exceed the pressure of 1.1 psig, this value is 542 SCFH.

If this equivalent of 2 percent of value is exceeded, the the primary containment volume per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at the action specified in 3.7.A.2.C shall be taken.

49.6 psig design basis accident pressure, P,.

n The containment leakage r 4

shall be demonstrated the

c. If N2 makeup to the primary containment averaged over following test se ule and 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (corrected for shall be det ned in accordanc th Appendix J pressure, temperature, and venting operations) exceeds to 10 50 as modified j

542 SCFH, it must be reduced b

pproved exemptions.

to < 542 SCFH within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or the reactor shall be a 3 ree typ t sts (overall integrated placed in Hot Shutdown containment leak rate) within the next 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months .

shall be cond ed at I

40 1 10 intervals i

during s tdown at P,,

49.6 g, during each 10- ar plant inservice spection.

D Hr AMENDMENT NO. 213 3.7/4.7-3 BW Unit 1

INSERT A TS 4.7.A.2 SECOND PARAGRAPH Perform leakage rate testing in accordance with the Primary Containment Leakage Rate Testing Program.

\\

3.7/4.7 CO]rrAI] DENT SYSTEMS g

SURVEf t.t.ANCE REOUIrim.nd Lih m us CONDITIONS FOR OPERATION 4.7.A.

Primary Containment 4.7.A.2. (Cont'd) b.

If any periodic type A test fails t meet 0.75 L.,

the test schedule for sub quent type A tests a 1 be reviewed and a proved by the Commissi If two co ecutive type A tests fa to meet 0.75L,,

type A test

[

shall e performed at less every 18 months unt two consecutive t e A tests meet 0 5 L, at which time he above test schedule may be resumed.

I '. Test duration shall

)

c.

be at least 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

j 2.

A 4-hour stabilization eriod will be requ ed and the contai ont I

atmosphor will be consid stabilized when change in weig ed average air tem rature averaged or r an hour does not viste by more than

.5'R/ hour from the average rate of

\\

change of temperature measured from the previous 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

1 1

E I41 3.7/4.7-4 BFit Unit I

... _ ~. _

9 4

FEB 0319 3.7/4.7 CONTAINMENT SYSTEMS I

LIMITING CONDITIONS FOR OPERATILN SURVEILLANCE REQUIREMENTS i

i 4.7.A, Primary Containment

{

1 l

4.7.A.2. (Cont'd) l r-

,y_

d. 1. At least 26 sets of-dat points at approirimate equal time interva and in i'

no case at inte s

greater t hour shall be provid or proper statistic analysis.

2. The sure of merit for the i-in rumentation systea 11 never estceed 0.25 L

a t

e! 'The test shall not - %

concluded creasing d le,ak rate.

4 l

n.

L

. The accuracy of each type j

test shall be verified by a i

supplemental test which:

l l

l'. Confirms the accura y of i

the test by verify ng that the difference be en the i

supplemental dat and the i

type A test dat is within j

0.25 L -

a f

2. Has duration sufficient to establish a curately the change in eskage rate between t a type A test and the supp emental test.

3. Requir the quantity of j*

gas i jected into the cont nment or bled from l

the ontainment during the su lesental test to be ivalent to at least 3

2 percent of the total

', fi asured leakage at Pa d

(49.6 psig).

,i l i

i arN 3.7.4.7-5 d,

Unit 1 AMENDMENT NO.141 t

3.7/4.7 CONTAInumrT SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIRE ENTS 4.7.A.

primary Containment 4.7.A.2. (Cont'd)

g. Local' leak rate tests (LLRT )

shall be performed on the primary containment testab o penetrations and isolatio valves, which are not pa of a water-sealed system, a not 1ess than 49.6 psig (ex opt for the main steam iso tion valves, see 4.7.A.2.1) and not less than 54.6 psis f e water-sealed valves ch operating cycle. Bo ted double-gasketed sea s shall be tested whenever th seal is closed after being opened and at least once per operating cycle. Acceptab methods of testing are hali e gas detection, soap

bbles, pressure decay, hydrostaticall pressurized fluid flow equivalent.

i The personne air lock shall be tested a 6-month intervals at an inte al pressure of not less than 9.6 psig. In

addition, f the personnel air lock is o ened during periods when con ainacat integrity is not req red, a test at the end of uch a period will be conduc ed at not less than 49.6 is. If the personnel air 1 ek is opened during a peri d when containment int city is required, a test at 2.5 psis shall be co ducted within 3 days after b ng opened.

If the air lock i opened more frequently than ce every 3 days, the air och shall be test-2 at least once every 3 days during the period of frequent openings.

N BrN 3.7/4.7-6 Unit 1

?

3.7/4.7 ColrratuurnT ani=r-gg LIMITING cnunITIONS FOR OPQ[7 ION surviiif ANCE Eisdiin 15 4.7.A.

Primary Contahant 4.7.A.2.g (Cont'd) c The total leakage from a penetrations and isolat n 1

valves shall not excee 60

)

percent of La Per 24' oura.

Leakage from contai t

j isolation valves t t 1

terminate below a pression pool water level y be excluded from t e total leakage pro d a sufficient fluid inv is available to ensure t sealing function f at least 30 days at a pres re of 54.6 pois.

Leakage rom containment isolati valves that are in close loop, seismic class I line that will be water sea d during a DBA will be me ured but will be excluded en computing the total eakage.

h j fl' AMmDMM NO.159 BFN 3.7/4.7-7 Unit 1

INSERT B 4

i PARAGRAPH 4.7.A.2.g Perform required local leak rate tests, including the primary containment air lock leakage rate testing in accordance with the Primary Containment Leakage Rate Testing Program.

Note:

An inoperable air lock does not invalidate the previous successful performance of the overall air lock leakage test.

The acceptance criteria for air lock testing are:

(1) Overall air lock leakage rate is s (0.95 L.) ir lock leakage rate is when tested at 2 Pa.

(2)

For door seal leakage, the overall a s (0.02 L.) when the air lock is pressurized to (2 2.5 psig for at least 15 minutes).

E 16 @

3.7/4.7 B&IEE 3.7.A & 4.7.A Primary Containment The integrity of the primary containment and operation of the core standby cooling system in combination, ensure that the release of radioactive materials from the containment atmosphere will be restricted to those leakage paths and associated leak rates asemed in the accident analyses. This restriction, in conjunction with the leakage rate limitation, will limit the. site boundary radiation doses to within the limits of 10 CFR Part 100 during accident conditions.

l During initial core loading and while the low power test program is being conducted and ready access to the reactor vessel is required, there will I

be no pressure on tbc system thus greatly reducing tha chances of a pipe break. The reactor may be taken critical during this period; however, restrictive operating procedures will be in effect to minimize the probability of an accident occurring.

The limitations on primary containment leakage rates ensure that the total containment leakage volume will not exceed the value asemed in the accident analyses at the peak accident pressure of 49.6 pais, P. As an l

added conservatism, the measured overall integrated leakage rate is l

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.

l

, n The surveillance testing for meas==81iik le=6-ae rates ara ~==*=t_t riti l

^ m _1. r-_i. vi a, tx J of 10 CFR Part 50 (type A, B, and C tests).

x.

The pressure suppression pool water provides the heat sink for the l

reactor primary system energy release following a postulated rupture of the system. The pressure suppression chamber water volume must absorb the associated decay and structural sensible heat release during primary I

system blowdown from 1,035 psig. Since all of the gases in the drywell are purged into the pressure suppression chamber air space during a loss of coolant accident, che pressure resulting from isothermal compression plus the vapor pressure of the liquid must not exceed 62 pais, the suppression ebanhar marimun pressure. The design volume of the suppression chamber (water and air) was obtained by considering that the total volume of reactor coolant to be condensed is discharged to the suppression-chamber and that the drysell volume is purged to the suppression chamber.

Using the minium or maximum water levels given in the specification, containment pressure during the design basis accident is approximately 49 peig, which is below the maxim a of 62 pais. The maximum water level indications of -1 inch corresponds to a downconer submergence of three feet seven inches and a vatar volume of 127,800 cubic feet with or 128,700 cubic feet without the drywell-suppression chamber differential pressure control. The minim e water level indication of -6.25 inches with differential pressure control and -7.25 inches without differential pressure control corresponds to a downconer submergence of approximat ely three feet and water volume of approximately 123,000 cubic feet.

3.7/4.7-25l AWDmMDR W. I 8 9 BFN Unit 1 I

~ _ - -.

JUN 211994 accuracy of the measurements of r'adioactive mat'erials in environmental sample' matrices are performed as part of the quality assurance program for environmental monitoring.

Tnfrrb C 6.8.5 PROGRAMS Postaccident Samolina Postaccident sampling activities will ensure the capability to obtain and analyze reactor coolant, radioactive iodines and particulates in plant gaseous effluents, and containment atmosphere samples under accident conditions. These activities shall include the following:

(i)

Training of personnel, (ii) Procedures for sampling and analysis, (iii) Provisions for maintenance of sampling and analysis.

6.9 REPORTING REOUIREMENTS 1

ROUTINE REPORTS 6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following identified reports shall be submitted to the Director of the Regional Office of NRC, unless otherwise noted.

6.9.1.1 STARTUP REPORT a.

A summary report of plant startup and power escalation testing shall be submitted following (1) receipt of an operating license, (2) amendment to the license involving a planned increase in power level, (3) installation of fuel that has a different design or has been manufactured by a BFN 6.0-24 AMENDMM NO. 2 0 7 Unit 1

.. ~.

t INSERT C Section 6.8.4.3 PRIMARY CONTAINMENT LEAKAGE RATE TESTING PROGRAM A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as. modified by approved exemptions.

This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163,

" Performance-Based Containment Leak-Test program,- dated September 1995".

The peak calculated containment internal pressure for the design basis loss of coolant accident, P.,

is 49.6 psig.

The maximum allowable primary containment leakage rate, L.,

at P.,

shall be 2% of primary containment air weight per day, i

Leakage Rate acceptance criteria are:

1 a.

Primary Containment leakage rate acceptance criterion is i

5 1. 0 L,.

During the first. unit startup following testing in accordance with this program, the leakage rate t

acceptance criteria are s 0.60 L, for the Type B and Type C tests and 5 0.75 L, for Type A tests; b.

Air lock testing acceptance criteria are:

1)

Overall air lock leakage rate is s 0.05 L, when tested at 2 P.,

2)

Air lock door seals leakage rate is s 0.02 L, when the overall air lock is pressurized to 2 2.5 psig for at least 15 minutes.

...y,y 7

m m..

3.7/4.7 CONTAINMENT SYSTEMS D E 0 7 19 H SURVEILLANCE REOUIREMENTS LIMITING CONDITIONS FOR OPERATION l

I' 3.7.A.

Pr4 mary Containment 4.7.A.

Pr4 mary Containment 2.

Interrated Leak Rate Testine 2.a.

Primary containment t

j integrity shall be maintained at all times Primary containment nitrogen when the reactor is critical consumption shall be monitored to determine the or when the reactor water temperature is above 212*F average daily nitrogen and fuel is in the reactor consumption for the last l

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Excessive leakage vessel except while performing "open vessel" is indicated by a N2 consumption rate of > 2% of physics tests at power l

levels not to exceed the primary containment free l

5 MW(t).

volume per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (corrected for drywell temperature, pressure, and b.

Primary containment integrity is confirmed if venting operations) at the maximum allowable 49.6 psig. Corrected to integrated leakage rate, normal drywell operating L,, does not exceed the pressure of 1.1 psig, this equivalent of'2 percent of value is 542 SCFH.

If this value is exceeded, the the primary containment volume per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at the action specified in 49.6 psig design basis 3.7.A.2.C shall be taken.

accident pressure, P,.

]

_o e containment leakage r s

shall be demonstrated the c.

If N2 makeup to the primary containment averaged over following tes e

ule and 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (corrected for shall be det ned in.

pressure, temperature, and accordan ith Appendix J to venting operations) exceeds 10 C 0 as modified by 542 SCFH, it must be reduced roved exemptions.

to < 542 SCFH within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ree type'A'tE ts (over or the reactor shall be a.

placed in Hot Shutdown integrated contai within the next 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months .

leakage rate) a be conducted a i 10-month interval uring shutdown at P 9.6 psig, during 10-year plant inservice inspection.

~

TAJEr BfN 3.7/4.7-3 AMENDMENT NO. 2 29 Unit 2

. - =.

-. _. - -. -.. ~

INSERT A TS 4.7.A.2 SECOND PARAGRAPH Perform leakage rate testing in accordance with the Primary Containment Leakage Rate Testing Program.

1 i

3.7/4.7 CONTAT15ENT SYSTEMS F

03 m SURVEIf1ANCE REQUIxiru.nis 1.iru11NG CONDITIONS FOR OPERATION 4.7.A.

Primary Containment 4.7.A.2. (Cont'd) b.

If any periodic type A test fails to a et l 0.75 L, the test schedule for subse ent type A tests shal be reviewed and app owed by the Commission If two cons cutive type A tests fai to meet 0.75L,,

type A test shall performed at least every 18 months I

unt two consecutive

[

t e A tests meet

.75 L,, at which time the above test schedule may be resumed.

c.

1.

Test'durat on shal

~

8 be at least 8 hou 2.

A 4-hour stabilization eriod will be requ ed and the contai t

atmosphe e ill be conside stabilized when th change in weight average air tempe ature averaged over an hour does not der ate by more than 0 *R/ hour from the erage rate of change of temperature measured from the previous 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

.~

3.7/4.7-4 Autwoutxt so. Ia 7 l

ar.

Unit 2 i

FE13 031988 3.7/4.7 CONTAINMENT SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 4.7.A.

Primary Containment 4.7.A.2. (Cont'd)

/d. 1. At least 20 sets of da Points at approximat y equal time interva and in no ease at inte is greater tha o hour shall be provide e proper statistic analysis.

2. The f ure of merit for the ins ntation system s 11 never exceed

.25 La*

e. The tiest shal not*be concluded nereasing sted leak rate.

f. Tlie ' accuracy olf each ype test shall be verified by a supplemental test which*

1

1. Confirms the accuracy of the test by verifyi that the difference bet en the supplemental data and the type A test data is within 0.25 L.

2. Has duration ufficient to establish a urately the change in akage rate between t type A test and the supp emental test.

3. Requir s the quantity of gas i jected into the con inment or bled from th containment during the pismental test to be j

uivalent to at least j

percent of the total measured leakage at P a

(49.6 psig).

j BFN 3.7/4.7-5 1

Unit 2 AMENDMENT NO.13 7 A

_.~4

_.. _ _~

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

t I

l 3.7/4.7 CONTATuarrNT 5YoiEris LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 4.7.A.

Primary Containment 4.7.A.2. (Cont'd)

3. Local leak rate tests LLRT )

shall be performed on the primary containment testab o penetrations and isolati valves, which are not pa of a water-sealed system, a not less than 49.6 pois (ex ept for the main steam iso tion valves, see 4.7.A.2.1) and not less than 54.6 psis f r water-sealed valves ch operating cycle. Bo ted double-gasketed sea s shall be tested when~ever t seal is closed after being opened and at least once per operating cycle. Acceptab methods of testing are hali e gas detection, soap bbles.

pressure decay, hydrostaticall pressurized fluid flow equivalent.

The personne air lock shall be tested a 6-month intervals at an inte al pressure of not less than 9.6 psig.

In addition, if the personnel air lock is ened during periods when con ainment integrity is not req iced, a test at the end of uch a period will be 4

conduc ed at not less than 49.6 is.

If the personnel air ek is opened during a peri d when containment int city is required, a test at 2.5 psig shall be e ducted within 3 days after ing opened. If the air lock opened more frequently than ce every 3 days, the air ock shall be tested at least once every 3 days during the period of frequent openings.

BFN 3.7/4.7-6 Unit 2 I

3.7/4.7 CONTAfteENT SYSTEMS g g g jggj LIMITING CODITIONS FOR OPERATION SURVEIUANCE REQuiannasus 4.7.A.

Primary Conta N ant 4.7.A.2.s (Cont'd)

The total path leakage f om l

all penetrations and I isolation valves shal not

' exceed 60 percent of a per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Leakage om containment isolat on valves that terminate b ow suppression poo water level may be exclude from the total leaka e rovided a sufficient id inventory is available t ensure the sealing f etion for at least 30 days a pressure of 54.6 ps Leakage from contai ent isolation valves that re in closed-loop, i

sei ic class I lines that vi be water sealed during a D

will be measured but will excluded when computing e total leakage.

1 fn fr r l

AMENDMENT NO.19 3 3.7/4.7-7 BFN Unit 2 i

I l ~

L

INSERT B i

PARAGRAPH 4.7.A.2.g Perform required local leak rate tests, including the primary

)

containment air lock leakage rate testing in accordance with the Primary Containment Leakage Rate Testing Program.

Note:

An inoperable air lock does not invalidate the previous j

successful performance of the overall air lock leakage test.

The acceptance criteria for air lock testing are:

(1) Overall i

air.'ock leakage rate is 5 (0.05 L.) when tested at 2 Pa.

(2)

For daor seal leakage, the overall air lock leakage rate is i

s (0.03 L.) when the air lock is pressurized to (2 2.5 psig for at least 15 minutes).

4 3.7/4.7 B&SLi Oj, $ g 3.7.A & 4.7.A Pri==ry conta4===nt The integrity of the primary containment and operation of the core standby cooling system in combination, ensure that the release of

~

radioactive materials from the containment atmosphere will be restricted to those leakage paths and associated leak rates asetaned in the accident smalyses. This restriction, in conjunction with the leakage rate limitation, will limit the site boundary radiation doses to within the limits of 10 CFR Part 100 during accident conditions.

During initial core loading and while the low power test program is being conducted and ready access to the reactor vessel is required, there will

~

be no pressure on the system thus greatly reducing the chances of a pipe break. The reactor may be taken critical during this period; however, restrictive operating procedures will be in effect to minimize the probability of an accident occurring.

The limitations on primary containment leakage rates ensure that the total containment leakage volme will not exceed the value assissed in the accident analyses at the peak accident pressure of 49.6 pais, 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 measur$ag ? ah==e rates men --f-tm C

__the reenh t- -f ^_;;-- % J os AU urs Part 50 (type A, B, and C tests).

The pressure suppression pool water provides the heat sink for the reactor primary system energy release following a postulated rupture of the system. The pressure suppression chamber water volume must absorb the ascociated decay and structural sensible heat release during primary system blowdown from 1,035 pais. Since all of the gases in the drywell are purged into the pressure suppression chamber air space during a loss of coolant accident, the pressure resulting from isothermal compression plus the vapor pressure of the liquid must not exceed 62 pais, the suppression chamber maximum pressure. The design volume of the suppression chamber (water and air) was obtained by considering that the total voltane of reactor coolant to be condensed is discharged to the suppression chamber and that the drywell volume is purged to the suppression chamber.

Using the minimum or maximum water levels given in the specification, containment pressure during the design basis accident is approximately 49 pois, which is below the marimum of 62 pais. The maximum water level indications of -1 inch corresponds to a downconer submergence of three feet seven inches and a water volume of 127,300 cubic feet with or 123,700 cubic feet without the drywell-suppression chamber differential pressure control. The minimum water level indication of -6.25 inches with differential pressure control and -7.25 inches without differential pressure control corresponds to a downconer submergence of approximately three feet and a water volume of approximately 123,000 cubic feet.

BFN 3.7/4.7-25 AMENDMDK NO. 2 0 4 Unit 2

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

l DEC 0 21993

'j. Limitations on the annual dose or dose cosmitment to any MEMBER OF THE PUBLIClue to releases of radioactivity and to radiation from uranium. fuel cycle sources conformina to 40 CFR Part 190.

-6.8.4.2 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM A program shall be provided to monitor the radiation and radionuclides in the environs of the plant. The program shall provide (1) representative measurements of radioactivity in the highest potential exposure pathways, and (2) verification of the accuracy of the effluent monitoring program and modeling of environmental exposure pathways. The program shall (1) be contained in the ODCM, (2) conform to the guidance of Appendix I i

to 10 CFR Part 50, and (3) include the following:

a.

Monitoring, sampling, analysis, and reporting of radiation and radionuclides in the environment in accordance with the methodology and parameters in the ODCM, b.

A Land Use Census to ensure that changes in the use of area at and beyond the SITE BOUNDARY are identified and that modifications to the monitoring program are made if required by the results of this census, and c.

Participation in an Interlaboratory Comparison Program to ensure that independent checks on the precision and accuracy of the measurements of radioactive materials in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring.

~

Y Lnfrrl (

6.8.5 PROGRAMS Postaccident Samnlina j

Postaccident sampling activities will ensure the capability to obtain and analyze reactor coolant, radioactive iodines and AMENDMENT NO. 2 2 0 BFN 6.0-23c Unit 2 4

m r

m.y

INSERT C Section 6.8.4.3 PRIMARY CONTAINMENT LEAKAGE RATE TESTING PROGRAM A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions.

This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163,

" Performance-Based Containment Leak-Test program, dated September 1995".

The peak _ calculated containment internal pressure for the design basis loss of coolant accident, P, is 49.6 psig.

The maximum allowable primary containment leakage rate, L.,

at P.,

shall be 2% of primary containment air weight per day.

Leakage Rate acceptance criteria are:

a.

Primary Containment leakage rate acceptance criterion is

$ 1.0 L,.

During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are 5 0.60 L,-for the Type B and Type C tests and s 0.75 L, for Type A tests; b.

Air lock testing acceptance criteria are:

1)

Overall. air lock leakage rate is s 0.05 L, when tested at 2 P.,

2)

Air lock door seals leakage rate is s 0.02 L when the overall air lock is pressurized to 2 2.5 psig for at least 15 minutes.

i

3.7/4.7 CONTAINMENT SYSTEMS DEC 0 71994 SURVEILIANCE REOUIREMENTS LIMITING CONDITIONS FOR OPERATION 3.7.A.

Pr4 mary Containment 4.7.A.

Primary Containment 2.

Interrated Leak Rate Testine 2.a.

Primary containment integrity shall be maintained at all times Primary containment nitrogen when the reactor is critical consumption shall be monitored to determine the or when the reactor water temperature is above 212*F average daily nitrogen and fuel is in the reactor consumption for the last vessel except while 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Excessive leakage perf onning "open vessel" is indicated by a N2 consumption rate of > 27. of physics tests at power levels not to exceed the primary containment free 5 MW(t).

volume per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (corrected for drywell temperature, pressure, and

b. Primary containment integrity is confirmed if venting operations) at the maximum allowable 49.6 psig. Corrected to integrated leakage rate, nonnal drywell operating L,, does not exceed the pressure of 1.1 psig, this equivalent of 2 percent of value is 542 SCTH.

If this the primary containment value is exceeded, the volume per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at the action specified in 49.6 psig design basis 3.7.A.2.c shall be taken.

accident pressure, P,.

p A

The containment l'ak ge rat e

shall be demonstrated he

c. If N2 makeup to the primary containment averaged over following tes e

ule and 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (corrected for shall be det ned in pressure, temperature, and accord with Appendix J to venting operations) exceeds 10 50 as modified by 542 SCFH, it must be reduced approved exemptions.

to < 542 SCFH within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or the reactor shall be a.

Three type A tests '

~

placed in Hot Shutdown (overall integrated within the next 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months .

containment lea rate) shall be con ed at 40 i 10-mon intervals during a tdewn at P,,

49.6 g, during each 10 ear plant intervice spection.

I nStr OU IB6 BFN 3.7/4.7-3 Unit 3

i INSERT A j

TS 4.7.A.2 SECOND PARAGRAPE Perform leakage rate testing in accordance with the Primary Containment Leakage Rate Testing Program.

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

i 3.7/4.7 C0lrrAT13elrr SYSTEMS FEB 031988 LIMifINC wnuITIONS FOR OPF# ATION SURVEIT.T AMCE REOUIh d 4.7.A.

Primary Contal====t 4.7.A.2. (Cont'd) b.

If any periodic et <}

type A test fails to 0.75 L.,

the test schedule for subse ant type A tests shal be reviewed and ap oved by the Commissi If two cons cutive type A tests fai to meet 0.75 L, a type A test shall performed at least every 18 months unt two consecutive t e A testa meet

.75 L,, at which time tne above test schedule f

may be resuned.

c.

1.

Test D ion'shal

~

be at least 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

2.

A 4-hour stabilization riod will be requ ed and the conta t

atmosphere 11 be consid stabilized when change in weight average air temps ature averaged ove an hour does not de ate by more than 0 'R/ hour from the verage rate of change of temperature measured from the previous 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

EII2 BFN 3.7/4.7_4 Unit 3

FEB 031988 3.7/4.7 CONTAINMENT SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 4.7.A.

Primary Containment 4.7.A.2. (Cont'd) m

d. 1. At least 20 sets of dat points at approximate l

equal time interva and in no case at inter is greater than o hour shall be provided or proper statisti.

analysis.

2. The f ure of merit for the in rumentation system 11 never exceed 0.25 L -

a

e. The test shal o

conclude an increasing

,ssa ated 1eak rate.

)

~

f. The accuracy of each type test shall be verified by supplemental test which:

~

1. Confirms the accura y of the test by verify ng that the difference be ween the supplemental dat and the d ^

type A test dat is within 0.25 L -

a

2. Has duratio sufficient to 2

establish curately the l

change in eskase rate between e type A test and the sup emental test.

3. Requi s the quantity of gas jected into the con inment or bled from th containment during the s plemental test to be uivalent to at least 2

5 percent of the total measured leakage at Pa (49.6 psig).

BFN 3.7/4.7-5 Unit 3 AMENDMENT NO. I 12 1

i

--"9

~_

. - ~..

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

3.7/4.7 CONTAINNENT SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 4.7.A.

primary Containment 4.7.A.2. (Cont'd)

g. Local leak rate tests (LLRT )

t shall be perfocined on' the primary containment testab e penetrations and isolatio valves, which are not pa of a water-sealed system, a not less than 49.6 psig (ex pt for the main steam isol tion valves, see 4.7.A.2.1) and not less than 54.6 psis f water-sealed valves e ch operating cycle. Bo ed double-gasketed seal shall be tested whenever the seal is

')

closed after being pened and i

at least once per perating cycle. Acceptabl methods of testing are halid gas detection, soap

bbles, pressure decay, hydrostatically pressurized fluid flow or uivalent.

The personne air lock shall be tested at -month intervals at an intern 1 pressure of not less than 4.6 psis.

In

addition, the personnel air lock is op ned during periods when cont nment integrity is not requi ed, a test at the end of s ch a period will be conduct at not less than 49.6 ps g.

If the personnel air lo is opened during a period when contairanent intog ity is required, a test at 1

.5 psig shall be cond eted within 3 days after bei opened.

If the air lock is pened more frequently than on e every 3 days, the air 1 k shall be tested at least o ce every 3 days during the eriod of frequent openings.

4 BFN 3.7/4.7-6 Unit 3

3.7/4.7 CorrATEMElff sIsir.rs NDY 1619KL Lin111mW C6HLITIONS FOR OPunAv10N SURvrTT.T.AMCE REQUTRDIElffs 4.7.A.

Priman Containment 4.7.A.2.s (Cont'd)

The total leakage from al penetrations and isolati valves shall not exceed 60 peret.nt of La Per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Leakage f contafament isolati valves that terminate bel suppression pool atar level may be excluded rom the total leakage ovided a sufficient id inventory is available ensure the sealing f tion for at least 30 days a a pressure of 54.6 psi. Leakage from contal eut isolation valves that e in closed-loop, sei c class I lines that wil be water sealed during a D

will be measured but will b excluded when computing he total leakage.

~

Tn.r t 0

Y NO.161 3 7/4 7-7 t3

INSERT B PARAGRAPH 4.7.A.2.g Perform required local leak rate tests, including the primary containment air lock leakage rate testing in accordance with the Primary Containment Leakage Rate Testing Program.

I Note:

An inoperable air lock does not invalidate the previous successful performance of the overall air lock leakage test.

The acceptance criteria for air lock testing are:

(1) Overall air lock leakage rate is s (0.05 L ) ir lock leakage rate is when tested at 2 Pa.

(2)

For door seal leakage, the overall a s (0. 02 L.) when the air lock is pressurized to (2 2.5 psig for at least 15 minutes).

N0Y.161992 3.7/4.7 BMZ1 i

3.7.A & 4.7.A Pri==rv Contain==nt The integrity of.the primary containment and operation of the core standby cooling system in combination, ensure 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 conjunction with the leakage rate limitation, will limit the site boundary radiation doses to within the*

limits of 10 CFR Part 100 during accident conditions.

During initial core loading and while the low power test program is being conducted and ready access to the reactor vessel is required, there will be no pressure on the system thus greatly reducing the chances of a pipe break. The reactor may be taken critical during this period; however, restrictive operating procedures will be in effect to minimize the probability of an accident occurring.

The limitations on primary 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 49.6 pais, 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 measur$agyleakage rat== m c r im.4.idi t _he reaufr - *- Of f.;; Zia.i or 10 wa Part 50 (type A, B, and C tests).

The pressure suppression pool water provides the heat sink for the reactor primary system energy release following a postulated rupture of the system. The pressure suppression chamber water volume must absorb the associated decay and structural sensible heat release during primary system blowdown from 1,035 pais. Since all of the gases in the drywell are purged into the pressure suppression chamber air space during a loss of coolant accident, the pressure resulting from isothermal compression plus the vapor pressure of the liquid must not exceed 62 pais, the suppression chamber maximus pressure. The design volume of the suppression chamber (water and air) was obtained by considering that the total volume of reactor coolant to be condensed is discharged to the suppression chamber and that the drywell volume is purged to the suppression chamber.

Using the minimum or maximum water levels given in the specification, containment pressure during the design basis accident is approximately 49 psig, which is below the maximum of 62 psig. The maximum water level indications of -1 inch corresponds to a downconer submergence of three feet seven inches and a vett.r volume of 127,800 cubic feet with or 128,700 cubic feet without the dryvell-suppression chamber differential pressure control. The minimum water level indication of -6.25 inches with differential pressure control and -7.25 inches without differential pressure control corresponds to a downconer submergence of approximately three feet and water volume of approximately 123,000 cubic feet.

BFN 3.7/4.7-24 EE R NO.I61 Unit 3

. ~..

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

j-j DEC 0 21993 3

l than 8 days in gaseous effluents released from each unit to

]

f areas beyond the 5fTB-BOUNDARY conforming to Appendix I to j

10 CFR Part 50.

I J.

Limitations on the annual dose or dose commitment to any MEMBER OF THE PUBLIC due to releases of radioactivity and to i

radiation from uranium fuel cycle sources conforming to 40 l

CFR Part 190.

I 6.8.4.2 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM A program shall be provided to monitor the radiation and radionuclides in the environs of the plant. The program shall provide (1) representative measurements of radioactivity in the highest potential exposure pathways, and (2) verification of the accuracy of the effluent monitoring program and modeling of environmental exposure pathways. The program shall (1) be cuuLained in the ODCM, (2) conform to the guidance of Apper.di: I to 10 CFR Part 50, and (3) include the following i

l a.

Monitoring, sampling, analysis, and reporting of radiation and radionuclides in the environment in accordance with the methodology and parameters in the ODCM, i

l b.

A Land Use Census to ensure that changes in the use of area at and beyond the SITE BOUNDARY are identified and that modifications to the monitoring program are made if required by the results of this census, and c.

Participation in an Interlaboratory Comparison Program to ensure that independent checks on the precision and accuracy of the measurements of radioactive materials in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring.

N lp Hef (

BFN 6.0-23cl AMENDMENT NO. I 74 Unit 3

INSERT C Section 6.8.4.3 PRIMARY CONTAINMENT LEAKAGE RATE TESTING PROGRAM l

A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix-J, Option B, as modified by approved exemptions.

This program shall be in accordance with the l

guidelines contained in Regulatory Guide 1.163,

" Performance-Based Containment Leak-Test program, dated September 1995".

The peak calculated containment internal pressure for the design basis loss of coolant accident, P.,

is 49.6 psig.

The maximum allowable primary containment leakage rate, L,

at P.,

shall be 2% of primary containment air weight per day.

Leakage Rate acceptance criteria are:

l a.

Primary Containment leakage rate acceptance criterion is y

s 1. 0 L.

During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are s 0.60 L, for the Type B and Type C tests and 5 0.75 L, for Type A tests; b.

Air lock testing acceptance criteria are:

1)

Overall air lock leakage rate is s 0.05 L, when tested at 2 P,

2)

Air lock door seals leakage rate is s 0.02 L, when the overall air lock is pressurized to 2 2.5 psig for at I

least 15 minutes.

i l

~

ENCLOSURE 3 1

TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)

UNITS 1, 2, AND 3 PROPOSED TECHNICAL SPECIFICATION (TS) CHANGE T8-364 REVISED PAGES I.

AFFECTED PAGE LIST Unit 1 Unit 2 Unit 3 i

3.7/4.7-3 3.7/4.7-3 3.7/4.7-3 j

3.7/4.7-4 3.7/4.7-4 3.7/4.7-4 3.7/4.7-5 3.7/4.7-5 3.7/4.7-5 3.7/4.7-6 3.7/4.7-6 3.7/4.7-6 3.7/4.7-7 3.7/4.7-7 3.7/4.7-7 3.7/4.7-25 3.7/4.7-25 3.7/4.7-24 6.0-23d 6.0-23c 6.0-23d 6.0-23e 6.0-23d 6.0-23e 6.0-23e II.

REVISED PAGES See attached.

l

1 3.7/4.7 CONTAIEMENT SYSTEMS LIMITING CONDITIONS FOR OPRDATION SURVEIT.T.ANCE REOUIREhinis i

3.7.A.

Primary Containment 4.7.A.

Primary Containment 2.a.

Primary containment

2. Intearated Leak Rate Testina integrity shall be maintained at all times Primary containment nitrogen when the reactor is critical consumption shall be or when the reactor water monitored to determine the temperature is above 212*F average daily nitrogen and fuel is in the reactor consumption for the last vessel except while 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Excessive leakage performing "open vessel" is indicated by a N2 I

physics tests at power consumption rate of > 2% of l

1evels not to exceed the primary containment free 5 MW(t).

volume per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

(corrected for drywell

b. Primary containment temperature, pressure, and integrity.is confirmed if venting operations) at the maximum allowable 49.6 peig. Corrected to integrated leakage rate, normal drywell operating L., does not exceed the pressure of 1.1 peig, this equivalent of 2 percent of value is 542 SCFH.

If this J

the primary containment value is exceeded, the i

volume per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at the action specified in i

49.6 pais design basis 3.7.A.2.C shall be taken.

accident pressure, P,.

]

c. If N2 makeup to the primary in accordance with the Primary containment averaged over Containment Leakage Rate i

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (corrected for

-Testing Program, i

pressure, temperature, and venting operations) exceeds 542 SCFH, it must be reduced to < 542 SCFH within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or the reactor shall be placed in Hot Shutdown within the next 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months .

l BFN 3.7/4.7-3 Unit 1 1

_ - ~

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

3.7/4.7 CONTAllWENT SYSTEMS LIMITING CONDITIONS FOR'0PRDATION SURVEIT.T.ANCE REOUImmr.n15 4.7.A.

Primary Containment 4.7.A.2. (Cont'd) b.

Deleted d

c.

Deleted BFN 3.7/4.7-4 Unit 1

. ~..

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

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

I 3.7/4.7 CONTAIMMENT SYSTEMS i

LIMITING CONDITIONS FOR OPRDATION SURVEIT.T.ANCE REOUIREMENTS t

4.7.A.

Primary Contain= ant 3

4.7.A.2. (Cont'd)

I

d. Deleted d

i

e. Deleted d

l j

f. Deleted i

I 4

i t

+

)

4 i

i 1

i j

BFN 3.7/4.7-5

)

Unit 1

3.7/4.7 CONTAINMENT SYSTEMS j

i LIMITING CONDITIONS FOR OPERATION SURVEIT.T ANCE REOUIREMENTS i

4.7.A.

Primary Containment 4

j 4.7.A.2. (Cont'd) g.

Perform required local leak rate tests, including the primary containment air lock leakage rate testing in accordance with the Primary Containment Leakage Rate Testing Program.

I Note: An inoperable air lock j

does not invalidate the previous succewaful performance of the 1

overall air lock 1eakage test.

The acceptance criteria for air lock testing are:

(1)

Overall air lock leakage rate j

is 1 (0.05 L.) when tested at 1 Pa.

(2) For door seal leakage, the overall air lock leakage rate is 1 (0.02 L.)

when the air lock is pressurized to (1 2.5 pais for at least 15 minutes).

BFN 3.7/4.7-6 Unit 1

a a

s.

THIS PAGE INTENTIONALLY LEFT BLANK l

1 1

I l

i BFN 3,7/4,7_7 Unit 1 l

i l

3.7/4.7 RASES 3.7.A & 4.7.A Primary Containment i

l The integrity of the primary containment and operation of the core standby cooling system in combination, ensure 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 conjunction with the leakage rate limitation, will limit the site boundary radiation doses to within the limits of 10 CFR Part 100 during accident conditions.

l During initial core loading and while the low power test program is being conducted and ready access to the reactor vessel is required, there will be no pressure on the system thus greatly reducing the chances of a pipe break. The reactor may be taken critical during this period; however, restrictive operating procedures will be in effect to minimize the probability of an accident occurring.

The limitations on primary 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 49.6 pais, 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 l

between leakage tests.

1 i

i The pressure suppression pool water provides the heat sink for the reactor primary system energy release following a postulated rupture of the system. The pressure suppression chamber water volume must absorb the associated decay and structural sensible heat release during primary system blowdown from 1,035 psig. Since all of the gases in the drywell i

are purged into the pressure suppression chamber air space during a loss

{

l of coolant accident, the pressure resulting from isothermal compression

- l plus the vapor pressure of the liquid must not exceed 62 psis, the suppression chamber maximum pressure. The design volume of the suppression chamber (water and air) was obtained by considering that the i

total volume of reactor coolant to be condensed is discharged to the suppression chamber and that the drywell volume is purged to the suppression chamber.

Using the minimum or maximum water levels given in the specification, containment pressure during the design basis accident is approximately 49 psig, which is below the maximum of 62 psig. The maximum water level l

indications of -1 inch corresponds to a downcomer submergence of three feet seven inches and a water volume of 127,800 cubic feet with or 128,700 cubic feet without the drywell-suppression chamber differential pressure control. The minimum water level indication of -6.25 inches with differential pressure control and -7.25 inches without differential pressure control corresponds to a downcomer submergence of approximately three feet and water volume of approximately 123,000 cubic feet.

BFN 3.7/4.7-25 Unit 1 i

l'

accuracy of.the measurements of radioactive materials in i

environmental sample matrices are performed as part of the quality assurance program for environmental monitoring.

6.8.4.3 PRIMARY CONTAIlOENT LEAKAGE RATE TESTING PROGRAM A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B,'as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163,

" Performance-Based Containment Leak-Test program, dated September 1995".

The peak calculated containment internal pressure for the design basis loss of coolant accid <ent, P, is 49.6 pais.

The maximum allowable primary containment leakage rate, L, at a

P., shall be 2% of primary containment air weight per day.

Leakage Rate acceptance criteria are:

a.

Primary Containment leakage rate acceptance criterion in 1 1.0 L. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are 1 0.60 L for the Type B and Type C tests and a

1 0.75 L for Type A tests; a

i b.

Air lor'.tsting acceptance criteria are:

(1) Overall air lock leakage rate is 1 0.05 L, when tested at 1 P.,

1 (2) Air lock door seals leakage rate is 1 0.02 L when the a

i overall air lock is pressurized to 1 2.5 pais for at j

1 east 15 minutes.

i BFK 6.0-23d Unit 1 i

l THIS PAGE INTENTIONALLY LEFT BLANK l

l i

l BFN 6.0-23e Unit 1

1 6.8.5 PROGRAMS Postaccident S==nlina Postaccident sampling activities will ensure the capability to obtain and analyze reactor coolant, radioactive iodines and particulates in plant gaseous effluents, and containment atmosphere samples under accident conditions. These activities shall include the following:

(i)

Training of personnel, (ii) Procedures for sampling and analysis, (iii) Provisions for maintenance of sampling and ana. lysis.

6.9 REPORTING REOUIREMENTS ROUTINE REPORTS 1

)

6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following identified reports shal: be submitted to the Director of the Regional Office of NRC, unless otherwise noted.

6.9.1.1 STARTUP REPORT a.

A summary report of plant startup and power escalation testing shall be submitted following (1) receipt of an operating license, (2) amendment to the license involving a planned increase in power level, (3) installation of fuel

~

that has a different design or has been manufactured by a l

l l

l BFN 6.0-24 Unit 1

3.7/4.7 CONTAIlOENT SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEIT.T.ANCE REOUIREMENTS 3.7.A.

Primary Containment 4.7.A.

Primary Containment 2.a.

Primary containment 2.

Intearated Leak Rate Testina integrity shall be maintained at all times Primary containment nitrogen when the reactor is critical consumption shall be or when the reactor water monitored to determine the temperature is above 212*F average daily nitrogen and fuel is in the reactor consumption for the last vessel except while 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Excessive leakage.

performing "open vessel" is indicated by a N2 physics tests at power consumption rate of > 2% of levels not to exceed the primary containment free 5 MW(t).

volume per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (corrected for drywell b.

Primary containment temperature, pressure, and integrity is confirmed if venting operations) at the maximum allowable

'49.6 pais. Corrected to integrated leakage rate, normal drywell operating L., does not exceed the pressure of 1.1 peig, this equivalent of 2 percent of value is 542 SCFH.

If this the primary containment value is exceeded, the volume per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at the action specified in 49.6 pais design basis 3.7.A.2.C shall be taken.

accident pressure, P.

Perform leakage rate testing c.

If N2 makeup to the primary in accordance with the Primary containment averaged over Containment Leakage Rate 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (corrected for Testing Program.

pressure, temperature, and venting operations) exceeds 542 SCFH, it must be reduced to < 542 SCFH within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or the reactor shall be l

placed in Hot Shutdown within the next 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months .

BFN 3.7/4.7-3 Unit 2

3.7/4.7 CONTAIIDENT Sisir.ns LIMITING CONDITIONS FOR OPERATION SURVEIT T ANCE REOUIREMENTS 4.7.A.

Primary Containment 4.7.A.2. (Cont'd) b.

Deleted

]

q c.

Deleted BFN 3.7/4,7_4 Unit 2

m m

3.7/4.7 CONTAINMENT SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEITTAMCE REOUIREMENTS 4.7.A.

Primary Containment 1

4.7.A.2. (Cont'd) j

d. Deleted 1

]

e. Deleted

f. Deleted i

l l

1 I

i l

BFN 3.7/4.7-5 l

Unit 2

_ _ -... _ _.. =

3.7/4.7 CONTATIDENT SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEIT.T.AMCE REOUIREiun15 4.7 A.

Primary Contain==nt 4.7.A.2. (Cont'd) i s.

Perform required local leak i

rate tests, including the primary containment air lock leakage rate testing in accordance with the Primary Containment Leakage Rate Testing Program.

I Note: An inoperable air lock does not invalidate the previous successful j

performance of the

{

overall air lock i

leakage test.

The acceptance criteria for air lock testing are:

(1)

Overall air lock leakage rate is 1 (0.05 L ) when tested at a

1 Pa.

(2) For door seal leakage, the overall air lock leakage rate is 1 (0.02 L.)

when the air lock is pressurized to (1 2.5 peig for at least 15 minutes).

l 1

BFN 3.7/4.7-6 Unit 2 i

1 1

)

THIS PAGE INTENTIONALLY LEFT BLANK i

l l

l I

i i

i I

4 i

t BFN 3.7/4.7-7 Unit 2

3.7/4.7 RASES 3.7.A & 4.7.A Primary Containment The integrity of the primary containment and operation of the core standby cooling system in combination, ensure 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 conjunction with the leakage rate limitation, will limit-the site boundary radiation doses to within the limits of 10 CFR Part 100 during accident conditions.

During initial core loading and while the low power test program is being conducted and ready access to the reactor vessel is required, there will be no pressure on the system thus greatly reducing the chances of a pipe break. The reactor may be taken critical during this period; however, restrictive operating procedures will be in effect to minimize the probability of an accident occurring.

The limitations on primary 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 49.6 pois, 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 bcrriers between leakage tests.

The pressure suppression pool water provides the heat sink for the reactor primary system energy release following a' postulated rupture of the system. The pressure suppression chamber water volume must absorb the associated decay and structural sensible heat release during primary system blowdown from 1,035 psig. Since all of the gases in the drywell are purged into the pressure suppression chamber air space during a loss of coolant accident, the pressure resulting from isothermal compression plus the vapor pressure of the liquid must not exceed 62 psig, the suppression chamber maximum pressure. The design volume of the suppression chamber (water and air) was obtained by considering that the total volume of reactor coolant to be condensed is discharged to the suppression chamber and that the drywell volume is purged to the i

suppression chamber.

Using the minimum or maximum water levels given in the specification, containment pressure during the design basis accident is approximately 49 pais, which is below the maximum of 62 psig. The maximum water level indications of -1 inch corresponds to a downcomer submergence of three feet seven inches and a water volume of 127,800 cubic feet with or 128,700 cubic feet without the drywell-suppression chamber differential pressure control. The minimum water level indication of -6.25 inches with differential pressure control and -7.25 inches without differential pressure control corresponds to a downconer submergence of approximately three feet and a water volume of approximately 123,000 cubic feet.

i BFN 3.7/4.7-25 Unit 2

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

j. Limitations on the annual dose or dose commitment to any MEMBER OF THE PUBLIC due to releases of radioactivity and to i

radiation from uranium fuel cycle sources conforming to.40 CFR Part 190.

I 6.8.4.2 RADIOLOGICAL ENVIROIDENTAL MONITORING PROGRAM A program shall be provided to monitor the radiation and

]

radionuclides in the environs of the plant. The program shall provide (1) representative measurements of radioactivity in the highest potential exposure pathways, and (2) verification of the accuracy of the effluent monitoring program and modeling of j

environmental exposure pathways. The program shall (1)_be i

contained in the ODCM, (2) conform to the guidance of Appendix I to 10 CFR Part 50, and (3) include the following:

a.

Monitoring, sampling, analysis, and reporting of radiation and radionuclides in the environment in accordance with the i

methodology and parameters in the ODCM, b.

A Land Use Census to ensure that changes in the use of area at and beyond the SITE BOUNDARY are identified and that modifications to the monitoring program are made if required by the results of this census, and c.

Participation in an Interlaboratory Comparison Program to ensure that independent checks on the precision and accuracy of the measurements of radioactive materials in environmental sample matrices are performed as part of the j

quality assurance program for environmental monitoring.

6.8.4.3 PRIMARY CONTAINMENT LEAKAGE RATE TESTING PROGRAM i

i A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved BFN 6.0-23c Unit 2

exemptions. 'This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163,

" Performance-Based Containment Leak-Test program, dated September 1995".

The peak calculated containment internal pressure for the design basis loss of coolant accident, P, is 49.6 psig.

The maximum allowable primary containment leakage rate, L, at P, shall be 2% of primary containment air weight per day.

Leakage Rate acceptance criteria are:

a.

Primary Containment leakage rate acceptance criterion is 1 1.0 L. During the first unit startup following testing a

in accordance with this program, the leakage rate acceptance criteria are 1 0.60 L for the Type B and Type C tests and a

1 0.75 L for Type A tests; a

b.. Air lock testing acceptance criteria are:

(1) Overall air lock leakage rate is 1 0.05 L, when tested

-l at 1 P,

(2) Air lock door seals leakage rate is 1 0.02 L when the j

a overall air lock is pressurized to 1 2.5 psig for at least 15 minutes.

6.8.5 PROGRAMS Postaccident Sampling Postaccident sampling activities will ensure the capability to obtain and analyze reactor coolant, radioactive iodines and BFN 6.0-23d Unit 2

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i BFN 6.0-23e Unit 2

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4 3.7/4.7 CONTAINMENT SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEIT.T.ANCE REOUIREMENTS 3.7.A.

Primary Containment 4.7.A.

Primary Containment l-2.a.

Primary containment 2.

Intearated Leak Rate Testina integrity shall be l

maintained at all times Primary containment nitrogen j_

when the reactor is critical consumption shall be or when the reactor water monitored to determine the temperature is above 212*F average daily nitrogen and fuel is in the reactor consumption for the last i

vessel except while 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Excessive leakage l

performing "open vessel" is indicated by a N2 physics tests at power consumption rate of > 2% of levels not to exceed the primary containment free 5 MW(t).

volume per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months j

(corrected for drywell j

b. Primary containment temperature, pressure, and integrity is confirmed if venting operations) at the maximum allowable 49.6 psig. Corrected to l

integrated leakage rate, normal drywell operating l

L, does not exceed the pressure of 1.1 psig, this f

j equivalent of 2 percent of value is 542 SCFH.

If this the primary containment value is exceeded, the volume per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at the action specified in 49.6 psig design basis 3.7.A.2.c shall be taken.

1 l

accident pressure, P.

a Perform leakage rate testing 1

c. If N2 makeup to the primary in accordance with the Primary containment averaged over Containment Leakage Rate l

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (corrected for Testing Program.

pressure, temperature, and e

venting operations) exceeds 542 SCFH, it must be reduced i

to < 542 SCFH within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months l

or the reactor shall be I

placed in Hot Shutdown within the next 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months .

i 1

l i

i i

4 i

i BFN 3.7/4.7-3 i

Unit 3

=-

. ~.......

l l

3.7/4.7 CONTAINMENT SYSTEMS LIMITING CONDITIONS Foi OPRDATION SURVEIffANCE REOUIREMENTS 4.7.A.

Primary Containment 4.7.A.2. (Cont'd) d b.

Deleted q

c.

D.1.t.d l

i i

BFN 3.7/4.7-4 Unit 3

3.7/4.7 CONTAI1WENT SYSTEMS LIMITING CONDITIONS FOR OPRDATION SURVEIT.T.ANCE REOUIREMENTS 4.7.A.

Primary Containment 4.7.A.2. (Cont'd) d

d. Deleted

e. De1eted 3

f. Deleted d

i l

i i

BFN 3.7/4.7-5 j

Unit 3

-... -... ~

e 3.7/4.7 CONTAIlOENT SYSTEMS LIMITING CONDITIONS FOR OPRRATION SURVEIf7AMCE REQUIREMENTS 4.7.A.

Primary Containment 4.7.A.2. (Cont'd) 3 Perform required local leak rate tests, including the primary containment air lock leakage rate testing in i

accordance with the Primary i

Containment Leakage Rate Testing Program.

Note: An inoperable air lock j

does not invalidate the previous successful performance of the overall air lock leakage test.

The acceptance criteria for air lock testing are:

(1)

Overall sir lock leakage rate is 1 (0.05 L.) when tested at 1 Pa.

(2) For door seal leakage, the overall air lock leakage rate is 1 (0.02 L )

when the air lock is pressurized to (1 2.5 psig for at least 15 minutes).

BFN 3.7/4.7-6 Unit 3 i

e y

<r

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w

=

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l BFN 3.7/4.7-7 Unit 3

3.7/4.7 BASES 3.7.A & 4.7.A Primary Containment l

The integrity of the primary containment and operation of the core standby cooling system in combination, ensure that the release of j

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 conjunction with the leakage rate limitation, will limit the site boundary radiation doses to within the limits of 10 CFR Part 100 during accident conditions.

During initial core loading and while the low power test program is being conducted and ready access to the reactor vessel is required, there will be no pressure on the system thus greatly reducing the chances of a pipe break. The reactor may be taken critical during this period; however,

)

restrictive operating procedures will be in effect to minimize the probability of an accident occurring.

The limitations on primary 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 49.6 psig, 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 pressure suppression pool water provides the heat sink for the reactor primary system energy release following a postulated rupture of the system. The pressure suppression chamber water volume must absorb the associated decay and structural sensible heat release during primary system blowdown from 1,035 psig.

Since all of the gases in the drywell are purged into the pressure suppression chamber air space during a loss of coolant accident, the pressure resulting from isothermal compression plus the vapor pressure of the liquid must not exceed 62 psig, the suppression chamber maximum pressure. The design volume of the suppression chamber (water and air) was obtained by considering that the total volume of reactor coolant to be condensed is discharged to the suppression chamber and that the drywell volume is purged to the suppression chamber.

Using the minimum or maximum water levels given in the specification, containment pressure during the design basis accident is approximately 49 psig, which is below the maximum of 62 psig. The maximum water level indications of -1 inch corresponds to a downcomer submergence of three feet seven inches and a water volume of 127,800 cubic feet with or 128,700 cubic feet without the drywell-suppression chamber differential pressure control. The minimum water level indication of -6.25 inches with differential pressure control and -7.25 inches without differential pressure control corresponds to a downcomer submergence of approximately three feet and water volume of approximately 123,000 cubic feet.

BFN 3.7/4.7-24 Unit 3

~ _ _.,

1 1

1 6.8.4.3 PRIMARY CONTAIl0ENT LEAKAGE RATE TESTING PROGRAM j

A program shall be established to implement the leakage rate l

testing of the containment as required by 10 CFR 50.54(o) and 10 i

CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163,

" Performance-Based Containment Leak-Test program, dated September 1995".

)

The peak calculated containment internal pressure for the design basis loss of coolant accident, P, is 49.6 psig.

j a

l The maximum allowable primary containment leakage rate, L.,

at P.,

shall be 2% of primary containment air weight per day.

Leakage Rate acceptance criteria :Ta:

a.

Primary Containment leakage rate acceptance criterion is 1 1.0 L. During the first unit startup following testing ia accordance with this program, the leakage rate acceptance criteria are 1 0.60 L for the Type B and Type C tests and a

1 0.75 L for Type A tests; a

b.

Air lock testing acceptance criteria are:

(1) Overall air lock leakage rate is 1 0.05 L,when tested at 1 P,

(2) Air lock door seals leakage rate is 1 0.02 L when the a

overall air lock is pressurized to 1 2.5 psig for at least 15 minutes.

BFN 6.0-23d Unit 3

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BFN 6.0-23e Unit 3

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ML20095E413

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