Text

Proposed Tech Specs Re Implementation of 10CFR50,App J, Option B,Performance Based Testing
	ML20096C440
Person / Time
Site:	Browns Ferry
Issue date:	01/10/1996
From:	; TENNESSEE VALLEY AUTHORITY
To:	;
Shared Package
ML20096C435	List: ML20096C434; ML20096C440;
References
	NUDOCS 9601170296
	Download: ML20096C440 (60)
	v • d • e

_

t i

ENCLO!URE 1 TENNESSEE VALLEY AUTHORITY BROWN 8 FERRY NUCLEAR PLANT (BFN)

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

'i 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.

4 1

)

9'6'01170296 960110 DR ADOCK 05000259 PDR

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

3.7/h.7 00NTAINMENT SYSTmS DEC 0 71994 FORVEI' ANCE REOUIREMENTS LIMITING CONDITIONS FOR OPERATION 4.7.A.

Pr4= mrv Containment 3.7.A.

Pr4 mary Conta4nment

2. Iritemted Lenk Rate Testine 2.a.

Primary containment integrity shall be maintained at all times Primary containament nitrogen

)

when the reactor is critical consumption shall be monitored to determine the or when the reactor water temperature is above 212*T average daily nitrogen and fuel is in the reactor consumption for the last 26 hours3.009259e-4 days 0.00722 hours 4.298942e-5 weeks 9.893e-6 months .

Excessive leakage vessel except while is indicated by a N2 performing "open vessel" consumption rate of > 21 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 temperature, pressure, and

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

If this equ value is exceeded, the l

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,.

The containment leakage r 4

shall be demonstrated the i

c. If N2 makeup to the primary containment averaged over following test e ule and 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (corrected for shall be det ed in accordane th Appendix J

)

pressure, temperature, and to 10 50 as modified venting operations) exceeds 542 SCFB, it must be reduced b

pyroved exemptions.

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

ree ty ta (overall integrated placed in Hot Shutdown containment i 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 40 i 10 intervals i

during s s down at P,,

49.6 g, during each 10 ar plant inservice spection.

.Tn.w/ A AMENDMDIT E 313 3.7/4.7-3 BFN Unit 1

i INSERT A l

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

}

l 1

4 k

I 4

3.7/4.7 COW AT19 TENT SYSTEMS SURVEft.flMCE REOUIki mni5 LiruriNG CONDITIONS FOR OPERATION 4.7.A.

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

If any periodic I

type A test fails t meet 0.75 L., the tiest schedule for quant type A tests 1 be reviewed and a roved by the Commissi If two e 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 testa meet 0 5 L, at which time he above test schedule may be resumed.

j c.

l '. Test duration shall 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 I

the conta t

atmosphor will be consid stabilized when change in wei ed avertse air t

reture averaged ov an hour does not viste by more than

.5'R/ hour from the average rate of change of temperature

/

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

EEE I4 I 3.7/4.7-4 BFR Unit 1

l FEB 0319 3.7/4.7 CONTAINMENT SYSTEMS 3

2 j

LIMITING CONDITIONS FOR OPERATION

SURVEILLANCE REQUIREMENTS i

4.7.A.

Primary Containment l

4.7.A.2. (Cont'd) r, I

d. 1. At least 26 sets of dat points at approicinate l

equal time interva and in no case at inte is i

greater the hour shall j

be provid or proper statistic analysis.

2. The sure of merit for the I

i rumentation system 11 never exceed j

j 0.25 L.

i

e. The test s' hall not be ___

l concluded ing i

a ed leak rate.

~

h

f. The ace'uracy of each' type test shall be verified by a supplemental test which:

l'. Confirms the accura y of the test by verify ng that i

the difference be the I

supplemental dat and the type A test dat is within 0.25 La*

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 gas i ected into the cont nment or bled from the ontainment during the su lesental test to be-ivalent to at least percent of the total red leakage at Pa

.}

(49.6 psig).

[E.j BFN 3.7.4.7-5 Utd t 1 AMENDMENT NO.141 6

l 1

i 3.7/4.7 COWv1T a n T SYSTEMS i

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 4.7.A.

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

{

g. Local' leak rate tests (LLRT )

shall be.perforeed en the i

l primary containment testa e

~

penetrations and isolati valves, which are not pa of a water-sealed system, a not less than 49.6 psis (

ept for the main steam iso tion 4

valves, see 4.7.A.2.1) and not less than 54.6 psis f r i

water-sealed valves ch j.

i operating cycle. Bo ted double-gasketed sea a shall be tested whenever 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 ivalent.

The personne air lock shall be tested a 6-month intervals at an into 1 pressure of not less than 9.6 pais. In

addition, f the personnel air lock is o ened during periois when' con ainment integrity is not req red, a test at the end of ch a period will be conduc ed at not less than 49.6 is.

If the personnel air 1 ek is opened during a peri when containment int city is required, a test at 2.5 pois shall be e

eted within 3 days after ng opened. If the air lock i 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.

N BFN 3.7/4.7-6 Unit 1 i,

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

3.7/4.7 cmrrifietruT s1sie g

4 1

1.mmsw uu-uTTIONS FOR OPr#Avf 0N suuv1.11T.ANCE mzuui - is l

4.7.A.

Pr4= mrv contain= nt l

4.7.A.2.3 (Cont'd)

=_

i The total leakage from a i

penetrations and isolat a valves shall not exce 60 j

percent of La per 24-urs.

i Leakage from conta t

j isolation valves t

I terminate below rossion pool water level y be l

excluded from e total leakage pro a sufficient fluid inv is available to ensure sealing function f at least 30 days at a pres te of 54.6 pais.

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

N 5/

AMSDMNT NO. f. 59 BFN 3.7/4.7_7 Unit 1

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.

1 Note:

An inoperable air lock door 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.) when tested at 2 Pa.

(2)

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

IbE 3.7/4.7 BASES 3.7.A & 4.7.A Pr4 mary containment N 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 chose leakage paths and associated leak rates aestuned in the accident saalyses. 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.

N 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 peig, P. As an a

added conservatism, the measured overall integrated leakass 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.

cn The surveillance tacting for mes.u--@ 1e=6==e rates ars = = 4 -t -a rii

::M. --- ;. or,

tx a or 10 CFR Part 50 (type A, B, and C tests).

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

a l purged into the pressure suppression chamber air space during a 2oss of coolant accident, the pressure resulting from isothermal compression plus the vapor pressure of the liquid must not exceed 62 pais, the suppression chamber =mwi=== 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 suppressioerchamber and that the dryvell volume is purged to the suppression chamber.

Using the minimum or parimum water levels given in the specification, containment pressure during the design basis accident is appravimately 49 peig, which is below the marian = of 62 peig. The maximum water level indications of -1 inch corresponds to a downconer submergence of three feet seven inches and a water volume of 127,800 cubic feet with or 128,700 cubic feet without the dryvell-suppression chamber differential pressure control. N minh== 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 approximat.ely three feet and water voltane of appravimately 123,000 cubic feet.

l AM DIDM DtiIlp. I 8 9 Unit 1

i 1

JW 211994 I

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

l f n yrr h C.

--y l

6.8.5 PROGRAMS i,

a Postaccident Samelins f

f Postaccident sampling activities will ensure the capability to J

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 (i)

Training of personnel, f

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

6.9 REPORTING REOUIREMEETS 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 summary report of plant startup and power escalation a.

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 i

I that has a different design or has been manufactured by a BFN 6.0-24 AMNDMNT110. 2 0 7 Unit 1

1 j

i 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.

Leaksge 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 s 0.60 L, for the Type B and Type C tests and s 0.75 L, for Type A tests; l

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.

n

---,e 4,

g(( Q 71994 3.7/4.7 CONTAINMENT SYSTEMS SURVEII1ANCE REOUIREMENTS LIMITING CONDITIONS FOR OPERATION 4.7.A.

Primary Containment Pr mary Containment i

3.7.A.

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 resetor water temperature is above 212*T average daily nitrogen and fuel is in the reactor consumption for the last vessel except while I

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Excessive leakage is indicated by a N2 l

performing "open vessel" consumption rate of > 2% of J

physics tests at power 1evels not to exceed the primary containment f ree

)

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

(corrected for drywell temperature, pressure, and b.

Primary containment integrity is confirmed if venting operations) at the maximum allowable 49.6 psig. Corrected to normal 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 49.6 psig design basis 3.7.A.2.C shall be taken.

accident pressure, P,.

]

e conta'inment leakag[e r a

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 accordan ith Appendix J to pressure, temperature, and venting operations) exceeds 10 0 0 as modified by roved exemptions.

542 SCFB, it must be reduced 1"_

to < 542 SCFH within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months red 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 10-month interval uring shutdown at P 9.6 psig, during 10-year plant inservice inspection.

~ TAJfr BFN 3.7/4.7-3 AMSDMGT NO. 2 29 Unit 2

1 INSERT A T8 4.7.A.2 8ECOND PARAGRAPH l

Perform leakage rate testing in accordance with the Primary containment Leakage Rate Testing Program.

l 4

a 1

~--

.. _. ~.......... ~..

i B 03 mi j

3.7/4.7 C01rf'Ainemni SYSTEMS StTRVEff 7 AMCE REQUIREMENTS 1.ihiliNC CONDITIONS FOR OPEtiTION I

i 4.7.A.

Primary Containment

^

i 4.7.A.2. (Cont'd) l b.

If any periodic l

type A test fails to a t 0.75 L.,

the test j

schedule for subse ant type A tests shal be i

reviewed and ap oved by the Commission If two cons utive type A tests fai' to meet O.75L,,

type A test shall perfor ut at less every 18 month 6

[

unt two consecutive

(

e A testa meet

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

c.

'1.

Test duration shal

~

be at least 4 hou 2.

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

atmosphe e 11 be conside stabilized when th chanse in weight average air tempo ature averased over an hour does not dev ste by more than 0

a/ hour from the erase 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 AMEND W NO. I 3 7 BFN Unit 2

4 i

FEB 'O 31988 I

3.7/4.7 CONTAINMENT SYSTEIR LIMITING CONDITIONS FOR OPERATION

SURVEILLANCE REQUIREMENTS 4.7.A.

Primary Containment 4.7.A.2. (Cont'd) 1

d. 1. At least 20 sets of da points at approximat y 4

equal time interva and in no case at inte is greater the o hour shall be provide e proper j

statistic

analysis, j

i

2. The f ure of merit for the

{

ins tation system s 11 never exceed

.25 La*

e. The tesf. shal not*be concluded increasing ated leak rate.

l

. -i~_

1

f. Th's accuracy of each Eype test shall be verified by a 1

supplemental test which:

1. Confirms the accurac of the test by verifyi that the difference bet the supplemental data and the type A test data is within

)

0.25 L.

j

2. Mas duration ufficient to establish a urately the change in skage rate between t type A test and j

the supp emental test.

3. Requir s the quantity of gas i ected into the i

con inment or bled from 4

t containment during the lesental test to be i

i uivalent to at least i

i j

5 percent of the total measured leakage at pa i

i (49.6 psig).

j 4

BFN 3.7/4.7-5 Unit 2 AMENDMENT NO.13 7 j

m-7 s

e

4 1

l 3.7/4.7 CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS i

LIMITING CONDITIONS FOR OPERATION j

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 e penetrations and isolatio valves, which are not pa of a water-sealed system, a not l

1ess than 49.6 psis (ex ept 1

for the main steam iso tion

-valves, see 4.7.A.2.1) and not less than 54.6 psis f r i

water-sealed valves ch operating cycle. Bo ted double-gasketed sea s shall be 4

tested whenl aver 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 pressues of not less than 9.6 psis.

In addition, if the personnel air lock is ened during periods when con ainment integrity is 4

not req ired, a test at the end of ch a period will be conduc ed at not less than 49.6 is.

If the personnel air ek is opened during a pari d when containment int city is required, a test at 2.5 pois shall be e

cted within 3 days after ng 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.

3.7/4.7-6 BFN

Unit ~2

v 3.7/4.7 convAinem e 5isi a g

susvili m CE REQuisanamis l

LirutlNC CONDITIONS FOR OPm? ION f

4.7.A.

Primmet Contm4 = =at 4.7.A.2.s (Cont'd) l The total path leakass f om l

i 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 os containment isolat on valves that terminate b ou suppression poo water level may be exc1M 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 conta eat inciation valves that re in closed-loop, sei e class I lines that vi be water sealed during a i

D will be measured but will excluded when computing a total leakage.

%,, ) B AMENDMENT NO.19 3 3.7/4.7-7 BFN Unit 2

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 i

Primary Containment Leakage Rate Testing Program.

j Note:

An inoperable air lock door 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.) when tested at 2 Pa.

(2)

For door seal leakage, the overall air lock leakage rate is j

s (0. 02 L.) when the air lock is pressurized to (2 2.5 psig for at least 15 minutes).

l

d e

NOM 6IL2 3.7/4.7 mm e

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 l

radioactive materials from the containment atmosphere will be restricted to those leakage paths and associated leak rates asetuned in the accident i

analyses. This restriction, in conjunction with the leakage rate limitation, will limit the site boundary radiation doses to within the d

limits of 10 CFR Part 100 dsring 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 j

I 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, F. 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 I

between leakage tests.

The surveillance testing for measuries ha6==e rates are E - -

- b.;

the a-ir ~ = " i;;

% J vs 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 reisase 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 systes 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 i

suppression chamber marimum 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 voltane is purged to the suppression chenber.

Using the minimus or marimum 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 mawimum water level indications of -1 inch corresponds to a downconer submergence of three feet seven inches and a water volume of 127,800 cubic feet with or 123,700 cubic feet without the drywell-suppression chamber differential 1

pressure control. The mini==a water level indication of -6.25 inches with differential pressure control and -7.25 inches without differential l

pressure control corresponds to a downconer submergence of approximately three feet and a water voltane of approximately 123,000 cubic feet.

i RFN 3.7/4.7-25 AMDMNIIO. 2 0 4 unit 2 a

,e w

I DEC 0 21993

j. Limitations on the annual dose or dose commitment to any MEMBER OF THE PUBLICaue 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 i

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 4

accuracy of the affluent monitoring program and modeling of environmental exposure pathways. The program shall (1) be 4

contained in the ODCM, (2) conform to the guidance of Appendix I

~

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

1 J

a.

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

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.

fe/ (

1

~

Y

/-- n 6.g.5 PROGRAMS Postaccident Samnlina 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

-w w

- - ~.

w--

wwm

(

j INSERT C i

l Section 6.8.4.3 PRIMARY CONTAINMENT LEAKAGE RATE TESTING PROGRAM l

A program shall be established to implement the leakage rate I

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 i

guidelines contained in Regulatory Guide 1.163,

" Performance-Based Containment Leak-Test program, dated

]

September 1995".

4 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 s 1.0 L,.

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

acceptance criteria are s 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,

4 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.

4 5

41k 4 '

1 4

1r

3. 7/4. 7 CONTAINMENT SYSTDiS DEC 0 71994 LIMITING CONDITIONS FOR OPERATION StfRVErt1JLNCE REOUIRDtENTS l

t 3.7.A.

Primary Containment 4.7.A.

Pr4 mary Containment r

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

Excessive' leakage vessel except while performing "open vessel" l

is indicated by a N2 physics tests at power l

consumption rate of > 2% of the primary containment free levels not to exceed 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

b. Primary containment temperature, pressure, and integrity is confirmed if venting operations) at the maxiansa 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 SCFR.

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,.

]

m The containment leakage rat

c. If N2 makeup to the primary shall be demonstrated he containment averaged over following tes ule and J

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 ventins operations) exceeds 10 50 as modified by 542 SCFB, it must be reduced approved exemptions.

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

or the reactor shall be a.

Three type A tests P aced in Hot Shutdown (overall integrated l

l within the next 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months .

containment les rate) j shall be e ed at 40 10-mon intervals during a tdewn at F,

49.6 g, during each 10 ear plant inservice spection.

I n.ff r NIIG.18 6 BFN 3.7/4.7-3 Unit 3

1 i

i 5

INSERT A i

i T8 4.7.A.2 SECOND PARAGRAPE i

Perform leakage rate testing in accordance with the Primary Containment Leakage Rate Testing Program.

i I.

A 4

I 3

6 i

l 5

4 1

j i

l J

t 1

s I

i I

I 4

i l

3.7/4.7 CONv1TfDeHT SYSTEMS FEB 031988 SURVEILLANCE REOUIismrus LThitiNC CONDITIONS FOR OPERATION 4.7.A.

Primary Conta h ant l

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

If any periodic type A test fails to et <

0.75 L.,

the test schedule for subes ont type A testa 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 j

i shall performed at least every 18 months unt two consecutive e A testa meet

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

c.

1.

Testduration' shad be at least 8 hou 2.

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

atmosphere 11 be consid stabilized l

when change in weight average air tempe sture averaged ove an hour does not d

ate by more than 0 *R/ hour from the i

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 .

I12 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)

d. 1. At least 20 sets of dat points at approximate equal time interva and in no case at inte 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,.

e. 13w t'est s'hal conclud an increasing sted leak rate.

l. The accuracy ofWtyp's

~

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 type A test dat is within i

0.25 L -

a

2. Mas duratio sufficient to establish curately the change in eskage rate between e type A test and the sup emental test.

3. Requi s the quantity of gas acted into the con inment or bled from t

containment during the 1emental test to be ivalent to at least 5 percent of the total l

measured leakage at Pa (49.6 psig).

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

3.7/4.7 CONTAINMENT SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 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 a penetrations and isolatio valves, which are not par 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 at least once per parating cycle. Acceptabl methods of testing are halid gas detection, soap

bbles, pressure decay, hydrostatically pressurized j

fluid flow or uivalent.

i The personnel air lock shall be tested at -month intervals l

at an intern 1 pressure of not j

less than 4.6 pais.

In i

addition, the personnel air i

lock is op ned during perioJs j

5 when cont inment integrity is not requi ed, a test at the end of 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 containment intog ity is required, a test at 2

.5 psis 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.

1 BFN 3.7/4.7-6 Unit 3

f I

3.7/4.7 CDETAT15ENT SYSTEMS

,Il0V1819EL Lininus Co-orff 0NS FOR OFenATION sunifilawa REOUIh-d l

l 4.7.A.

Primary Containment 4.7.A.2 3 (Cont'd)

~,

i The total leakage from a1 penetrations and isolati valves shall not exceed i

60 percent of La per l

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

{

containment isolati valves that terminate bel I

suppression pool ater level may be excluded rom the l

total leakas ovided a sufficient id inventory is available ensure the i

sealing f tion for at least j

30 days a a pressure of-

}

54.6 pai. Leakage free contai t isolation valves l

that e in closed-loop, sei e class I lines that j

wil be water sealed during a 3

D will be measured but will b axeluded when computing a total leakage.

$n Jrr h

OIE 16 I 3.7/4.7-7 Unit 3

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

i

)

INSERT B 4

!~

PARAGRAPH 4.7.A.2.g i

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 l_

Note:

An inoperable air lock door does not invalidate the previous successful performance of the overall air lock leakage l

test.

The acceptance criteria for air lock testing are:

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

(2)

For door seal leakage, the overall air lock leakage rate is 5 (0.02 L.) when the air lock is pressurized to (2 2.5 psig for at least 15 minutes).

l l

t

I W16 m E

\\

3. rid.7 nuts 3.7.A & 4.7.A Primmer cont =4===

t l

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

radioactive materials from the containment atmosphere will be restricted l

to those leakage paths and associated leak rates assumed in the accident j

This restriction, in conjunction with the leakage rate analyses.

l limitation, will limit the site boundary radiation doses to within the i

{

limits of 10 CFR Part 100 during accident conditions.

s 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 f

i be no pressure on the system thus greatly reducing the chances of a pipe j

i break. Ths reactor may be taken critical during this period; however, restrictive operating procedures will be in effect to minimise the probability of an accident occurring.

The limitations on primary containment leakage rates ensure that the f

total containment leakage volue will not exceed the value asemed in the 1

accident analyses at the peak accident pressure of 49.6 pois, P. As an a

added conservatism, the measured overall integrated leakage rate is further limited to 0.75 La during performance of the periodic tests to account for possible degradation of the containment leakage barriers between leakage tests.

m, c_

o The surveillance testing for measur4=+,leakane raena =r crim. eui the r =l e ~- c f ".J

% J or 10 wa Part 50 (type A, B, and C tests).

i l

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 system blowdown from 1,035 peig. 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 maxim a pressure. The design volume of the suppression chamber (water and air) was obtained by considering that the total voluse of reactor coolant to be condensed is discharged to the j

l suppression chamber and that the drywell volume is purged to the j

suppression chamber.

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

indications of -1 inch corresponds to a downcomer submergence of three feet seven inchas 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 2

with differential pressure control and -7.25 inches without differential pressure control corresponds to a downcomer submergence of approximately three feet and water volme of approximately 123,000 cubic feet.

RFN 3.7/4.7-24 Unit 3 N W NO.I61

i I

l DEC 0 21993 1-than 8 days in gaseous effluents released from each unit to l

areas beydad the SITIF BOUNDARY conforming to Appendix I to 10 CFR Part 50.

4 6

j. Limitations on the annual dose or dose commitment to any MDIBER OF THE PUBLIC due to releases of radioactivity and to radiation from uranium fuel cycle sources conforming to 40 CFR Part 190.

6.4.4.2 RADIOLOGICAL ENVIROl9fENTAL 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 i

environmental exposure pathways. The program shall (1) be cuus.alued in the ODCM, (2) conform to the guidance of Appcnf.i= I to 10 CFR Part 50, and (3) include the followings 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.

w

//

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

INSERT C l

Section 6.8.4.3 PRIMARY CONTAINMENT LEAKAGE RATE TESTING PROGRAM A program shall be established to implement the leakage rate j

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

10 CFR 50, Appendix J, Option B, as modified by' approved exemptions.

This program shall be in accordance with the I

guidelines contained in Regulatory Guide 1.163, 1

" 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.

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

at l

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

Leakage Rate acceptance criteria are:

i Primary Containment leakage rate acceptance criterion is a.

5 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 least 15 minutes.

ENCLOSURE 2 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)

UNITS 1, 2, AND 3 i

PROPOSED TECHNICAL SPECIFICATION (TS) CRANGE T8-364 REVISED PAGES I

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 i

3.7/4.7-25 3.7/4.7-25 3.7/4.7-24 j

6.0-23d 6.0-23c 6.0-23d 1

6.0-23e 6.0-23d 6.0-23e 6.0-23e 1

II.

REVISED PAGES i

See attached.

k

3.7/4.7 C0!rfATHMENT SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEIt.f.AltCE REOUIREMENTS 3.7.A.

Primary Containment 4.7.A.

Primary Containment 2.a.

Primary containment

2. Intearated Leak Rate Testinn 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 dryvell

b. Primary containment temperature, pressure, and 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 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.

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 SCFB, 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 .

4 BFN 3.7/4.7-3 Unit 1

3,7/4.7 CONTAINMENT SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEftTAMCE REOUIREMENTS 4.7.A.

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

Deleted d

c.

Deleted BFN 3.7/4.7-4 Unit 1

3.7/4.7 CONTAIMMENT SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEIT.T AMCE REQUIREMENTS 1

4.7 A.

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

d. Deleted d

e. Deleted d

f. Deleted

]

I BFN 3.7/4.7-5 Unit 1

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

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

Perform required local leak rate tests, including the primary containment air lock 1eakage rate testing in accordance with the Primary Containment Leakage Rate i

Testing Program.

l Note: An inoperable air lock door does not invalidate the previous successful performance of the overall air lock l

1eakage test.

The acceptance criteria for air lock testing are:

(1)

Overall air 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 1

THIS PAGE INTENTIONALLY LEIT BLANK l

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

i l

3.7/4.7 B&gE),

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 j

i 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.

I 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 i

i added conservatism, the measured overall integrated leakage rate is further limited to 0.75 L during performance of the periodic tests to

]

a 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 pais. Since all of the gases in the dryvell 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 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 peig, 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 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 i

Unit 1

. _. _. _ _ _ _ _ _ _ _ _ _ _. _ _ _.. _... _.. _ _ _ _ _ _... _. ~... _ _

- accuracy of the measurements of radioactive materials in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring.

6.8.4.3 PRIMARY CONTAIlWENT 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 Cont:ainment 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 s **

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

BFN 6.0-23d Unit 1

THIS PAGE INTERTIONALLY LEIT BLANK 1

BFN 6.0-23e Unit 1

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

6.8.5 PROGRAMS t

Postaccident Samnlina 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:

j (i)

Training of personnel, (ii) Procedures for sampling and analysis,

~

(iii) Provisions for maintenance of sampling and analysis.

6.9 REPORTING REQUIREMENTS l

ROUTINE REPORTS 6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following identified i

reports shall be submitted to the Director of the Regional Office of NRC, unless otherwise noted.

1 6.9.1.1 STARTUP REPORT a.

A sumsary 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 1

i BFN 6.0-24 Unit 1

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

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

Primary Containment 4.7.A.

Primary Containment 1

2.a.

Primary containment 2.

Interrated Leak Rate Testina integrity shall be 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 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 l

physics tests at power consumption rate of > 2% of 1

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 4

integrated leakage rate, normal drywell operating L., does not exceed the pressure of 1.1 pais, this equivalent of 2 percent of value is 542 SCFE. If this 1

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.

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

$42 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 .

BFN 3.7/4.7-5 Unit 2

3.7/4.7 CONTAINMENT SYSTEMS I

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS 4.7.A.

Primary Containment 1

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

Deleted y

l c.

Deleted e

i i

i 4

1, 3

j O

44 1

1 E

3.7/4.7-4 Unit 2

l i

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

Primary Containment 4.7.A.2. (Cont'd) 1

d. Deleted q

e. Deleted

f. Deleted l

l 1

l i

e BFN 3.7/4.7-5 Unit 2

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

Primary Containment 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.

Note: An inoperable air lock i

door does not invalidate the previous j

successful performance 4

of the overall air lock leakage test.

The acceptance criteria for air lock testing are:

(1)

Overall air 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).

J l

3.7/4.7-6 Un t 2

i 1

i THIS PAGE INTENTIONALLY LEFT BLANK 1

l 1

l BFN 3 7/4*7-7 Unit 2

t 3.7/4.7 BAIX3.

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 radioactive materials from the containment atmosphere will be restricted a

i 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.

i 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 ministize the probability of an accident occurring.

4 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 a

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 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 chamher (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 pais, which is below the maximum of 62 peig. 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 ministan 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 a water volume of approximately 123,000 cubic feet.

f 4

BFR 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 radiation from uranium fuel cycle sources conforming 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 i

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 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 j

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.

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, Appendiz J, Option B, as modified by approved BFN 6.0-23c Unit 2

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

i 4

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

" Performance-Based C'ntainment Leak-Test program, dated o

September 1995".

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

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

a Leakage Rate acceptance criteria are:

a.

Primary Containment leakage rate acceptance criterion is 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

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

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

THISPAGEkNTENTIONALLYLEFIBLANK BFN 6.0-23e Unit 2

~... - - -

3.7/4.7 CONTAINMENT SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEIT.T.AMCE REQUIREMENTS t

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 psig. Corrected to integrated leakage rate, normal drywell operating L, does not exceed the pressure of 1.1 pais, 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

$42 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 .

BFN 3.7/4.7-3 Unit 3

3.7/4.7 CONTAINMENT SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEItiANCE REOUIR.EMENTS 4.7.A.

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

Deleted

]

c.

Deleted l

)

a f

I e

e BFN 3.7/4.7_4 Unit 3

3.7/4.7 CONTAIlOENT SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS 4.7.A.

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

d. Deleted

e. Deleted i

f. Deleted d

1

)

l 4

O BFN 3.7/4.7-5 Unit 3

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

Primary Containment 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 j

Containment Leakage Rate Testing Program.

Note: An inoperable air lock door 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 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).

i 4

BFN 3.7/4.7-6 Unit 3

THIS PAGE kNTENTIONALLY LEFT BIJugt l

l BFN 3.7/4.7-7 Unit 3

3.7/4.7 3A374 3.7.A & 4.7.A Primary Containment t

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 a

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 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 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 chamher.

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 downcomer submergence of approximately three feet and water volume of approximately 123,000 cubic feet.

BFN 3.7/4.7-24 Unit 3

\\

l i

i h~

1 i

6.8.4.3 PRIMARY CONTAIletENT 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 j

CFR 50, Appendix J, Option B, as modified by approved j

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

" Performance-Based Containment Leak-Test program, dated j

September 1995".

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

The marimum allowable primary containment leakage rate, L.,

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

Leakage Rate acceptance criteria are:

Primary Containment leakage rate acceptance criterion is a.

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

b.

Air lock testing acceptance criteria aret (1) Overall air lock leakage rate is 1 0.05 L when tested a

at 1 P.,

(2) Air lock door seals leakage rate is 1 0.02 L, when the overall air lock is pressurized to 1 2.5 pais for at least 15 minutes.

?

BFN 6.0-23d

. Unit 3

~

THIS PAGE IETENTIONALLY LEFT BIJLNK l

BFN 6.0-23e Unit 3

1 BP-213 LICENSING TRANSMITTAL TO NRC

SUMMARY

AND CONCURRENCE SHEET A concurrence signature reflects that the signatory has assured that the submittal is appropriate and consistent with TVA Policy, applicable commitments are approved for implementation, and supporting documentation for submittal completeness and accuracy has been prepared.

DATE 1/04/96 DATE DUE NRC ASAP SUBMITTAL PREPARED BY (1)

S, M.

Kane Name Signature SUBJECT'Sucolement to Technical Soecification Chance TS 364 Imolementation of 10 CFR 50. Anoendix J.

Ootion B.

Performance Based Testina_

Does this submittal contain Corrective Action / Commitment?

Yes

_X_ No INDEPENDENT REVIEW (2)

DATE CONCURRENCE (3)

NAME ORGANIZATION SIGNATURE DATE PORC Chairman N/A cer Bob Moll-1/03/96 J. M.

Corev RadChem j

J.

E. Maddox Maintenance J.

E.

McCarthy Mech./Nuc. Ena.

T.

J.

McGrath NSRB Chairman R.

J.

Moll Ooerations G.

D.

Pierce Tech. SuoDort E.

Preston Plant Manaaer Pedro Salas Site Licensino T.

D.

Shriver NA&L Manacer H.

L. Williams Enc. & Matls. Mar.

NRC SUBMITTAL SUPPLEMENTAL INFORMATION FOR TECHNICAL SPECIFICATION (TS) 364 -

IMPLEMENTATION OF 10 CFR 50, APPENDIX J, OPTION B, PERFORMANCE BASED TESTING CONCURRENCE COPY PLEASE RETURN TO SITE LICENSING ASAP PLEASE CONTACT:

STEVEN KANE (x7854)

WHEN REVIEW IS COMPLETED

.

ML20096C440

Text

SUMMARY

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