Text

Proposed Tech Specs,Allowing Bfn,Units 2 & 3,to Operate at Uprated Power Level of 3458 Mwt
	ML18038B961
Person / Time
Site:	Browns Ferry
Issue date:	10/01/1997
From:	; TENNESSEE VALLEY AUTHORITY
To:	;
Shared Package
ML18038B959	List: ML18038B958; ML18038B960; ML18038B961;
References
	NUDOCS 9710070320
	Download: ML18038B961 (207)
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ENCLOSURE 3 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)

UNITS 2 AND 3 PROPOSED TECHNICAL SPECIFICATION (TS) CHANGE TS-384 MARKED PAGE S I. AFFECTED PAGE LIST The following pages have been revised and an 'X'as been placed in the right hand margin to indicate where changes OCCQr.

0 eratin License Unit 2 Unit 3 Technical S ecifications Bases Unit 2 Unit 3 Unit 2 Unit 3 1.1-5 1.1-5 B 3.1-44 B 3.1-44 3 1 23

~ 3 1 23

~ B 3.4-3 B 3.4-3

3. 3-6 3.3-6 B 3.5-4 B 3.5-4 3 3 7

~ 3 3 7

~ B 3.5-12 B 3.5-12 3.3-34 3.3-34 B 3.5-24 B 3.5-24 3.4-4 3.4-4 B 3.5-28 B 3.5-28 3.4-7 3.4-7 B 3.6-2 B 3.6-2 3.5-5 3.5-5 B 3.6-7 B 3.6-7 3.5-13 3.5-13 B 3.7-1 B 3.7-1 3 7 1

~ 3 7 1

~ B 3.7-2 B 3.7-2 3 7 3

~ 3 7 3

~ B.3.7-3 B 3.7-3 3 s 7 3a 3 s 7 3a B 3.7-5 B 3.7-5 3.7-5 3. 7-5 B 3.7-6 B 3.7-6 B 3.7-8 B 3.7-8 II. MARKED PAGES See attached.

97iQ07032Q 97iOOi PDR ADOCK 05000260 I P PDR

0 e SRPTENBER 18, 1996 (2) pursuant co the Act and 10 cFR Parts 40 and 70. co receive, possess, and use ac any time source and special nuclear material as reactor fuel in accordance with the limitations for storage and amounts zequixed for reactor operation, as described in the Final Safety Analysis Repozt as supplemented and amended; (3) Pursuaat co the Act and 10 cPR Parts 30, 40, and 70, co receive, possess, and use'at any time any byproduct, source, and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumeacacioa and radiatioa monitoring equipmeat calibration. and as fission detectors 'ia aaxnxacs as required:

Pursuant "o che Act and 10 cFR Parts 30, 40, and 70 'o'receive,

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possess, and use in amounts as xequized any,byproduct, or special nuclear material ~ichout restriction to chemical ox physical form for sample aaalysis or equipmenc and instrument calibration ox associated with radioaccivs apparatus ox compoaencs:

'.'4)

(5) pursuant co che Act and 10 cFR Parts 30 and 70, to possess but aoc sepazace. such bypmduct and specia1 nuclear materials as may be produced by the operation of che facQity.

A This license shall be deemed co contain and is subject co the conditions specified in che following Coaxnission xegulatioas in 10 CPR Chaptez I: Part 20, Section 30.34 of Part 30, Sectioa 40.41 of Part 40, SeCtiOnS 50.54 and 50.59 Of Part 50, aad SeCtiOa 70.32 Of part 70: is subject to all applicable provisions of the Act and co the rules, regulations, and orders of che Commission aov ox hereafter in effect; and is subject to the'dditional condicioas specified or incorporated below:

The licensee is authorized co operate che facility at steady etete teeetot tote Ooeet levele oot ta exocet o~ggteeeeeeeete thezaal. t~gysg (2)

The Technical Specifications contained in Appendices A, aad B. as revised through Amendmenc No. 246, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

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I Definitions I

)

I

1. 1 Definitions (continued)

PHYS/CS TESTS PHYSICS TESTS sha11 be those tests performed to measure the fundamental nuclear characteristics of the! reactor core and related instrumentation.

These tests are:

a. Described in Section 13.10, Refueling Test Program; of the FSAR;

b. Authoriaed under the provisions of

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10 CFR 50.59; or c.. Otherwise approved by the. Nuclear Regulatory Commission.

~QUES'8 RAT THERNAL POMER RTP shall be a total reactor/core heat transfer

~ 'RTP rate to the reactor coolant o+I~9.Nt.

SHU1l}OHN HARBIN (SDN) SON shall be the amount of reactivity by which ghe

) ~ reactor is subcritical or would be subcritical assuming that:

a.. The reactor is xenon free;

b. The moderator temperature is 68 F: and

c. All control rods are fully inserted except for the single control rod of highest reactivity worth, ~hich is assumed to be fully withdrawn.

Mith control rods not capable of being fully inserted, the reactivity worth of these control rods must he accounted for in the determination of SDH.

STAB)BRED TEST BASIS, A STAGGERED TEST BASIS shall consist of the testing of one of the systews, subsysteas, chinnels, or other designated components during the interval specified by the Surveillance Frequency, so that all systems, subsystems, channels, or other designated cowponents are tempted during n Surveillance Frequency intervals, where n is the total number of systems, subsystems, channels, or other designated capponents in the associated function.

(continued)

SFN+NIT 2 1-1-5 Amendment

SLC System 3-1.7 1

SUR E? LLANCE REQUIREHENTS continued SURYKlLLANCK FREQUENCY

~ I SR 3.1.7.5 Verify the SLC conditions satfsfy the following equation: 31 days 13 wt- %) {86 Qpm) (19.8 atoag)

Once withfn 24 where, hours after water or boron C sodium pentaborate solutfon fs added to the concentration (weight percent) solution.

Q pump flow rate (gpm)

E Boron-lo, enrichment {atom percent Boron-10)

$$ 3.1.7.6 Verify each pump develops a flow rat>> 18 months 39 pm at a discharge pressure psfgo

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Sp 3.l.T.7 Verify flow through one SLC subsystel free 18 months on a

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pump into reactor pressure vessel. STAGGEREO TEST BASIS 1

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g 3.1.1.8 Verffy all piping between storage tank puap suction is unblocked.

and 18 aonths (contfnued) 0 ~

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Q-UNlT R 3.1-23 Amendment

RPS tnstl'LIRHIntation 3.3.1.1 Tjblt 3o3o1 ~ 1 1 (pa00 1 Of 3)

Reactor prataatfan syata Intnaentatfan APPLI CASLE NOES Ol

'CNDREouIREO IT IOIS RESEREECED OTHER CHAHHELS fROI SPECI PIEO PER TRIP REQJIREP SURVEILLAHCE ALLCQIS LE tuHCTIOI CONITIOIS SYSTBI ACTIOI Oot REOJlaQCETS VALLEI 1~ Interaidiete Ranfft Nanl tora

~. Neutron fLux -Hfah SR 5.5o1 ol ~ 1 8 120/125 dfvfefono of 5o5ololo3 SR 3.3,1.1.5 fULL jcjL0 SR 3o3o1 ol oh SR 3o3ol of o9 SR 3 3.1 ~ 1.'l4 g(a) SR 5o'5.1.1.1 x 120/125 SR 5o5o I ~ 1 o4 dfvfafae af Sk 3.3.1.1.O fULL jejfa SR 3.3.1 ~ 1.14

b. Inop SR 3.3o 1 o1 o3 SR 3.3.1.1.14 o

g(a) SR 3o5o 1 o1.4 SR 3o3o 1 ~ 1 ~ 14 Awrege Pauer RjnS0 Nanf tora

a. Neutron P lux -Hfgho 3(b) SR 3.3.'f ~ lol 5 153 RTP Setdaun SR 3o3.1 ~ fob SR 3.3.1.1.7 SR 3.5o I ~ 1 o13 SR 3.3.F 1.15

b. Ffau Rfajed Sfafated 3(b) SR 5.5. l 1 ~ I II Thereof Pauer -HfSh SR 3.3.1 1.2 + 7IX TP and SR 5.5.1.1.7 S TP SR 3.3.1.1.13 SR 3o3o1 ~1 ollf( 66%

a. Heutran flux -Hfeh 3(b) SR 3o3o 1 ~1 ~ 1 s 1203 RTP SR 3.3o'I ~ 1 o2 SR 3.3.1.1o7 SR 5.5.1 1.15 SR 5.5 ~ 1 ~ 1 ~ 16 (continued)

(0) Vfth any central rod Hfthdraal froN ~ tort otLL ccntafnfnff ona or aero fmf asaoeblfeao (b) Each APRIL chaail Pravidea inPuta to both trfP ayateeeo BFN-UNlT 2 3.3-6 Amendment

RPS Instrlimentatkon 3.3.1.1 Table 3.3.1.1-1 (page 2 of 3)

Reactor Protection Systes Instrumentation APPLICABLE CONDITIONS NODES OR REOUI RED REFERENCED OTHER CHANNELS FRON SPECIF IEO PER TRIP REOUIRED SURVEILLANCE ALLOVABLE COND I T I ON 5 STSTEH ACTION D.1 REOUIRENENTS VALUE

2. Average Pouer Range Nonitors (continued)

d. Downscale SR 3.3.1.1.7 8 3X RTP SR 3.3.1.1.8 SR 3.3.1.1.14

e. Inop 1,2 SR 3.3.1.1.7 SR 3.3.1.1.8 SR 3.3.1.1.14

3. Reactor Vessel Steaa 1,2 SR 3.3.1.1.1 I 8 1055 psig Dose Pressure -High SR 3.3.1.1.8 SR 3.3.1.1.10 IOg O SR 3,3.1.1,14 4~ Reactor Vcsscl Vater 1,2 SR 3.3.1.1.1 8 538 inches Level -Ltw, Level 3 SR 3.3.1.1.8 above vessel SR 3.3.1.1.13 zcIo SR 3.3.1.1.14

5. Hain Stean Isolation SR 3.3.1.1.8 S 10X closed Valve -Ciosul e SR 3.3.1.1. T3 SR 3.3.1.1.14

6. Dryucli Pressure -High 1,2 SR 3.3.1.1.8 8 2.5 psig SR 3.3.1.1.13 SR 3.3.1.1.14

7. Scrota Discharge Volvo..

Vater Level -High

~. Resistance Teaperature 1,2 SR 3.3.1.1.8 8 50 gallons Detector SR 3.3.1.1.13 SR 3.3.1.1.14 5(a) SR 3.3.1.1.8 8 50 gallons SR 3.3.1.1.13 SR 3.3.1.1.14

b. Float Suitch 1,2 SR 3.3.1.1.8 S 50 gallons SR,3.3.1.1.13 SR 3.3.1.1.14 5(a) SR 3.3.1.1.8 8 50 gallons SR 3.3.1.1.13 SR 3.3.1.1.14 (cont inued)

(a) Mith any control rod Nithdraun fron a core cell containing one or sore fuel asseablies.

BFN-UNIT 2 3,3-7 Amendment

ATWS-RPT Instrumentation 3.3.4.2 SURVEILLANCE REOUIREMENTS continued SURVEILLANCE FREQUENCY 4

SR 3.3.4.2.2 Perform CHANNEL FUNCTIONAL TEST. 92 days SR 3.3.4.2.3 Perform CHANNEL CALIBRATION. The 18 months Allowable Values shall be:

a. Reactor Vessel Water Level Low Low, Level 2: a 471.52 inches above vessel zero; and

b. 'e Steam Oome Pressure High:

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SR 3.3.4.2.4 Perform LOGIC SYSTEM FUNCTIONAL TEST 18 months including breaker actuation.

BFN-UNIT 2 3.3-34 Amendment

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v r Region I -. 0 iRodUne Rod Line

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g po D

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L'0 gale; Opesatlon got fetmN&In Thh Reeloa 4 30 Q

I 10 0 5 10 ~

15 20 25 30 06 4) 45 50 65 00 65 .70 7S 60 OS QO QS 100 106 K CCea ROVt fPerCeIII Ot ra(ad)

O 7l 0

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t S/RVs 3.4.3 3..4< REACTOR COOLANT SYSTEM (RCS) 3.4L3 Safety/Relief Valves {5/RVs)

LCO: 3.4.3 The safety function of 12 S/RVs shall be OPERASLE.

APP

) ICAB ILITY:

1 MODES 1, 2, and 3.

4 A T QNS REQUIRED ACTION COMPLETION TIME I

A.. One or more required A. 1 Se in MODE.3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months S/RVs inoperable.

A.2 8e in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months SUR EILLANCE REQUIREMENTS FREQUENCY SR.'.4.3.1 Verify the safety function the required S/RVs are within lift+ settings of the of In accordance 3% with the setpoint as follows: Inservice Testing Program Number of Setpoint

~+Vs 4 105 ) l 3%

1 1115 ~ ff~S 5 12 Jl55

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Following testing, lift settings shall be within t 1%.

(continued)

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BF UNIT 2 3.4-7 Amendment

t SU SQ

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EILLANCE REOUIREHENTS 3.5.1.5 continued SURVE ILLA'K f

<<<<<<<<<<e <<<<<<<<<<<<<<<<<<<<<<<<<<NOT <<<<<<<<<<<<<<<<e <<<<<<<<<<<<<<<<e <<<<

ECCS -Operating 3.5.1 FRE(UENCY Only required to be performed prior to entering NOE, 2 from HOOK 3 or 4, shen in HOOK 4 > 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

<<<<<<<<<<<<<<e e<<<<<<<<<<<<<<<<<<<<e<<<<<<<<<<e ee<< <<ee Verify each recirculation pump discharge 31 days valve cycles through one complete cycle of full travel.

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I Sg 3.5.1.6 Verify the 'following KCCS pumps develop the In accordance specified flow rate against a system head With the corresponding to the specified reactor Inservice pressure. Testing SYSTEM HEAD Program CORRESPomuj:

TO A VESSEL. TO NQ TORVS OF DIFF ERENTIAL

~SS Bf FLOM ~T ~PIPS PR SSUR Core Spray a 6250 gpm 2 a 105 psid LPCI e 19,200 gp 2 a 20 psid LPCI z 10,450 gpss 20 psid l

,sa 3.5.1.I <<<<ee<<<<<<e e<<<<<<ee<<<<<<<<<<NOTE<<e ee<<eee ee<<<<eee<<<<<<ee Not required to be perfonaed unt)1 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure and floe are adequate to perfonu the test.

<<reeeeeee<<<<eee<<e<<ee<<<<<<eee<<<<<<ee<<eeeeeee<<<<<<ee

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Vor<f oith reactor Orassore cma%nd 92 days 2 sig, the HPCI pump can develop a flow rate a'.5000 gpss against a systaa head corresponding to reactor pressure.

{continued) 3.5-5 Amendment

RCIC System 3.5.3

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SQR ILLANCK REQUIREHEMTS SURVEILLANCE FRE(UEHCY SR . $ .5.3el Verify the RCIC System piping is filled 31 days with water from the pump discharge valve to the injection valve.

SR} 3.5.3.2, Verify each RCIC System manual, power 31 days operated, and automatic valve in the flow path, that is not locked, sealed, or otherwise secured fn position, is in the correct position.

SRj 3-5.3.3 aaeeaaaeaaa>> >>ac>>>>a N0TEeeeea>>a aeeaaaeeeaea Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure and flow are adequate to p'erform the test.

eaaeeeaeaaeeeeaaeeaaeeeaeaaaeeeaaeeeeeaaeee ID'.O Yerk y, u' reactor pressure singe fsig pZ says an psi'g, the RCIC pump can develop a l~o flow rate a 600 gpe against a system head

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corresponding to reactor pressure. X'$

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3.3.3.4 e e e e e a a e e a e a e a a a a e a NPTE e a a a e a e e a a e a e a a a a a a e Hot required to be performed unt)l 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steaa pressure and flow are adequate to perfore the test.

eeaaeeaeaaeeaaaaaeaae'e ~ 'a>>a>>ace>>aeeeaaaaeeea Verify, with reactor pressure a 165 psig, 18 months the RCIC pump can develop a flow rate a 600 gpss against a system head corresponding to reactor pressure.

(continued)

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~ ~ 8 -UNT 2 3.5-13 Amendment I

RHRSW System 3.7.1 5 ( ~II<)

3.7.1 Residual Heat Removal Service Water (RHRSW) System a~d Mlk: -<e He 4 4 LCO 3.7.1 NOTES

1. With 1 or 2 units fueled, each subsystem must have at least. one OPERABLE RHRSW pump.

2. With 3 units fueled, two RHRSW subsystems must have two OPERABLE RHRSW pumps.

Four RHRSW subsyste s 11 be OPERABLE.

QHS APPLICABILITY: MODES 1, 2, and 3.

ACTIONS NOTE Enter appl'icable Conditions and Required Actions of LCO 3.4.7, "Residual Heat Removal (RHR) Shutdown Cooling System- Hot Shutdown," for RHR shutdown cooling made inoperable by RHRSW System.

CONDITION REQUIRED ACTION COMPLETION TIME A. One RHRSW subsystem or A.l Restore RHRSW 30 days required pump subsystem or required inoperable. pump to OPERABLE status.

(continued)

BFN-UNIT 2 3.7-1 Amendment

RHRSW System 3.7.1 CONDITION REQUIRED ACTION 'OHPLETION TIHE D. Required Action A.l, 0.1 Be in HODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months B.3, or C. 1 and associated Completion AND Time not met.

0.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months oR a%f5 i nope ale SURVEILLANCE REQUIREHENTS SURVEILLANCE FREQUENCY SR 3.7.1.1 Verify each RHRSW manual and power operated 31 days valve in the flow path, that is not locked, sealed, or otherwisesecured in position, is in the correct position'r can be aligned to the correct position.

~< LO<<.> QgS ir~)...t<<.c Sg, 3.7. l. 2 Verify il e ouri<<>r wi4r fr~prri4u<e of 9I F'

01'l,. [;;4 spe.,'Pi'.J i h/0 Fi gare 3 7,~ l l Atl5 ir r<<<< t<<<< ~

~p BFN-UNIT 2 317 3 Amendment

RHRSW System end 371 tARS Figure 3.7.1-1 Reactor Thermal Power Versus Ultimate Heat Sink Temperature Limit Unacceptable e~ 99 E

P- 98 O

Acceptable 97 93 93.5 94 94.5 Ulthnate Keat Sink Temperature (degrees F)

BFN-UNIT 2 Qt 7~ 3 Amendment

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EECM System and UHS 3.7.2 SURVEILLANCE FREQUENCY SR 3.7.2.1 Verify the average water temperature of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months UHS is w 95'F.

SR 3.7.2.2 NOTE Isolation of flow to individual components does not render EECM System inoperable.

Verify each EECM subsystem manual and power 31 days operated valve in the flow paths servicing safety related systems or components, that is not locked, sealed, or otherwise secured in position, is in the correct position.

SR 3.7.2.3 Verify each required EECM pump actuates on 18 months an actual or simulated initiation signal.

BFN-UNIT 2 3.7-5 Amendment

V SLC System S 3.1.7 2'A)ES I

S YE ILLANCE SR 3 7.4 R UIREHENTS continued) This Surveillance requires the amount of Boron-10 in the SLC solution tank to be determined every 31 days. The enriched sodium pentaborate solution is made by combining stoichiometric quantities of borax and boric acid in demineralized water, Since the chemicals used have known Boron-10 quantities, the Boron-10 quantity in the sodium pentaborate solution formed can be calculated. This parameter is used as input to determine the volume requirements fot SR 3.1.7.1. The 31 day Frequency of this Surveillance is appropriate because of the relatively slow

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0 variation of boron concentration between surveillances.

~SR 7. 5

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Oemonstrating that each SLC System develops a flow rate a 39 gpm at a discharge pressure a psig ensures that-pump performance has not degraded during the fuel cycle.

This minimum pump flow rate requirement ensures that, when combined with the sodium pentaborate solution concentration and enrichment requirements, the rate of negative reactivity insertion from the SLC System will adequately compensate for the positive reactivity effects encountered during power reduction, cooldown of the moderator, and xenon decay. This test confirms one point on the pump design .curve and is indicative of overall performance. The l8 month Frequency is acceptable since inservice testing of the pumps, performed every 92 days, will detect any adverse trends in pump performance.

R 3 7. a SR ')~~7.

These Surveillance.es ensure that there is a functioning flow path fraa the boron solution storage tanlr, to the RPV,

, including the firing of an explosive valve. The for the explosive valve shall be from the 'same replacement'harge manufactured batch as the one fired or Frol another batch that has boon certified by having one of that batch successfully fired. The pump and explosive valve tested should be alternated such that both cowplete flow paths are tested every 36 months at alternating 18 month intervals.

The Surveillance may be performed in separate steps to O. (continued)

Pl-UNIT 2 B 3.1-44 Amendment

Recfrculat1an Laaps Operatfng

$ .4.1

)ASES WPPL I CABLE Safety analyses performed for FSAR Chapter 14 implicftly ISAFETY ANALYSES assume core conditions are stable. However, at the high

{continued) power/law flaw earner of the power/ftaw map, an increased prababflity for limit cycle oscillations exists (Ref. 3) depending on cambfnatians of operating conditions (e.g.,

power shape, bundle power, and bundle flow). Generic

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evaluations indicate that when regional power ascillat1ons become detectable on the APRHs, the safety margin may ba

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insufficient under some operating conditions to ensure actions taken to respond to the APRES signals would prevent violation of the HCPR Safety Lfmft (Ref. 4). HRC Generic Letter 86'-02 {Ref. 5) addressed stability calculation methodology and stated that due to uncertainties, 10 CFR SO, Appendix A, General Oesign Criteria (GOC) 10 and IZ could nat be met using analytic procedures on a BMR 4 design.

However, Reference 5 concluded that .operating limftatfons which provide for the detection (by monitoring neutron flux noise levels) and suppression of flux oscillatfons in operating regions of potential instability consistent with the recaaeendatfans of Reference 3 are acceptable to demonstrate camplfance with GOC I0 and l2. The HRC concluded, that regions of potential instability could occur at calculated decay ratios of 0.8 or greater by the General Electric methodology.

Stability tests at operating BMRs were reviewed to 4etermfne a gener1c region of the power/flow map in which surveillance of neutron flux noise levels should be performed. A conservative decay ratio was chosen as the'basis for determfnfng the generic region for surveillance to account for the plant to plant variability of decay rat1o with core and fuel, designs. This decay ratio also helps ensure suffic1ent margin to an instability occurrence is maintained. The ric region has been determined to be bounded hy th rad line and the 50% core flow line. BFN conserv e y implements this generic region with the

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74k'l. Operation Hot Permitted'egion and Regions I and II of Figure 3.4.I-I. This conforms ta Reference 3 recoreendatfons, Operation is permitted in Region II provided neutron flux noise levels are verified to be within limits. ,The reactor made s~itch must be placed in the shutdown position (an immediate scram is required) if Region I is entered.

Recirculation loops operating satisfies Criterion 2 of the HRC Policy Statement (Ref. 6).

t1nued Amendment "I 8FH-UNIT 2 B 3.4-3

KCCS -Operating B 3.5.$

I

~ sos SA GROUND The HPCI System t l7Q fs designed to provide core oo i for a

( ontinved} wide range of reactor pressures {l50 psig to psig).

Upon receipt of an initiation signal, the HPCI turbine stop valve and turbine control valve open and the turbine accelerates. to a specified speed. As the HPCI flow increases, the turbine governor valve is automatically ad(usted to maintain design flow. Exhaust steaa frow the HPCI turbine is discharged to the suppression pool. A full flow test line is provided to route water from and to the CST to allow testing of the HPCI System during normal operation w'ithout injecting water into the RPV.

The ECCS pumps are provided with minimum flow bypass lines,

~hich disch'arge to the suppression pool. The valves in these lines automatically open {for CS and RHR they are already open) to prevent pump damage due to overheating when other discharge line valves are closed. To ensure rapid delivery of water to the RPV and to minimize water haaeer effects, all KCCS pump discharge lines are filled with water. The LPCI and CS Systeja discharge lines are kept full of water using the pressure suppression chamber head tank or i.o 5

condensate head tank. The HPCI SysteII is normally aligned to the CST. The height of water in the CST is sufficient to maintain the piping full of water up to the first isolation valve. The relative height of the. feedwater line connection for HPCI is such that the water in the feedwater lines keeps the remaining portion of the HPCI discharge line full of water.

The ADS {Ref. 4} consists of 6 of the 13 S/RVs. It is designed to provide depressuritation of the RCS during a small break LOCA if HPCI fails or is unable to maintain level'n faired witer the RPV. A6S operation reduces the RPV pressure to within the operating pressure range of the low pressure ECCS subsystews (CS and LPCI), so that these subsystems can provide coolant inventory makeup. Each of the S/RVs used'for automatic depressurization is equipped with one air accumulator and associated inlet check valves.

The accumulator provides the pneeaatic power to actuate the valves.

LJCABLE The ECCS performance is evaluated for the entire spectrum of ETY NALYSES break sizes for a postulated LOCA. The accidents for which ECCS operation is required are presented in References 5 (continued}

B 3.5-4 Aaendmen t

ECCS -Operating

& 3.5.1 SAS j SURVEILLANCE S .5 (continued)

RE{NIIRKHENTS The specified Frequency is once per'l days. However, this SR is modified by a Note that states the Surveillance is only required to ba performed prior to entering HOOK 2 frea HOOE 3 or 4, when in HOOK 4 > 4& hours. Verification during or fallowing each entry inta NME 4 > 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to entering HOOK 2 from NSE 3 or 4 is an exception to the normal Inservice Testing Program generic valve cycling Frequency of 92 days, but is considered acceptable due to the demonstrated reliability af these valves. The 4& hours is intended to indicate an outage of sufficient duration ta allow for scheduling aod proper perfanaance of the Surveillance. If the valve is inoperable and in the open position, the associated LPCI subsystem oust be declared inoperab)e.

S 3 5.I.6 SR 3.5.1.7 nd SR 3.5 The perfarmance'requirements of the low pressure ECCS pujwps 0 are determined through applicatian of the 10 CFR 50, Appendix K criteria (Ref. 7). This periodic Surveillance is performed (.in accordance with the AS% Code,Section XI, requirements for the,ECCS pumps) to verify that the'.KCCS puops will develop the flow rates required by the respective analyses. The low pressure KCCS pump flow rates ensure that adequate core cooling is provided to satisfy the acceptance criteria of Reference 13. The pump flow rates are verified against a system head equivalent to the RPV pressure expected during a LOCA. The total systea puap outlet pressure is adequate to overcame the elevation head pressure between the puep suction and the vessel discharge. the pipin9 friction losses, and RPV pressure present during a UKA. These values may be established bg testing ar analysis or during preoperational testing.

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I The flow tests for the HPCI System are performed at two different pressure ranges such that system capability to provide rated flow is tested at both the higher'and lower operating ranges of the system. Add)tionally, adequate steaa flow.must be passing through the Wain turbine or turbine bypass valves to continue ta control reactor pressure when the HPCI Sgstea diverts steaa flow. Reactor steaa pressure must be aMP psig ta perfam SR 3.5.1.7 and a l50 psig,to perform SR/3.5.1.8. Adequate steaII flow is

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{continued)

I -UNIT Z 8 3.5-12

RC?C System 8 3.5.3 I

g 31.5 EMERtBENCY,CORE COOLING SYSTEMS (ECCS) ANO REACTOR CORE ISOLATION COOLIHS (RCIC) SYSTEM B 3l.5.3 RCIC System s'he SA JKQROUND RCIC System

-System is not part of the ECCS; however, the RCIC is included with the ECCS section because of their similar functions.

The RCIC System is designed to operate either automatically or manually following reactor pressure vessel (RPV) isolation accompanied by a loss of coolant flow from the feedwater system to provide adequate cor e cooling and contr>1 of the RPV water level. Under these conditions, the High Pressure Coolant Injection (HPCI) and RCIC systems perform similar functions. The RCIC System design requirements ensure that the criteria of Reference I are satisf~ed.

The RCIC System (Ref. 2) consists of a steam driven turbine-pump unit, piping, and valves to provide stela to the turbine, as well as piping a'nd valves to transfer water from the suction source to the core via the feedwater system line, where the coolant is distributed within the RPV through the feedwater sparger. Suction piping is provided frea the condensate storage tank (CST) and the suppression pool. Pump suction is normally aligned to the CST to Iiniaize in)ection of suppression pool water into the RPV.

However, if the CST water supply is low, or the suppression pool level is high, a manual transfer to the suppression pool water source ensures a water supply for continuous operation of the RCIC System. The steaa supply to the turbine is piped frea a main steam line upstream of the associated inboard main steam line isolation valve.

fl74.

The RCIC Systla is designed to provide core cooli for a

. wide range of reactor. pressures {l50 psig t sig).

Upon receipt of an initiation signal, the ACIC tu ine accelerates until a 600 gpm flow rate (design flow) fs achieved. As the RCIC turbine flow varies, the turbine control valve is automatically atQusted to maintain design flow. Exhaust steam from the RCIC turbine is discharged to the suppression pool. A fu'll flow test line is provided to route water from and to the CST to allow 'testing of the RCIC (continued)

Bf-UNIT 2 B 3.5-24 Amendment

l

'e I

RCIC System B 3.5.3

~ SASS

~E... t) d)

SVRV ILLNCE )

RE/V REMFNTS steam flow path for the turbine and the f1ow controller posi ti on.

The 3l day Frequency of this SR was derived from the Inservice Testing Program requirements for performing valve testing at least once every 92 days. The Frequency of 31 days is further justified because the valves are operated under procedural control and because improper valve position would affect only the RCIC System. This Frequency has been shown to be acceptable through operating experience.

.3.3 a d R 3.5.3.l The RCIC pump flow r ates ensure that the system can maintain reactor coolant inventory during pressurized conditions with the RPY isolated. The flow tests for the RCIC Systea are performed at two different pressure ranges such that systea capability to provide rated flow is tested both at the higher and lower operating ranges of the systea.

Additionally, adequate steam floe must be passing through the main turbine or turbine bypass valves to continue to control reactor pressure uhen the RCIC S steo diverts stean tlou Re.actor steas) pressure oust he h. gs$ 5 to perforo SR 3.5.3.3 and a ISO psIR to per/ore SR 3.5... Therefore, sufficient time is allowed after adequate pressure is achieved to perforl these SRs. Reactor startup is allowed prior to perfoming the 1ow pressure Surveillance because the reactor pressure is low and the time allowed to satisfactorily perform the Surveillance is short.

Alternately, the low pressure Surveillance test may be performed prior to startup using an auxiliary steam supply.

The reactor pressure is allowed to he increased to normal operatine pressure since it is assumed that the low pressure Surveillance has been satisfactorily completed and there is no indication or reason to believe that RCIC is inoperable.

Therefore, these SRs are modified by Notes that state the Surveillances are not required to be perforaed until I2 hours after the reactor steaa pressure and flow are adequate to perform the test.

0 (continued) 8FN+NIT 2 8 3 '-28 Aaendment

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Primary Containment (

s s.n.l.l l SAS)S (contjnued)

APPL,? CABLE The safety design basis for the primary containment is that

'SAPPY ANALYSES it must withstand the pressures and temperatures of the limiting OBA without exceeding the design leakage rate.

The OBA that postulates the maximum release of radioactive material within primary containment is a LOCA. In the analysis of'his accident, it is assumed that primary containment. is OPERABLE such that release of fission products to the environment is controlled by the rate of primary containment leakage.

Analytical methods and assumptions involving the primary containment are presented in References stand 2. The safety analyses assume a nonmechanistic fission product release following a OBA,.which forms the crasis.for. determination of offsite doses. The fission product release is, in turn, based on an assumed leakage rate frea the primary containment. OPERABILITY of the primary containment ensures that the leakage rate assumed in the safety analyses is not exceeded.

The maxieaa allowable leakage rate for the primary containment (L,) is Z.OX by weight of the containment air per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at the design basis LOCA maximum peak containment pressure {P,) o<5K psig (Ref. I).

L 50.3 Primary containment satisfies Criterion 3 of the NRC Policy Statement (Ref.').

Primary containment OPERABILITY is maintained. by limiting leakage to s 1.D L except prior to the first startup after perforaing. a required Primary Containment Leakage Rate Testing iNograia leakage test. At this time, applicable leakage 1iaits eust be met. Coipliance with this LCO will ensure a p'rtmary containment configuration, including equ)pient hatches, that is structurally sound and that will liait leakage to those leakage rates assumed in the safety analyses.

Individual leakage rates specified for the primary containment air lock are addressed in LCD 3.6.1.2.

h 0 (continued)

Sljll-UNST 2 8 3.6-2 Amendment

I Primary Containment Air Lock 8 3.6.1.2

~'6 BAQGROUNO end leak tightness are essential for maintaining primary (continued) containment leakage rate to within limits in the event of a OSA. Not maintaining air lock integrity or leak tightness may result in a leakage rate in excess of that assttmed in the unit safety analysis.

AP ICASLE The"OSA that postulates the maximum release of radioactive ANALYSES material within primary containment is a LOt;A.

analysis of this accident, it is assumed that primary containment is OPERABLE, such that release of fission products to the environment is controlled by the rate of primary containment leakage. The primary containment is designed wi'ih a maximum allowable leakage rate {L,) of. 2.la by weight of the containment air per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at the calculated maximum peak containment pressure {P.) of Gag psig (Ref. 3), This allowable leakage rate forms the osis for the acceptance criteria imposed on the SRs associated, with the air lock.

primary-containment air lock OPERABILITY is also required to mfnfyfze the amount of fission product gases that may escape primary containment through the air lock and contaminate and pressurize, the secondary containment.

The primary containment air lock satisfies Crfterion 3 of the NRC Polfcy Statement (Ref. 4).

As part of prfmary containment, the air lock's safety function fs related to control of containment leakage rates following a DBl. Thus, the air lock's structural integrity and leak tightness are essential to the successful mitigation'f such an event.

The primary containment air lock is required to be OPERABLE.

For the air lock to be considered OPERABLE, the air lock interlock mechanism must be OPERABLE, the air lock must be in complia'nce with the Type B air lock leakage test, and both air lock doors must be OPERABLE. The interlock allows only one air lack door*'to be opened at a tice. Thfs provision ensures that a gross breach df primary containment does not exist when primary containment is required to be

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C (continued) 8 -VNIT Z B 3.6-7 Amendment

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RHRSW System ~d >+~

B 3.7.1 BASES (continued)

APPLICABLE The RHRSW System removes heat from the suppression pool to SAFETY ANALYSES limit the suppression pool temperature and primary containment pressure following a LOCA. This ensures that the primary containment can perform its function of limiting the release of radioactive materials to the environment following a LOCA. The ability of the RHRSW System to support long term cooling of, the reactor or primary containment is discussed in the FSAR, Chapters 5 and 14 (Refs. 2 and 3, respectively). These analyses explicitly assume that the RHRSW System will provide adequate cooling support to the equipment required for safe shutdown. These analyses include the evaluation of the long term primary containment response after a design basis LOCA.

The safety analyses for long term cooling were performed for various combinations of RHR System failures and considers the number of units fueled. With one unit fueled, the worst case single failure that would affect the performance of the-RHRSW System is any failure that would disable two subsystems or pumps of the RHRSW System (e.g, the failure of an RHR Suppression Pool Cooling/Spray return line valve which effectively disables two RHRSW subsystems or pumps).

With two and three units fueled, a worst case single failure could also include the loss of two RHRSW pumps caused by losing a 4 kV shutdown board since there are certain configurations that allow two RHRSW pumps to be'lignment

. powered from the same 4 kV shutdown board. As discussed in the FSAR, Section 14.6.3.3.2 (Ref. 4) for these analyses, manual initiation of the OPERABLE RHRSW subsystems and the associated RHR System is assumed to occur 10 mi af oooo DBA. The RHRSW flow assumed in the analyses is 4 gpm per pump with two pumps operating in one loop. In thss case, the maximum suppression chamber water temper re and pressure are 177'F (as reported in Referenc an X psig, respectively, well below the design temperature of 281'F and maximum allowable pressure of 62 psig. This is also below the 200'F limit imposed by Design Criteria BFN-50-7064A (Ref. 5) for all plant transients involving SRV ope a ions fg ~ 04 1'4c 9 Hs The RHRSW System"satisfies Criterion 3 of the NRC Policy x

Statement (Ref 6).

(continued)

BFN-UNIT 2 B 3.7-2 Amendment

OPERAQl LlTV'g ol QH5 P~ Rttasm

~ 4 ~+st A o~ li>>vessel g ~+>' g

+<>pate>>34 rt l 4 sf f g p' B 3.7.1 g gg BASES (continued)

LCO Four RHRSll subsystems are required to be OPERABLE to provide the required redundancy to ensure that the system functions to remove post accident heat loads, assuming the worst case single actfve failure occurs cofncfdent with the loss of offsite power.

An OPERABLE RHRSM subsystem consists of:

a. The required number of OPERABLE RHRSM pumps dependent upon the number of units fueled; and

b. An OPERABLE flow path capable of taking suction from the intake structure and transferring the water to the required RHR heat exchangers at the assumed flow rate.

The LCO fs modified by two Notes. Note 1 specifies that when 1 or 2 units are fueled, there must be at least one OPERABLE pump per RHRSM subsystem. Note 2 specifies that when 3 units are fueled, two of the RHRSM subsystems aust have two OPERABLE RHRSM pumps.

a~4 C($ 5 ave APPLICABILITY In NEES 1, 2, and 3, the RHRSN Syste "P4 equired to be OPERABLE to support the OPERABILITY o RHR System for primary containment cooling (LCO 3.6.2.3, 'Residual Heat Removal (RHR) Suppression Pool Cooling," and LCO 3.6.2.4,

'Residual Heat Removal (RHR) Suppression Pool Spray" ) and decay heat removal (LCO 3.4.7, "Residual Heat Removal (RHR)

Shutdown Cooling System-Hot Shutdown" ). The Applicability is therefore consistent with the requirements of these systems.

In NSES 4 and 5, the OPERABILITY re rements of the RHRSM System are determined by the system g'ports.

a~ QHS t ACTIONS The Actions are modified by a Note indicating that the applicable Conditions of LCO 3.4.7 be entered and Required Actions taken if the inoperable RHRSM subsystem results in inoperable RHR shutdow'n cooling. This is an exception to LCO 3.0.6 and ensures the proper actions are taken for these components.

(continued)

BFN-UNIT 2 B 3.7-3 Amendment

~ p RHR% System rJ 8 3.7.1 BASES ACTIONS C.l (continued)

The requisite number of subsystems and pumps must be restored to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . The 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Completion Time is based on the Completion Times provided for the RHR suppression pool cooling and ra unct'ions.

Or tha tA HS i5 and 0.2 deie. 'd '~~e--S}e If the RHRSM subsystems cannot be restored to OPERABLE status within the associated Completion Time , the unit must be placed in a NODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least HODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in NODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

SURVEILLANCE S 3.7 RE(UIREHENTS Verifying the correct alignment for each manual and power operated valve in each RHRSif subsystem flow path provides assurance that the proper flow paths will exist for RHRSQ operation. This SR does not apply to valves that are locked, sealed,. or otherwise secured in position, since these valves are verified to be in the correct position prior to locking, sealing, or securing. A valve is also allowed to be in the nonaccident position, and yet considered in the correct position, provided it can be realigned to its accident position. This is acceptable because the RHRSM System is a manually initiated system.

This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves.

The 31 day Frequency is based on engineering judgment, is consistent with the procedural controls governing valve operation, and ensures correct valve positions.

TH5 ca~

n c~t (continued~

B 3.7-5 Amendle.'.

BFN-UNIT 2

RHRSW System nd' 3.7.1 QgS BASES (continued)

REFERENCES 1. FSAR, Section 10.9.

2. FSAR, Chapter 5.

3. FSAR, Chapter 14.

4. FSAR, Section 14.6.3.3.2.

5. Design Criteria BFN-50-7064A, Primary Containment Systems - Units 2 and 3

6. NRC No.93-102, "Final Policy Statement on Technical Specification ?mprovements " Jul 23, 1993.

QneeP SR Ve~i 4:c~4>'+~ of+4e UH5 0e~pe'<>@err QHscvrs gee 'kc<<4 re~a c~pa4i'1:~y <4 i(< RATS~ 5)she~;5 ~'i." ~ 5"< ~$ &- }I&~ 4g f"g l

e~p4.'o c~ re(healg i QvcMclg o4'gL M4 i +)i'ong Ja,y;~p

.I pl;..S), gOOE.

BFN-ONIT 2 B 3.7-6 Amendment

EECW System and UHS 8 3.7.2 BASES APPLICABLE The ability of the EECM System to provide adequate cooling SAFETY ANALYSES to the identified safety equipment is an implicit assumption (continued) for the safety analyses evaluated in References I and 2.

The ability to provide onsite emergency AC power is dependent on the ability of the EECM System to cool the DGs.

The long term cooling capability of the RHR and core spray pumps is also dependent on the cooling provided by the EECM System.

The EECM System, together with the UHS, satisfy Criterion 3 of the NRC Policy Statement (Ref. 4).

LCO The EECW loops are independent of each other to the degree that each has separate controls, power supplies, and the operation of one does not depend on the other. In the event of a DBA, two EECM pumps are required to provide the minimum heat removal capability assumed in the safety analysis for the system to which it supplies cooling water. To ensure

~- this requirement is met, three EECM pumps must be OPERABLE.

At least two pumps will operate if the worst single active failure occurs coincident with the loss of offsite power.

The EECM System is considered OPERABLE when it has an OPERABLE UHS, three OPERABLE pumps, and two OPERABLE flow paths capable of taking suction from the intake structure and transferring the wat appropriate equipment.

P,~ Eacw The OPERABILITY of the U "is v' water temperature of 95'F. wdA;a; .j <.," . s> E'- u~s t.-< 8-.

~rc )ri~>'cle3 i~ SR 3. 7. l. 2.

The isolation'of the EECM System compon s may render those components or systems inoperable, but does not affect the OPERABILITY of the EECM System.

APPLICABILITY In MODES I, 2, and 3, the EECM System and UHS are required to be OPERABLE to support OPERABILITY of the equipment serviced by the EECM System. Therefore, the EECM System and UHS are required to be OPERABLE in these NODES:

In MODES 4 and 5, the OPERABILITY requirements of the EECM System and UHS are determined by the systems they support.

(continued)

BFN-UNIT 2 B 3.7-8 Amendment

SEPTEMBER 3.8, 1996 (3) Pursuanc co the Act and 10 CFR Parts 30, 40 and 70 to receive, possess and use at any time any byproduct, source and special nuclear material as sealed neutxon sources for'eactor staxtup, sealed sources for raactox instrumentation and radiation monitoring equipmenc calibration, and as fission detectors in amounts as required; r

~ ~

~

(4) pursuant to the Act and 10 cFR pares 3o, 40 an& Io, to receive, possess and use in clmouncs as required any byproduct, source or special nuclear material without. restriction to chemical or physical foxm for sample analysis or instrumenc calibration or associated with xadioaccive appax'atus or components; (5) Pursuant to the Acc and 10 CPR Parts 3o and 7o, to possess not separate, such byproduct and special nucleax'aterials as may be produced by the operation of'he facility.

~

This license shall be deemed to contain ancl is subject to the conditions specified in the following Commission regulacions in 10 CHL Chapter I: Part 20, Section 30.34 of Part: 30. Section 40.41 of part 40, sections 50.54 an& 50.$ 9 of Pact. 50, and section 70.32 of Part 70; and is subject to all applicable provisions of the Act. and to the xules. xegulations. and orders of the Commission now or hexeaf ter effect,: and is subject to the additional conditions specified or incorporated below:

The licensee is authorized to operate the facility ac steady seaee reaceor sore poser levels oor, Ia excess ofgSFs.Poepaeaccs thermal. ~3Ia5g g

The Technical Specifications contained in A~~ces A and B, as reviSed thxough Atnendmenc Ho. 206 axe hereby incorporate& in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

Xn the operation of the facility. the licensee shall, pursuant to the Federal Rater Pollution control Act Amendments of 1972 (public Law 92-500), comply with all applicable thexmal water quality standards of the State of Alabama and the United States-

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~ I BFN,;

()nit,'

I

Definitions

~

1.1 t~

I s

).1 $ nfinitions

~

(continued)

I PHYSICS TESTS PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation.

These tests are:

a. Described fn Section 13.10, Refueling Test Program; of the FSAR;

b. Authorized under the provisions of 10 CFR 50.59; or

c. Otherwise approved by the Nuclear Regulatory Commission.

3%4 RAT THERMAL POMER RTP shall be a tota1 reactor core heat transfer

{RTP rate to the reactor coolant o %It.

SH((HI GMN MARGIN (SON) SDM shall be the amount of reactivity by which the reactor is subcritical or would be subcritical assuming that: s

a. The reactor is xenon free;

b. The moderator temperature is 68 F; and

c. A11 control rods are fully inserted except for the single control rod of highest reactivity worth, which is assumed to be fully withdrawn.

lith control rods not capable of being fully inserted, the reactivity worth of these

~

'I

~ control rods must be accounted for in the determination of SQH.

STA KREO TEST BASIS A STAGGERED TEST BASIS shal1 consist of the testing of one of the systems, subsystems, channels, or other designated coaponents during the interval specified by the Surveillance

.Frequency, so that all systems, subsystems, channels, or other designated components are tested during n Surveillance Frequency intervals,

~here n is the total number of systems, subsystems, channels, or other designated components in the associated function.

(continued)

UNIT 3 I.?-5

SLC System 3.1.7 SURV ILLANCE REQUIREMENTS continued SURVEILLANCE FREQUENCY SR i 3.1.7.5 Verify the SLC conditions satisfy the following equation: 31 days (13 wt. X) (B6 gpm) (19.6 atoag)

Once within 24 where, hours after water or boron i'8 C o sodium pentaborate solution is added to the concentration (weight, percent) solution.

g pump flow rate (gpm)

E 8oron-10 enrichment (atom percent Bor on-10)

SR 3.1.7.6

~ ~ Verify each pump develops a flow rate months a 39 m at a discharge pressure I ~psig.-

SRi 3.1.7.7 Verify flow through one SLC subsystem from 18 months on a pump into reactor pressure vessel, STAGGEREQ TEST 8ASIS SR!, 3.1.7.S Verify all piping between storage tank and 18 months puap suction is unblocked.

(continued) 3.1-23

RPS Instrumentation 3.3.1.1 Table 3.3.1.1-1 (paya 1 of 3)

Reactor rrotectlca Syatea Iratnaentatlon APFLICAELE CQS) IT Ills

)RNES cR RSR) )RED REFERENCED DINER C RANEE LS FRINI SPECl F I ED PEN TRI P aSa)ISED Q4VEILLANCE AL LISLE fNCTIDR COO ITION SYSTE)I ACTI(I D o1 REISJIRE%NYS VAL1%

Intenaedlete RanOe Konl tore

a. Neutral FLus -Nish SR 3.3.1 1 ' S 'l20/125 SR 3.3.1 '1,3 dlvleione of SR 3.3.1olo5 full ecaLa SR 3.3.1.1.6 SR 3.3.1.1.9 SR 3+3.1,1,14 5(a) SR 331 ~ 'I ~ 1 S 120/125 ss 3.3-1 ~ Te4 dlvlel one of SR 3.3.1o1.9 full seal ~

SR 3 3.1i1i14

b. Inop SR 3.3.1.1+3 SR 3o3II ~ 1.14 5(e) SR 3o3o1,1.4 SR 3 3 1 I 14

2. Averaae rouer Ranee Konl tore

~. Ileutron FLus -Nfgh, 3(b) SR 3.3.'l.1.1 8 153 RTP Setdoun SR 3+3,1.1.6 SR 3.3.1.1.7 SR 3,3.1,1,13 SR 3a3o le le 16

b. Floe SIeeed Stwlatcd 3(b) SR 3.3.1.1 ~1 50 V Thermal reer -Hleh SR 3.3,1.1.2 X TP and SR 3.3.1.1.'F RTP SR 38,1.1 ~ 13 SR 3.3.1.1 ~ 16 cw '(

c. Neutron Flue -Rloh 3(b) SR 3.3.1m I +1 S 12'TP

$R 3.'3 1 ~ I o2 sR 3.3,1.1.7 sa 3.3.1.1.13 SR 3+3 o1 ~ 1 ~ 16 (continued)

( ~ ) Vlth any cantroL rod ulthdrae free a core ceLL containing ane or mre fuel aeeeahllee.

(b) Each APR)1 charnel provides inputs to bath trip ayetcm.

BFN-UNlT 3 3.3-6 Amendment

0 l,

RPS Instrumentation 3.3.1.1 Table 3.3.1.1.1 (page 2 of 3)

Reactor Protection Systea Instrunentat I on APPLICABLE CONDITIONS H(mES OR REOUIRED REFERENCED OTHER CHANNELS F Rol SPECI F I ED PER TRIP REQUIRED SURVEILLANCE ALLOMABLE FUNCTION CONDITIONS SYSTEN ACTION D. 1 REQUIRENENTS VALUE

2. Average Pouer Range Honi tora (continued)

d. Downscale SR 3.3.1.1.7 B 3X RTP SR 3.3.1.1.8 SR 3.3.1.1.14

e. Inop 1,2 SR 3.3.1.1.7 SR 3.3.1.1.8 SR 3.3.1.1.14

3. Reactor Vessel Stem 1,2 SR 3.3.1.1.1 ps ig D~ Pressure -High SR SR 3.3 1.1.8

~

3.3.1.1.10 ioqO SR 3.3.1.1.14

4. Reactor Vessel Mater 1,2 SR 3.3.1.1 ~ 1 R 538 inches Level -Lou, Level 3 SR 3.3.1.1.8 above vessel SR 3.3.1.1.13 zero SR 3.3.1.1.14

5. Hain Stem Isolation SR 3.3.1.1.8 8 10X closed Ve lve C losure SR 3.3.1.1.13 SR 3.3.1.1.14

6. Dryuell Pressure -High 1,2 SR 3.3.1.1.8 S 2.5 psig SR 3.3.1.1.'l3 SR 3.3.1.1.14 7e Scrm 0'ischarge Vol~

Mater Level -High

e. Resistance Teepereture 1,2 SR 3.3.1.1.8 8 50 gallons Detector SR 3.3.1.1.13 SR 3.3.1.1.14 5(e) SR 3.3.1.1.8 S 50 gallons SR 3.3.1.1.13 SR 3.3.1.1.14

b. Float SNitch 1,2 SR 3.3.1.1.8 S 50 gallons SR 3.3.1.1.13 SR 3.3.1.1.14 5(e) SR 3.3.1.1.8 S 50 gallons SR 3.3.1.1.13 SR 3.3.1.1.14 (continued)

(e) Mith any control rod Nithdrasn fraa e core cell containing one or sore fuel esseahlies.

BFN-UNIT 3 3.3-7 Amendment

t ATMS-RPT Instrumentation 3.3.4.2 SURVEILLANCE REQUIREMENTS continued SURVEILLANCE FREQUENCY SR .3.3.4.2.2 Perform CHANNEL FUNCTIONAL TEST. 92 days SR 3.3.4.2.3 Perform CHANNEL CALIBRATION. The 18 months Allowable Values shall be:

a. Reactor Vessel Mater Level Low Low, Level 2: a 471.52 inches above vessel zero; and

b. "

Re Steam Dome Pressure High:

w IT%~ psig.

ll~. O SR 3.3.4.2.4 Perform LOGIC SYSTEM FUNCTIONAL TEST 18 months including breaker actuation.

BFN-UNIT 3 3.3-34 Amendment

Reghn nRod Une-l - - QtOL Rod Lho

~'P .Rto s lo Rod Lhw L-2 Nolo: Opaalaa

. Nol pemlllocl 7hls l4glon la-0~k 0

/CO ce I l0 0 6 Sl 15 20 25 30 . SS 40 4S 50 56 60 65 70 75 60 65 QO QS l00 l0$

Core Row ]parceat ot ralod) 0 0

4 CI g

I i

I S/RVs 3.4.3 I

3.4 REACTOR COOLANT SYSTEM (RCS)

, 3:4.3 Safety/Relief Valves {S/RVs)

LCO i3.4.3 The safety function of lZ S/RVs shall be OPERABLE.

APPL)CABILITY: MOOES 1, 2, and 3.

ACTI NS REQUIRED ACTION COMPLETION TINE A. < One or moie required A.1 Be in MODE 3: 12 hours

S/RVs inoperable.

4.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months 0 I SVR EILLANCE REOUIREHEHTS FREQUENCY

.SR 3.4.3.1 Verify the safety function lift settings of In accordance the required S/RVs are within + 35 of the with the setpoint as fo11ows: ItLseFvice Testing Program Nvmber of Setpoint

~(RV<L ~ising/

105 'I l35 1115 f I 45 12 t)S 5 Following testing, 1ift settings shal1 be within + 1%.

(continued) 0 8$ -UNIT 3 3.4-7 A@end%en t

ECCS -Operating 3-5.1 I

r SURVEILLANCE REOUERENEHTS continued SURVE Il.LANCE FREQuENCY SR; 3.5.1.5 N PTE~ ~~~~ae'w aeaa~mu~~~

Only required to be performed prior to entering NODE 2 from NODE 3 or 4, when in NODE 4 > 48 hours.

il Verify each recirculation pump discharge 31 days valve cycles through one complete cycle of full travel.

', SR,. $ .5.1.6 Verify the Following ECCS. pumps develop the In accor'dance specified flow rate against a system head with the corresponding to the specified reactor Inservice pressure. Testing SYSTEM HEAD Program CORRESPONDING TO A VESSEl. TO 0 No.

Of PUMPS TORUS DIFFERENTIAL PRfSSURf 0 Core Spray a 6250 gpm 2 a 105 psid LPCI ~ 19,200 gpm 2 a 20 psid LPCE a 10,450 gpi a 20 psid

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I

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3.5.1.1 ma a a w a a w a e o e e ~ a m w a o NOTE ~ a a e e o o s e e a a m a ~ w a e o o Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactar steal pressure and flaw are

~ i adequate to perfarm the test.

%%&w&WA&&&$%%%&&WWwWV&%%M&w&ww&%&AA&w&&Www+

gerir with reactor pressure sig, the HPCI pump can develop e~ gand a 1CP40 92 days X

>So flow rate a 5000 gpw against a systen head correspond)n9 to reactor pressure.

(continued)

SFll-UNIT 3 3.5-5

RCIC System 3.5.3 I

SUAVE LLAHCE REQUIREMENTS SURVEILLANCE FREQUENCY SR '3.5.3.l Verify the RCIC System piping is filled 31 days with water from the pump discharge valve to the injecti on val ve.

SR .3.'5.3.2 Yerify each RClC System manual, powe~ 31 days operated, and automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position.

~

SR l3.5.3.3 <<<<<<<<<<<<<<<<<<<<<<<<<<NOTE<<<<<<<<<<>><<<<<<>>>><<<<<<<<<<<<>>>>>>

Not required to be performed until lZ hours after reactor steam pressure and flow are adequate to perform the test.

<<<<>>>>>><<<<<<>><<>><<>><<<<<<<<<<>><<<<>>>>>><<<<<<<<>><<<<\<<>>Q<<<<<<<<<<>>>>>>

~F045 Verif, w'th reactor pressure 8Q gsig an psig, the RCLC puep can develop a flow rate a 600 gpm against a system head corresponding to reactor pressure.

SR: 3,5.3.4 <<>><<>>>>>>>>>> <<>>>><<<<<<<<<<<<<<<<NOTE<<>> <<<<<<<<<<>>>><<<<>><<>> <<<<<<>>>>>>

I Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steai pressure and flow are adequate to perforo'the test.

Verify,, with reactor pressure a 165 psig, M months the RCIC pump can develop a flow rate a 600 gpss against a system head corresponding to reactor pressure.

{continued)

~ ~

I SR[-NIT 3 3.5-]3 Amendment j ~

RHRSW Syste 3.7.1

~JQ ~

3.7.1 Residual Heat Removal Service Water (RHRSW) Syst a J Ultan 3e He4 S: k (gag)

LCO 3.7. 1 NOTES

1. With 1 or 2 units fueled, each subsystem must have at least one OPERABLE RHRSW pump.

2. With 3 units fueled, two RHRSW subsystems must have two OPERABLE RHRSW pumps.

Four RHRSW subsys ems s a l be OPERABLE.

n e~d Ll HS APPLICABILITY: MODES 1, 2, and 3.

ACTIONS

- NOTE Enter applicable Conditions and Required Actions of LCO 3.4.7, "Residual Heat Removal (RHR) Shutdown Cooling System-Hot Shutdown," for RHR shutdown cooling made inoperable by RHRSW System.

CONDITION REQUIRED ACTION COMPLETION TIME C

A. One RHRSW subsystem or A.l Restore RHRSW 30 days required pump subsystem or required inoperable. pump to OPERABLE status.

(continued)

BFN-UNIT 3 3.7-1 Amendment

0 RHRSM System ~d 3 7'1 AS Q

e CONDITION REQUIRED ACTION COMPLETION TIME

0. Required Action A. 1, D. 1 Be in NODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months B.3, or C. 1 and associated Completion ANO Time not met.

0.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months og 4 85 ln appar pig.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.1.1 Verify each RHRSM manual and power operated 31 days valve in the flow path, that is not locked, sealed, or otherwise secured in position,

'is in the correct position or can be aligned to the correct position.

IAH5 ie l~oil

37. per f~g fk<<~<<~pc t it~~ i <~p+" i~<<

e I, 2, 9i P g H5 jg ~'th'n Vhe llaw '4 5'ee.'Cs eS Figure 3, 7. l I .

I gy~r Q A5 +c' li'E 3,7-3 Amendmen'.

BFN-UNIT 3

RHRSW System 3.7.1

~tc (A l~S Figure 3.7.1-1 Reactor Thermal Power Versus Ultimate Heat Sink Temperature Limit Unacceptable Acceptable 97 91 93 93.5 94 Ultknate Heat SInk Temperature (degrees F) 3I 7 BFN-UNIT 3

Jt WJ

oie.'

EECM System and UHS 3.7.2 e SURVEIL CE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.2.1 Verify the average water temperature of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months UHS is c 95'F.

SR 3.7.2.2 NOTE Isolation of flow to individual components does not render EECM System inoperable.

Verify each EECM subsystem manual and power 31 days operated valve in the flow paths servicing safety related systems or components, that is not locked, sealed, or otherwise secured in position, is in the correct position.

( SR 3.7.2.3 Verify each required EECQ pump actuates on 18 months an actual or simulated initiation signal.

BFN-UNIT 3 3.7-5 Amendment

0 1 t 0

X.

SLC System~

8 3.1'.7 BAS(S SUR EILLANCE SR 3.1.7.4 RE( lREHENTS

( ontinued) This Surveillance requires the amount of Boron-10 in the SLC solution tank to be determined every 31 days. The enriched sodium pentaborate solution is made by combining stoichiametric quantities of borax and baric acid in demineralized water, Since the chemicals used have known Boron-}0 quantities, the Boron-10 quantity in the sodium pentabar ate, solution formed can be calculated. This parameter is used as input to determine the volume requirements for SR 3.1.7.1. The 31 day Frequency of this Surveillance is appropriate because of the relatively slow variation af boron concentration between surveillances.

~R Oemonstrating thaC each SLC System um develops a flow rate m 39 gpm at a discharge pressure m sig ensures that pump performance has not degraded during the fuel cycle.

This minimum pump flow rate requirement ensures thit, when combined with the sodium pentabarate solution concentration and enrichment requirements, the rate of negative reactivity insertion from the SLC System will adequately compensate for the positive reactivity effects encountered during power reduction, cooldown of'he moderator, and xenon decay. This test confirms one point on the pump design curve ahd fs indicative of overall performance. The 18 month Frequency is acceptable since inservice testing of the pumps, performed every 92 days, will detect any adverse trends in pump per'formance.

3 .7 7 and SR 3 1.7.

These Surveillances ensure that there is a functioning flow path from the boron solution storage tank to the RPU, including the firing of an explosive valve. The replacement charge for the explosive valve shall be fram the same manufactured batch as the one fired or fram another batch that has been certified by havin9 one of that batch successfully fired. The pump and explosive valve tested should be alternated such. that both complete flow paths are tested every 36 months at a1ternating l8 month intervals.

The Surveillance may be performed in separate steps to (continued)

I 8+-UNIT 3 8 3.1-44 Amendment

Recirculatidn Laaps Operating 3.4.1 SArS AgP K

~

L?CABLE Safety ANALySpS assume cantfnued) analyses performed For FSAR Chapter 14 implicitly care canditions are stable. However, at the high pawer/law flaw corner of the pawer/flow map, an increased probability for limit cycle ascillatians exists (Ref. 3) depending on cambinations of operating conditions {e.g.,

powel shape, bundle power, and bundle flow). Generic evaluatians indicate that when regional power ascillations became detectable on the APNEA, the safety margin may be insufficient under some operating conditions to ensure actions taken to respond to the APRNs signals would prevent violation of the HCPR Safety Limit (Ref. l). NRC Generic Letter B6-02 (Ref. 5) addressed. stability calculation methodology and stated that due to uncertainties, 10 CFR 50, Appendix A, General Oesign Criteria (GOC) 10 and 12 could not be met using analytic procedures onw BlR 4 design. However, Reference 5 concluded that operating limitations which provide for the detection {by monitoring neutron Flux noise 1evels) and suppression of flux oscillations in operating regions of potential instability consistent with the recaaeendatians of Reference 3 are acceptable to demonstrate compliance with GOC 10 and 12. The NRC concluded that 0 regions of potential instability could occur at calculated decay ratios of 0.8 or greater by the General E1ectric methodology.

Stability tests at operating BNb were reviewed to determine a generic region of the power/flow map in which surveillance of neutron fTux noise levels should be performed.

decay ratio was chosen as the basis for A'onservative determining the generic region for surveillance to account for the plant to plant variability of decay ratio with core and fue1 designs. This decay ratio also helps ensure sufficient margin to an instability occurrence is maintained.

e generic region has been determined to be bounded by the rod line and the S0% core flow line. BFN conservatively o

lnts this generic region with the 'Operation Not Permitted'egion and Regions I and II of Figure 3.4.1-1.

This conforms to Reference 3 recoaaendat$ ons. Operation is permitted in Region lI provided neutron flux noise levels are verified to he within limits. The reactor mode switch must he placed in the shutdown position (an iamediate scram fs required) if Region I is entered.

Recirculation loops operating satisfies Criterion 2 of the NC Policy Statement (Ref. 6).

)FN-UNIT 3 B 3.4-3 Amendment

0 ECCS-Operating

~ I S 3.5.j

~ BASIIS BAC)G!taVNO The HPCI System is designed to lflg provide core coolin for a

{gontinued) wide r ange of reactor pressures {150 psig t psig).

C Upon receipt of an initiation signal, the HPCI turbine stop valve and turbine control valve open and the turbine accelerates to a specified speed. As the HPCI flee increases, the turbine governor valve is automatically adjusted to maintain design flow. Exhaust steam from the HPCI turbine is discharged to the suppression pool. A full flow test line is provided to route water from and to the CST to al'low testing of the HKI Systen during normal operation without injecting ~ater into the RPY.

The ECCS pumps are provided with minimum flow bypass lines, which discharge to the suppression pao1. The valves in these lines automatically open (for CS and RHR they are

. already open) to prevent pump damage due to overheating..when other discharge 1ine valves are closed. To ensure rapid delivery of water to the RPY and to minimize water haIIner effects, all ECCS pump discharge lines are filled with water. The LPCI and CS System discharge lines are kept full of water using the pressure suppression chamber head tank or condensate 'head tank. The HPCI Systea is normally aligned to the CST. The height of water in the CST is sufficient to maintain the piping full of water up to the first isolation valve. The relative height of the feedwater line connection for HPCI is such that the water in the feedwater 14nes keeps the remaining portion of the HPCI discharge line full of water.

The AQS {Ref. h) consists of 6 of the l3 S/RYs. It is designed to provide depressurization of the RCS during a small breik LOCA if HPCI fails or is unable to maintain required water level in the RPV. AOS operation reduces the RPV pressure to within the operating pressure range of the low pressure EC 'ubsystems (CS and LPCI), so that these subsysteas can pIovide coolant inventory makeup.- Each of the $/RVs used for automatic depressurization is equipped with one air accumulator and associated inlet check valves.

The accumulator provides the pneumatic power to actuate the valves.

A AMICABLE The ECCS performance is evaluated for the entire spectrum of S ETY ANALYSES break sizes for a postulated I.OCA. The accidents for which ECCS operation is required are presented in References 5 (continued)

B)N-UNIT 3 8 3.5-h Apendment

ECCS - Operating 8 3.5.1 e aags SU EILLANCE ~KR 3...5 I < dl

~

RE IREHENTS The specified Frequency is once per 31 days. However, this SR is modified by a Note that states the Surveillance is only required to be performed prior to entering HODE 2 from N)pE 3 or 4, when in NODE 4 > 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . Verification during or following each entry into NODE 4 > 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to entering %DE 2 from NODE 3 or 4 is an exception to the normal Inservice Testing Program generic valve cycling Frequency of 92 days, but is considered acceptable due to the dattonstrated reliability of these valves. The 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months is intended to indicate. an outage of sufficient duration to allow for scheduling and proper performance of the Surveillance. IF the valve is inoperable and in the open position, the associated LPCI subsystem must be declared h inoperable-R 3.5.1. SR 3.5.1 7 and SR 3.

The performance requir ements of the low pressure ECCS pumps are determined through application of the 10 CFR 50, Appendix K criteria (Ref. 7). This periodic Surveillance is performed (in accordance with the ASIDE Code,Section XI, requirements for the ECCS pumps) to verify that the ECCS pumos will develop the flow rates required by the respective ana)yses. The low pressure ECCS pump flow rates ensure that adequate core cooling is provided to satisfy the acceptance criteria of Reference 13. The pump flow rates are verified against a system head equivalent to the RPV pressure expected during a LOCA- The total system pump outlet pressure is adequate to overcome the elevation head pressure between the pump suction and the vessel discharge, the piping friction losses, and RPV pressure present during a LOCA. These values may be established by testing or analysis or during preoperational testing.

The flow tests for the HPCI System are performed at,two different pressure ranges such that system capability to provide rated flow is tested at both the higher and lower operating ranges of the system. Addit)ona11y, adequate steam flow.must be passing through the main turbine or turbine byaass valves to continue to control reactor pressure ihen the HPCI S~tem diverts steam flow. Reactor steam pressure must be v~2 psig to perform SR 3.5.2.7 and m 150 psig to perform SR 3.5-1.8. Adequate steaa flow is 9'continued)

I -UNIT 3 8 3.5-12 Amendment

RClC System 8 3.5.3

~

I B 3]5 EMERGENCY CORE COOLIN SYSTBlS (ECCS) AND REACTOR CORE ISOLATlON

'OOLING (RCIC) SYSTEH

~

B 3!S.3 RCIC System BAClfG

)OUND The RClC System is not part of the ECCS; however, the RClC

\ System is included with the ECCS section because of their I similar functions.

The RCIC System is designed to operate either automitically or manua11y fo11owing reactor pressure vessel (RPV) isolation accompanied by a loss of coolant flow from the feedwater system to provide adequate core cooling and contro1 of the RPV water 1eve1. Under these conditions, the High Pressure Coolant Injection (HPCl) and RClC .systems perform similar functions. The RCIC System design requirlnts ensure that the criteria of Reference 1 are satisfied.

The RCLC System (Ref. 2) consists of a steam driven turbine unit, piping, and valves to provide steam to the 0

pump turbine, as well as piping and valves to transfer water from the suction source to the core via the feedwater system line, ~here the coolant fs distributed within the RPV through the.feedwater sparger. Suction piping is provided from the condensate storage tank (CST) and the suppression pool. Pump suction is normally aligned to the CST to minimize in5ection of suppression poo1 water into the RPV.

However, if. the CST water supply is low, or the suppression pool level is high, a manual transfer to the suppression pooi water source ensures a water supply for continuous operation of the RCIC System- The steam supply to the turbine is piped from a main steam line upstream oF. the associated inboard main steam 1ine isolation valve.

I is designed to provide core ooli for a 74-'he RCIC System wide range of reactor pressures (150 psig to Isig).

Upon receipt of an initiation signal, the RCIC tur inc accelerates until a 600 gpm flow rate (design flow) is achieved. As the RCIC turbine flow varies, the turbine control valve is automatically a@)usted to maintain'esign flow. Exhaust steam from the RC?C turbine is discharged to the suppression pool. A full flow test line is provided to route water from and to the CST to allow testing of the RClC (continued)

B 3,5-24 Amendmen t 8%j UNIT 3

RCIC System B 3.5-3 e eas SURVEILLANCE ~(.(.. ( (( d(

REglJIREHENTS steam flaw patn for the turbine and the flow controller position.

The 31 day Frequency of this SR was derived from the Inservice Testing Program requirements for performing valve testing at least once every 92 days. The Frequency of 31 days is further justified because the valves are operated under procedural control and because improper valve .position would affect only the RCIC Systel. This Frequency has been shown to be acceptable through operating experience.

SR 3.S 3.3 and SR 3.5 3.4 The RCIC pump flow rates ensure that the system can maintain reactor coolant. inventory during pressurized conditions with the RPV isolated. The flow tests for the RCIC System are performed at two different pressure ranges such that systea-e II ~

capability to provide rated flow is tested both at the higher and lower operating ranges of the systew.

0

~

Additionally, adequate steai flow must be passing through the main turbine or turbine bypass valves to continue to control reactor pressure when the RC?C stem diverts steaa floe ee.actor steam pressure must be a si to perform SR 3.5.3.3 and a 150 psig to perfora SR 3.5. . . erefore, sufficient time is allowed after adequate pressure is achieved to performs these SRs. Reactor startup is allowed prior to performing the low pressure Surveillance because the reactor pressure is low and the time allowed to satisfactorily perform the Surveillance is short.

Alternately, the low pressure Surveillance test may be perforaed prior to startup using an auxiliary steaa supply.

The reactor pressure is allowed to be increased to dorsal operating pressure since it is assumed that the low pressure Surveillance has been satisfactorily coepleted and there is no indication or reason to believe that RCIC is inoperable.

Therefore, these SRs are modified by Notes that state the Surveillances are not required to be perforeed until I2 hours after the reactor steai pressure and flow are adequate to perform the test.

0 ~

BFNI I

NIT 3 B. 3.5-2S (continued)

Aaendment I

Primary Containment B 3.6.1.1 le The safety design basis for the primary containment is that APP CABLE SAFETY ANALYSES it must ~ithstand the pressures and temperatures of the limiting DBA without exceeding the design leakage rate.

The OBA that postulates the maximum release of radioactive material within primary containment is a LOCA. In the analysis of this accident, it is assumed that primary containment is OPERABLE such that release of fission products to the environment is controlled by the rate of primary containment 'leakage.

Analytical methods and assumptions invalving the pr)mary containment are presen)ed fn References 1 and 2. The safety analyses assume a nonmechanistic fission product release following a DBA, which Forms the basis for determination of offsite doses. The fission product release is, fn turn, based on an assumed leakage rate from the primary containment. OPERABILITY of the primary containment ensures that the leakage rate assumed in the safety analyses is not-exceeded-The maximum allowable leakage rate for the primary containment (L,) is 2.'OL by weight of the containment air per 24 bourg at the design basis LOCA maximum peak containment pressure (P,) of~~sig (Ref. l).

Sb L, Primary containment satisfies Criterion 3 of the NRC Policy Statement (Ref. 6).

~ ~

Primary containment OPERABILITY is maintained by limiting leakage to 1.0 L., except prior to the first startup after erforaing a required Primary Containaent Leakage Rate esting Program leakage test. At this time, applicable leakage liaits Iust be let. Compliance with this LCQ will ensure a pHmary containment configuration, including equipment hatches, that is structurally sound and that vill 15llit leakage to those leakage rates assumed in the safety analyses Individual leakage rates specified for the primary containment air lock are addressed in LCO 3.5.1.2.

(continued)

~

~URIT 3 B 3.6-2 Amendment

1 I Primary Containment Air Lock I

~ ~

8A GROUND and leak tightness are essentia1 for maintaining primary

{ ontinued) containment leakage rate to within limits in the event of a 08A. Hot maintaining air 1ock integrity or leak tightness may result in a leakage rate in excess of that assumed in the unit safety analysis.

I APPEICABLK The QBA that postulates the maximum release of radioactive SAFE ANALYSES material within primary containment is a LOCA. In the analysis of this accident, it is assumed that primary

.containment is OPERABLE. such that release of fission products to the environment is controlled by the rate of primary containment leakage. The primary containment is designed with a maximum allowable leakage rate (L,) of 2.0%

by weight of the containment air per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at the cale ated maximum peak containment pressure (P,) of sig (Ref. 3). This allowable leakage rate forms the sos for the acceptance criteria imposed on the SRs associated with the air lock.

Primary containment air lock OPERABILITY is also required to minimize the amount of fission product gases that may escape primary containment through the air lock and contaminate and pressurize, the secondary containment.

The primary containment air lock satisfies Criterion 3 of the NC Policy Statement (Ref. 4).

LC4 As part of prieary containment, the air lock's safety function is related to control of containment leakage rates following a OBA. Thus, the air lock's structural 4ntegrity and leak tightn..s are essential to the successful mitigation of such an event.

The primary containment air lock is required to be OPERABLE.

for the air lock to be considered OPERABLE. the air lock interlock mechanism must be OPERABLE, the air lock must be in compliance with the Type 8 air lock leakage test, and both air lock doors must be OPERABLE. The interlock allows only one air lock door to be opened at a time. This provision ensures that a gross breach of primary containment does not exist when primary containment is required to be

~ 1 (continued)

BQ-NIT B B 3.6-7 Amendment 1

O.

RHRSM System nd 4(~

B 3.7.1 B 3.7 PLANT SYSTEMS B 3.7.1 Residual Heat Removal Service Mater (RHRSM) System an~ ~ j+<~+~ (+"

BASES BACKGROUND The RHRSW System is designed to provide cooling water for

. the Residual Heat Removal (RHR) System heat exchangers, required for a safe reactor shutdown following a Design

'o Basis Accident (DBA) or transient. The RHRSW System is operated whenever the RHR heat exchangers are required to operate in the shutdown cooling mode or in the suppression pool cooling or f the RHR System.

the P HS a~d The RHRSW Sy tern the three BFN units and consists of four independent and redundant loops, each of which feeds one RHR heat exchanger in each unit. Each loop is made up of a header, two 4500 gpm pumps, a suction source, valves, piping, and associated instrumentation. One loop with one pump operating is capable of providing 5K of the required cooling capacity to maintain safe shutdown conditions for one unit following a design basis accident.

However, one pump is capable of providing sufficient cooling capacity to maintain a safe shutdown condition for each of the non-accident units. As such, a subsystem consists of a loop with one or two OPERABLE pumps dependent upon the number of fueled units. The RHRSW System is designed with sufficient redundancy so that no single active component failure can prevent it from achieving its design function.

The RHRSW System is described in the FSAR, Section 10ag (Ref. 1).

Cooling water is pumped by the RHRSW pumps from the Wheeler Reservoir through the tube side of the RHR heat exchangers, and discharged back to the Wheeler Reservoir.

The system is initiated manually from each of the three units control rooms. If operating during a loss of coolant accident (LOCA), the system is automatically tripped on degraded bus voltage to allow the diesel generators to automatically power only that equipment necessary to reflood the core. The system can be manually started any time the degraded bus voltage signal is manually overridden or clears, and is assumed to be manually started within 10 minutes after the LOCA.

(continued'mendmen.

BFN-UNIT 3 8 3.7-1

RHRSW System nd 3.7.1 B

~ pq BASES (continued)

APPLICABLE The RHRSM System removes heat from the suppression pool to SAFETY ANALYSES limit the suppression pool temperature and primary containment pressure following a LOCA. This ensures that the primary, containment can perform its function of limiting the release of radioactive materials to the environment following a LOCA. The ability of the RHRSW System to support long term cooling of the reactor or primary containment is discussed in the FSAR, Chapters 5 and 14 (Refs. 2 and 3, respectively). These analyses explicitly assume that the RHRSW System will provide adequate cooling support to the equipment required for safe shutdown. These analyses include the evaluation of the long term primary containment response after a design basis LOCA.

The safety analyses for long term cooling were performed for various combinations of RHR System failures and considers the number of units fueled. With one unit fueled, the worst case single'failure that would affect the performance of the RHRSM System is any failure that would disable two subsystems or pumps of the RHRSM System (e.g, the failure of an RHR Suppression Pool Cooling/Spray return line valve which effectively disables two RHRSW subsystems or pumps).

With two and three units fueled, a worst case single failure could also include the loss of two RHRSM pumps caused by losing a 4 kV shutdown board since there are certain alignment configurations that allow two RHRSM pumps to be powered from the same 4 kV shutdown board. As discussed in the FSAR, Section 14.6.3.3.2 (Ref. 4) for these analyses, manual initiation of the OPERABLE RHRSW subsystems and the Rouo associated RHR System is assumed to occur 10 mi af DBA. The RHRSM flow assumed in the analyses is gpm per pump with two pumps operating in one loop. In this case, the maximum suppression chamber water temper re and pressure are 177'F (as reported in Reference and

%HER psig, respectively,

~ well below the design temperature of 281'F ~

and maximum allowable pressure of 62 psig. This is also below the 200'F ~

limit imposed by Design Criteria BFN-50-7064A (Ref. 5) for all lant transients involving SRV op rations',$ g, il, glg/5 The RHRSM System"satisfies Criterion 3 of the NRC Policy Statement (Ref 6).

(continued) 3.7-2 Amendme l".

BFN-UNIT 3 B

RAS(LIT'g ~f gee QH5 F Rg~~ 7a

'b g d 0 4 p cg

~gp ~) f RHRSM System <nd

g. T, f -/

I~

'the t>~ ~ ~> <(ceo P~'r2 i~ Ft'panic 8 3.7.1 ~85-B ES (continued)

CO Four RHRSW subsystems are required to be OPERABLE to provide the required redundancy to ensure that the system functions to remove post accident heat loads, assuming the worst case single active failure occurs coincident with the loss of offsite power.

An OPERABLE RHRSW subsystem consists of:

a. The required number of OPERABLE RHRSW pumps dependent upon the number of units fueled; and

b. An OPERABLE flow path capable of taking suction from the intake structure and transferring the water to the required RHR heat exchangers at the assumed flow rate.

The LCO is modified by two Notes. Note 1 specifies that when 1 or 2 units are fueled, there must be at least one OPERABLE pump per RHRSW subsystem. Note 2 specifies that-when 3 units are fueled, two of the RHRSW subsystems must have two OPERABLE RHRSW pumps.

gad F405 APPLICABILITY In HOOES 1, 2, and 3, the RHRSW Syste 'quired to be )(

OPERABLE to support the OPERABILITY of e RHR System for primary containment cooling (LCO 3.6.2.3, "Residual Heat Removal (RHR) Suppression Pool Cooling," and LCO 3.6.2.4, "Residual Heat Removal (RHR) Suppression Pool Spray" ) and decay heat removal (LCO 3.4.7, "Residual Heat Removal (RHR)

Shutdown Cooling System'-Hot Shutdown" ). The Applicability is therefore consistent with the requirements of these systems.

In HOOES 4 and 5, the OPERABILITY re ements of the RHRSW Syst a termined by the system 'g upports.

coed QH5 <4y I

ACTIONS The Actions are modified by a Note indicating that the applicable Conditions of LCO 3.4.7 be entered and Required Actions taken if the inoperable RHRSW subsystem results in inoperable RHR shutdown cooling. This is an exception to LCO 3.0.6 and ensures the proper actions are taken for these components.

(continued)

BFN-UNIT 3 B 3.7-3 Amendment

i 0

RHRSW System ~d B 3.7.1 ACTIONS C. 1 (continued)

The requisite number of subsystems and pumps must be restored to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . The 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Completion Time is based on the Completion Times provided for the RHR suppression pool cooling and spray functions.

D. 1 and 0.2 Os the QHS ~~

Is Qc~e~<ACA iradsp eem tile If the RHRSW subsystems cannot be restored to OPERABLE status within the associated Completion Time the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating =

experience, to reach the required unit conditions from full t

power conditions in an orderly manner and without challenging unit systems.

SURVEILLANCE 3.7.1.1 REQUIREMENTS Verifying the correct alignment for each manual and power operated valve in each RHRSW subsystem flow path provides assurance that the proper flow paths will exist for RHRSW operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these valves are verified'o be in the correct position prior to locking, sealing, or securing. A valve is also allowed to be in the nonaccident position, and yet

'Rconsidered in the correct position, provided it can be realigned to its accident position. This is acceptable because the RHRSW System is a manually initiated system.

This rather, SR does not require any testing or valve manipulation; it involves verification that those valves capable of being mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves.

The 31 day Frequency is based on engineering judgment, is consistent with the procedural controls governing valve operation, and ensures correct valve positions.

>ra S ewf

'Prowrates t' 4)C ~

(continued)

BFN-UNIT 3 B 3.7-5 Amendment

~ '

RHRSM System ~~ 0 B 3.7.1

~H~

BASES (continued)

REFERENCES 1. FSAR, Section 10.9.

2. FSAR, Chapter 5.

3. FSAR, Chapter 14.

4. FSAR, Section 14.6.3.3.2.

5. Oesign Criteria BFN-50-7064A, Primary Containment Systems - Units 2 and 3

6. NRC No.93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.

0 e},ical Z ~Berk Sp p,7 V'Rl I)I c RW Idvi (jP tgc 0HS ke~)er~ }uie enzLlvc'a i '4q< tLe hei< r em ovh I Ca Pc 4 '1 '<y a< eh. RHR>~ Sys~~~ <> ~ >l.; th ASS~~~>'~~a aF Vk< GO>

@we>yS'i I he' 9 ho~~ c~J t ho< r Fre gui~ r i'rr p we gage 3 a~ a)erik>'n~

Q)( PI 0Nce

$ re }c>' "c ><<~~i'~y ~< >~t 4 r s~ c ~ev'ar lR } I i%) JKrl J ~"e P

ap pj'~.4}, l'fobES.

BFN-UNIT 3 B 3.7-6 Amendne

EECW System and UHS B 3.7.2 BASES APPLICABLE The ability of the EECW System to provide adequate cooling SAFETY ANALYSES to the identified safety equipment is an implicit assumption (continued) for the safety analyses evaluated in References I and 2.

The ability to provide onsite emergency AC power is dependent on the ability of the EECW System to cool the DGs; The long term cooling capability of the RHR and core spray pumps is also dependent on the cooling provided by the EECW System.

The EECW System, together with the UHS, satisfy Criterion 3 of the NRC Policy Statement (Ref. 4).

LCO The EECW loops are independent of each other to the degree that each has separate controls, power supplies, and the operation of one does not depend on the other. In the event of a DBA, two EECW pumps are required to provide the minimum heat removal capability assumed in the safety analysis for the system to which it supplies cooling water. To ensure this requirement is met, three EECW pumps must be OPERABLE.

At least two pumps will operate if the worst single active failure occurs coincident with the loss of offsite power.

The EECM System is considered OPERABLE when it has an OPERABLE UHS, three OPERABLE pumps, and two OPERABLE flow paths capable of taking suction from the intake structure and transferring the wa r o ppropriate equipment.

pBEcw '

The OPERABILITY of the S" is b s max um water temperature of 95 QJJi>4~I <~guirc~e~4s ~>y >HS i'c'~pc~~< e PryvileA i~ Sg '3. 7. I. 2.,

The isolation of the EECW System n may render those components or systems inoperable, but does not affect the OPERABILITY of the EECW System.

APPLICABILITY In MODES I, 2, and 3, the EECW System and UHS are required to be OPERABLE to support OPERABILITY of the equipment serviced by the EECW System. Therefore, the EECW System and UHS are required to be OPERABLE in these NODES.

In NODES 4 and 5, the OPERABILITY requirements of the EECM System and UHS are determined by the systems they support.

(continued)

BFN-UNIT 3 B 3.7-8 Amendment

ENCLOSURE 4 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)

UNITS 2 AND 3 PROPOSED TECHNICAL SPECIFICATION (TS) CHANGE TS-384 REVISED PAGES I. AFFECTED PAGE LIST The following pages have been revised. A revision bar has been placed in the left hand margin to indicate where changes occur.

0 eratin License Unit 2 Unit 3 3 3 Technical S ecifications Bases Unit 2 Unit 3 Unit 2 Unit 3 1.1-5 1.1-5 B 3.1-44 B 3. 1-44 3 1 23

~ 3 1 23

~ B 3.4-3 B 3. 4-3

3. 3-6 3.3-6 B 3.5-4 B 3. 5-4 3 3 7

~ 3 3 7

~ B 3.5-12 B 3. 5-12 3.3-34 3.3-34 B 3.5-24 B 3. 5-24 3.4-4 3.4-4 B 3.5-28 B 3. 5-28 3.4-7 3.4-7 B 3.6-2 B 3. 6-2 3.5-5 3.5-5 B 3.6-7 B 3. 6-7 3.5-13 3.5-13 B 3.7-1 B 3. 7-1 307 1 307 1 B 3.7-2 B 3. 7-2 3 7 3

~ 3 7 3

~ B 3.7-5 B 3. 7-5 3 s 7 3a 3 0 7 3a B 3.7-6 B 3. 7-6

3. 7-5 3.7-5 B 3.7-8 B 3. 7-8 II. REVISED PAGES See attached.

(2) Pursuant to the Act and 10 CFR Parts 40 and 70, to receive, possess, and use at any time source and special nuclear material as reactor fuel in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report as supplemented and amendedg (3) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use at any time any byproduct, source, and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use in amounts as required any byproduct, source, or special nuclear material without restriction to chemical or physical form for sample analysis or equipment and instrument calibration or associated with radioactive apparatus or components; (5) Pursuant to the Act and 10 CFR Parts 30 and 70, to possess but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C. This license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter Is Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is sub)ect to the additional conditions specified or incorporated below:

(1) Maximum Power Level The licensee is authorised to operate the facility at steady state reactor core power levels not in excess of 3458 megawatts thermal.

(2) Technical- S ecifications The Technical Specifications contained in Appendices A and 8, as revised through Amendment No. 248, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

Unit 2

Definitions 1.1

l. 1 Definitions (continued)

PHYSICS TESTS PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation.

These tests are:

a. Described in Section 13.10, Refueling Test Program; of the FSAR;

b. Authorized under the provisions of 10 CFR 50.59; or

c. Otherwise approved by the Nuclear Regulatory Commission.

RATED THERMAL POWER RTP shall be a total reactor core heat transfer

( RT P ) rate to the reactor coolant of 3458 MWt.

SHUTDOWN MARGIN (SDM) SDM shall be the amount of reactivity by which the reactor is subcritical or would be subcritical assuming that:

a. The reactor is xenon fr ee;

b. The moderator temperature is 68'F; and

c. All control rods are fully inserted except for the single control rod of highest reactivity worth, which is assumed to be. fully withdrawn.

With control rods not capable of being fully inserted, the reactivity worth of these control rods must be accounted for in the determination of SDM.

STAGGERED TEST BASIS A STAGGERED TEST BASIS shall consist of the testing of one of the systems, subsystems, channels, or other designated components during the interval specified by the Surveillance Frequency, so that all systems, subsystems, channels, or other designated components are during n Surveillance Frequency intervals,

'ested where n is the total number of systems, subsystems, channels, or other designated components in the associated function.

(continued)

BFN-UNIT 2 1.1-5 Amendment

SLC System 3.1.7 SURVEILLANCE REQUIREMENTS continued SURVEILLANCE FREQUENCY SR 3. 1.7.5 Verify the SLC conditions satisfy the following equation: 31 days AND (13 wt. X) (86 gpm) (19.8 atomX)

Once within 24 where, hours after water or boron C - sodium pentaborate solution is added to the concentration (weight percent) solution ~

pump flow rate (gpm)

E = Boron-10 enrichment (atom percent Boron-10)

'R 3.1.7.6 Verify each pump develops a flow rate

~ 39 gpm at a discharge pressure 18 months

~ ~ 1325 psig.

SR 3.1.7.7 Verify flow through one SLC subsystem from 18 months on" a pump into reactor pressure vessel. STAGGERED TEST BASIS SR 3. 1.7.8 Verify all piping between storage tank and 18 months pump suction is unblocked.

(continued)

BFN-UNIT 2 3.1-23 Amendment

RPS Instrumentation 3.3.1.1 Table 3.3.1.1-1 (page 1 of 3)

Reactor Protection System Instruaentation APPLICABLE COND I T I ON S HOOES OR REQUIRED REFERENCED OTHER CHANNELS FROH SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLNIABLE FUNCTION CONDITIONS STSTEH ACTION D.1 REQUIREHENTS VALUE

1. Intermediate Range Honitors

a. Neutron Flux--High SR 3.3.1.1.1 5 120/125 SR 3.3.1.1.3 divisions of SR 3.3.1.1.5 full scale SR 3.3.1.1.6 SR 3.3.1.1.9 SR 3.3.1.1.14 5(a) SR 3.3.1.1.1 5 120/125 SR 3.3.1.1.4 divisions of SR 3.3.1.1.9 full scale SR 3.3.1.1 14

~

b. Inop SR 3.3.1.1.3 NA SR 3.3.1.1.14 5(a) SR 3.F 1.1.4 NA SR 3.3.1.1.14

2. Average Pouer Range Honitors

a. Neutron Flux High, SR 3.3.1.1.1 < 15X RTP Setdown SR 3.3.1.1.3 SR 3.3.1.1.6 SR 3.3.F 1.7 SR 3.3.1.1.9 SR 3.3.1.1.14

b. Fiou Biased Simulated SR 3.3.1 1.1

~ 5 0.58 M Thermal Pomr -High SR 3.3.1.1.2 + 66X RTP and SR 3.3.1.1.7 < 120X RTP SR 3.3.1.1.8 SR 3.3.1.1.9 SR 3.3.F 1.11 SR 3.3.1.1.14

c. Neutron Flux -High SR 3.3.1.1.1 S 120X RTP SR 3.3.1.1.2 SR 3.3.1.1.7 SR 3.3.1.1.8 SR 3.3.1.1.9 SR 3.3.1 1.14

~

(cont inued)

(a) Mith any control rod uithdraun from a core cell containing one or more fuel assemblies.

BFN-UNIT 2 3.3-6 Amendment

RPS Instrumentation 3.3.1.1 Table 3.3.1.1-1 (page 2 of 3)

Reactor Protection Systea Instrlsaentati on APPLICABLE COND IT IOHS HOOES OR REQUIRED REFERENCED OTHER CHANNELS FROH SPECI FIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION COND I T IOIS SZSTEH ACTION D.1 REOUIREHENTS VALUE

2. Average Power Range Honitors (continued)

d. Downscale SR 3.3.1.1.7 2 3X RTP SR 3.3 ~ 1 ~ 1.8 SR 3.3.1.1.14 em lnop 1,2 SR 3.3.1.1.7 SR 3.3.1.1.e SR 3.3.1.1 '4

3. Reactor Vessel Steasl II2 SR 3.3.1.1.1 S 1090 psig Dane Pressure -High SR 3.3.1.1.8 SR 3.3.1.1 10 ~

SR 3.3.1 1.14

~

4. Reactor Vessel Water 1,2 SR 3.3 ~ 1 ~ 1.1 2 538 inches Level -Lou, Level 3 SR 3.3.1.1.8 above vessel SR 3.3.1.1.13 tel'0 SR 3.3.1.1.14

5. Hain Steam Isolation SR 3.3 ~ 1 ~ 1.8 s 10X closed Valve Closure SR 3.3.1.1.13 SR 3.3.1.1 14

6. Drywall Pressure -High I~2 SR 3.3.1.1.e < 2.5 psig SR 3.3.1.1.13 SR 3.3.1.1.14

7. Scram Discharge Volune Mater Level -High

a. Resistance Tesperature 1,2 SR 3.3.1.1.8 5 50 gallons Detector SR 3.3.1.1.13 SR 3.3.1.1.14 5(a) 2 H SR 3.F 1.1.8 S 50 gallons SR 3.3.1.1.13 SR 3.3.1.1.14

b. Float Switch 1,2 SR 3.3.1.1.8 5 50 gallons SR 3.3.1.1 13 ~

SR 3.3.1.1.14 5(a) SR 3.3.1.1.8 5 50 gallons SR 3.3.1.1.13 SR 3.3.1.1.14 (cont inued)

(a) With any control rod withdralal fros a core cell containing one or more fuel assemblies.

BFN-UNIT 2 313 7 Amendment

ATWS-RPT Instrumentation 3.3.4.2 SURVEILLANCE REQUIREMENTS continued SURVEILLANCE FREQUENCY SR 3.3.4.2.2 Perform CHANNEL FUNCTIONAL TEST. 92 days SR 3.3.4.2.3 Perform CHANNEL CALIBRATION. The 18 months Allowable Values shall be:

a. Reactor Vessel Water Level Low Low, Level 2: a 471.52 inches above vessel zero; and

b. Reactor Steam Dome Pressure High:

a 1175.0 psig.

SR 3.3.4.2.4 Perform LOGIC SYSTEM FUNCTIONAL TEST 18 months including breaker actuation.

BFN-UNIT 2 3.3-34 Amendment

c I

'n z

BFN Power/Flow Stability Regions Region Region I II 10 2.9% R od Lin 100.0 % Rod Line O 70.00 96.2'ne

/o Rod L 0 ~O 0

Ill K 76 2% Rod Line

~ Ill 6o.oo 0Z O CL

< ~~Z Note: Operation ol Not Permitted In D y0o U g 5o.oo This Region

~-m~ I-(~ 40.00 a~

Ill 30.00 0 M zM Natural Circulation Une 10.00 0.00 0 5 10 15 20 25 30 35 40 45 50 55 60 65 70 75 80 85 90 95 100 105 Core Flow (% of rated)

i 3.4 3.4.3 LCO REACTOR COOLANT SYSTEH (RCS)

Safety/Relief Valves (S/RVs) 3.4.3 The safety function of 12 S/RVs shall be OPERABLE.

S/RVs

3.4.3 APPLICABILITY

HODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.l Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months S/RVs inoperable.

AND A.2 in 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months t

Be MODE SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.3. Verify the safety function lift settings of In accordance 1

the required S/RVs are within setpoint as follows:

i of the

31. with the Inservice Testing Program Number of Setpoint

~SRVs ~siqi 1135 1145 1155 Following testing, lift settings shall be within i 1%.

(continued)

BFN-UNIT 2 3.4-7 Amendment

ECCS -Operating 3.5.1 SURVEILLANCE REQUIREMENTS continued SURVEILLANCE FREQUENCY SR 3.5.1.5 NOTE--

Only required to be performed prior to entering MODE 2 from MODE 3 or 4, when in MODE 4 > 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

Verify each recirculation pump discharge 31 days valve cycles through one complete cycle of full travel.

SR 3.5.1.6 Verify the following ECCS pumps develop the In accordance specified flow rate against a system head with the corresponding to 'the specified reactor Inservice pressure. Testing SYSTEM HEAD Program CORRESPONDING TO A VESSEL TO NO. TORUS OF DIFFERENTIAL SYSTEM FLOW RATE PUMPS PRESSURE OF Core Spray )~ 6250 gpm 2 a

)~ 10520 psid LPCI 19,200 gpm 2 psid LPCI ~ 10,450 gpm 1 20 psid SR 3.5.1.7 NOTE Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure and flow are adequate to perform the test.

Verify, with reactor pressure a 1040 and 92 days

~ 950 psig, the HPCI pump can develop a flow rate a 5000 gpm against a system head corresponding to reactor pressure.

(continued)

BFN-UNIT 2 3.5-5 Amendment

RCIC System 3.5.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.3. 1 Verify the RCIC System piping is fill ed 31 days with water from the pump discharge valve to the injection valve.

SR 3.5.3.2 Verify each RCIC System manual, power 31 days operated, and automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position.

SR 3.5.3.3 -NOTE Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure and flow are adequate to perform the test.

Ol 0

Verify, with reactor pressure z 1040 psig 92 days and a 950 psig, the RCIC pump can develop a flow rate a 600 gpm against a system head corresponding to reactor pressure.

SR 3.5.3.4 -NOTE Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure and flow are adequate to perform the test.

Verify, with reactor pressure a 165 psig, 18 months the RCIC pump can develop a flow rate a 600 gpm against a system head corresponding to reactor pressure.

(continued)

BFN-UNIT 2 3.5-13 Amendment

RHRSM System and UHS 3.7.1 3.7 PLANT SYSTEMS 3.7.1 Residual Heat Removal Service Mater (RHRSM) System and Ultimate Heat Sink (UHS)

LCO 3.7.1 NOTES

1. Mith 1 or 2 units fueled, each subsystem must have at least one OPERABLE RHRSM pump.

2. Mith 3 units fueled, two RHRSW subsystems must have two OPERABLE RHRSM pumps.

Four RHRSW subsystems and UHS shall be OPERABLE.

APPLICABILITY: MOOES 1, 2, and 3.

I ACTIONS NOTE Enter applicable Conditions and Required Actions of LCO 3.4.7, "Residual Heat Removal (RHR) Shutdown Cooling System-Hot Shutdown," for RHR shutdown cooling made inoperable by RHRSM System.

CONDITION RE(UIRED ACTION COMPLETION TINE A. One RHRSM subsystem or A.l .Restore RHRSM 30 days required pump subsystem or required inoperable. pump to OPERABLE status.

(continued)

BFN-UNIT 2 3.7-1 Amendment

RHRSW System and UHS 3.7.1 CONDITION RE(UIRED ACTION COMPLETION TIME B. Two RHRSW subsystems B. I --------NOTE--------

inoperable. Only applicable when

- two units are fueled.

Verify at least three I hour OPERABLE RHRSM pumps are associated with two OPERABLE RHRSM subsystems.

B.2 --------NOTE--------

Only applicable when three units are fueled.

Verify at least four I hour OPERABLE RHRSW pumps are associated with two OPERABLE RHRSW subsystems.

~AD B.3 Restore one 7 days inoperable RHRSW subsystem to OPERABLE status.

E C. Three or more RHRSM C.1 Restore the required B hours subsystems inoperable. RHRS'W "subsystems to OPERABLE status.

+0 F

Required Action B.l or B.2 and associated Completion Time not met.

(continued)

BFN-UNIT 2 3&7 2 Amendment

RHRSM System and UHS 3.7.1 CONDITION RE(U I RED ACTION COMPLETION TIME D. Required Action A. 1, 0.1 Be in MODE 3. 12 hours 8.3, or C.l and associated Completion ~ND Time not met.

D.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months OR UHS inoperable.

SURVEILLANCE RE(UIREMENTS SURVEILLANCE FREQUENCY SR 3.7.1.1 Verify each RHRSQ manual and power operated 31 days valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position or can be aligned to the correct position.

SR 3.7.1.2 Verify the average water temperature of UHS 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months UHS is within the limits specified in Figure temperature <

3.7.1-1. 91'F AND 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months UHS temperature )

91'F BFN-UNIT 2 3t7 3 Amendment

0 RHRSW System and UHS 3.7.1 Figure 3.7.14 Reactor Thermal Power Versus Ultimate Heat Sink Temperature Limit 101 Unacceptable Acceptable 97 91 93 93.5 94 Ultimate Heat Sink Temperature (degrees F) 3.7-3a Amendment

EECW System and UHS 3.7.2 3.7 PLANT SYSTEMS 3.7.2 Emergency Equipment Cooling Water (EECW) System and Ultimate Heat Sink (UHS)

LCO 3.7.2 The EECW System with three pumps and UHS shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One required EECW pump A.1 Restore the required 7 days inoperable. EECW pump to OPERABLE status.

B. Required Action and B.l Be in MODE 3. 12 hours associated Completion Time of Condition A AND not met.

8.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months OR Two or more required EECW pumps inoperable.

OR UHS inoperable.

BFN-UNIT 2 3.7-4 Amendment

Il 0

EECW System and UHS 3.7.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.2.1 NOTE Refer to SR 3.7. 1.2 for additional UHS requirements.

I I

Verify the average water temperature of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months UHS i s w 95'F.

SR 3.7.2.2 NOTE Isolation of flow to individual components does not render EECW System inoperable.

Verify each EECW subsystem manual and power 31 days operated valve in the flow paths servicing safety related systems or components, that is not locked, sealed-, or otherwise secured in position, is in the correct position.

SR 3.7.2.3 Verify each required EECW pump actuates on 18 months an actual or simulated initiation signal.

BFN-UNIT 2 3.7-5 Amendment

SLC System B 3.1.7 SURVEILLANCE SR 3.1.7.4 RE(UIREMENTS (continued) This Surveillance requires the amount of Boron-10 in the SLC solution tank to be determined every 31 days. The enriched sodium pentaborate solution is made by combining stoichiometric quantities of borax and boric acid in demineralized water. Since the chemicals used have known Boron-10 quantities, the Boron-10 quantity in the sodium pentaborate solution formed can be calculated. This parameter is used as input to determine the volume requirements for SR 3. 1.7. 1. The 31 day Frequency of this

, Surveillance is appropriate because of the relatively slow variation of boron concentration between surveillances.

SR 3.1.7.6 Demonstrating that each SLC System pump develops a flow rate

> 39 gpm at a discharge pressure z 1325 psig ensures that pump performance has not degraded during the fuel cycle.

This minimum pump flow rate requirement ensures that, when combined with the sodium pentaborate solution concentration and enrichment requirements, the rate of negative reactivity insertion from the SLC System will adequately compensate for the positive reactivity effects encountered during power reduction, cooldown of the moderator, and xenon decay. This test confirms one point on the pump, design curve and is indicative of overall performance. The 18 month Frequency is acceptable since inservice testing of the pumps, performed every 92 days, will detect any adverse trends in pump performance.

SR 3.1.7.7 and SR 3.1.7.8 These Surveillances ensure that there is a functioning flow path from the boron solution storage tank to the RPV, including the firing of an explosive valve. ,The replacement charge for the explosive valve shall be from the same manufactured batch as the one fired or from another batch that has been certified by having one of that batch successfully fired. The pump and explosive valve tested should be alternated such that both complete flow paths are tested every 36 months at alternating 18 month intervals.

The Surveillance may be performed'n separate steps to (continued)

BFN-UNIT 2 B 3.1-44 Amendment

Recirculati'on Loops Operating B 3.4.1 APPLICABLE Safety analyses performed for FSAR Chapter 14 implicitly SAFETY ANALYSES assume core conditions are stable. However, at the high (continued) power/low flow corner of the power/flow map, an increased probability for limit cycle oscillations exists (Ref. 3) depending on combinations of operating conditions (e.g.,

power shape, bundle power, and bundle flow). Generic evaluations indicate that when regional power oscillations become detectable on the APRNs, the safety margin may be insufficient under some operating conditions to ensure actions taken to respond to the APRHs signals would prevent violation of the HCPR Safety Limit (Ref. 4). NRC Generic Letter 86-02 (Ref. 5) addressed stability calculation methodology and stated that due to uncertainties, 10 CFR 50, Appendix A, General Design Criteria (GDC) 10 and 12 could not be met using analytic procedures on a BWR 4 design.

However, Reference 5 concluded that operating limitations which provide for the detection (by monitoring neutron flux noise levels) and suppression of flux oscillations in operating regions of potential instability consistent with the recommendations of Reference 3 are acceptable to demonstrate compliance with GDC 10 and 12. The NRC concluded that regions of potential instability could occur at calculated decay ratios of 0.8 or greater by the General Electric methodology.

Stability tests at operating BWRs were reviewed to determine a generic region of the power/flow map in which surveillance of neutron flux noise levels should be performed. A conservative decay ratio was chosen as the basis for determining the generic region for surveillance to account for the plant to plant variability of decay ratio with core and fuel designs. This decay ratio also helps ensure sufficient margin to an instability occurrence is maintained. The generic region has been determined to be bounded by the 76.2'od line and the 455 core flow line.

BFN conservatively implements this generic region with the "Operation Not Permitted" Region and Regions I and II of Figure 3.4. I-I. This conforms to Reference 3 recommendations. Operation is permitted in Region II provided neutron flux noise levels are verified to be within limits. The reactor mode switch must be placed in the shutdown position (an immediate scram is required) if Region I is entered.

Recirculation loops operating satisfies Criterion 2 of the NRC Policy Statement (Ref. 6).

{continued)

BFN-UNIT 2 B 3.4-3 Amendment

ECCS -Operating B 3.

5.1 BACKGROUND

The HPCI System is designed to provide core cooling for a (continued) wide range of reactor pressures (150 psig to 1174 psig).

Upon receipt of an initiation signal, the HPCI turbine stop valve and turbine control valve open and the turbine accelerates to a specified speed. As the HPCI flow increases, the turbine governor valve is automatically adjusted to maintain design flow. Exhaust steam from the HPCI turbine is discharged to the suppression pool. A full flow test line is provided to route water from and to the CST to allow testing of the HPCI System during normal operation without injecting water into the RPV.

The ECCS pumps are provided with minimum flow bypass lines, which discharge to the suppression pool. The valves in these lines automatically open (for CS and RHR they are already open) to prevent pump damage due to overheating when other discharge line valves are closed. To ensure rapid delivery of water to the RPV and to minimize water hammer effects, all ECCS pump discharge lines are filled with water. The LPCI and CS System discharge 'lines are kept full of water using the pressure suppression chamber head tank or condensate head tank; The HPCI System is normally aligned to the CST. The height of water in the CST is sufficient to maintain the piping ful] of water up to the first isolation valve. The relative height of the feedwater line connection for HPCI is such that the water in the feedwater lines keeps the remaining portion of the HPCI discharge line full of water.

The ADS (Ref. 4) consists of 6 of the 13 S/RVs. It is designed to provide depressurization of the RCS during a small break LOCA if HPCI fails or is unable to maintain required water level in the RPV. ADS operation reduces the RPV pressure to within the operating pressure range of the low pressure ECCS subsystems (CS and LPCI), so that these subsystems can provide coolant inventory makeup. Each of the S/RVs used for automatic depr'essurization is equipped with one air accumulator and associated inlet check valves.

The accumulator provides the pneumatic power to actuate the valves.

APPLICABLE The ECCS performance is evaluated for the entire spectrum of SAFETY ANALYSES break sizes for a postulated LOCA. The accidents for which ECCS operation is required are presented in References 5 (continued)

BFN-UNIT 2 B 3.5-4 Amendment

ECCS -Operating B 3.5.1 BASES SURVEILLANCE SR 3.5. 1.5 (continued)

RE(UIREMENTS The specified Frequency is once per 31 days. However, this SR is modified by a Note that states the Surveillance is only required to be performed prior to entering MODE 2 from MODE 3 or 4, when in HODE 4 > 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . Verification during or following each entry into MODE 4 > 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to entering MODE 2 from HODE 3 or 4 is an exception to the normal Inservice Testing Program generic valve cycling Frequency of 92 days, but is considered acceptable due to the demonstrated reliability of these valves. The 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months is intended to indicate an outage of sufficient duration to allow for scheduling and proper performance of the Surveillance. If the valve is, inoperable and in the open position, the associated LPCI subsystem must be declared inoperable.

SR 3.5.1.6 SR 3.5.1.7 and SR 3.5.1.8 The performance requirements of the low pressure ECCS pumps are determined through application of the 10 CFR 50, Appendix K criteria (Ref. 7). This periodic Surveillance is performed (in accordance with the ASME Code,Section XI, requirements for the ECCS pumps) to verify that the ECCS pumps will develop the flow rates required by the respective analyses. The low pressure ECCS pump flow rates ensure that adequate core cooling is provided to satisfy the acceptance criteria of Reference 13. The pump flow rates are verified against a system head equivalent to the RPV pressure expected during a LOCA. The total system pump outlet

'ressure is adequate to overcome the elevation head pressure between the pump suction and the vessel discharge, the piping friction losses, and RPV pressure present during a LOCA. These values may be established by testing or analysis or during preoperational testing.

The flow tests for the HPCI System are performed at two different pressure ranges such that system capability to provide rated flow is tested at both the higher and lower operating ranges of the system. Additionally, adequate steam flow must be passing through the main turbine or

'urbine bypass valves to continue to control reactor pressure when the HPCI System diverts steam flow. Reactor steam pressure must be ~ 950 psig to perform SR 3.5. 1.7 and

~ 150 psig to perform SR 3.5. 1.8. Adequate steam flow is (continued)

BFN-UNIT 2 B 3.5-12 Amendment

RCIC System B 3.5.3 SYSTEMS' 8 3.5 EMERGENCY CORE COOLING (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEH B 3.5.3 RCIC System BASES BACKGROUND The RCIC System is not part of the ECCS; however, the RCIC System is included with the ECCS section because of their similar functions. r, The RCIC System is designed to operate either automatically or manually following reactor pressure vessel (RPV) isolation accompanied by a loss of coolant flow from the feedwater system to provide adequate core cooling and control of the RPV water level. Under these conditions, the High Pressure Coolant Injection (HPCI) and RCIC systems perform similar functions. The RCIC System design requirements ensure that the criteria of Reference I are satisfied.

The RCIC System (Ref. 2) consists of a steam driven turbine pump unit, piping, and valves to provide steam to the turbine, as well as piping and valves to transfer water from the suction source to the core via the feedwater system line, where the coolant is distributed within the RPV through the feedwater sparger. Suction piping is provided from the condensate storage tank (CST) and the suppression pool. Pump suction is normally aligned to the CST to minimize injection of suppression pool water into the RPV.

However, if the CST water supply is low, or the suppression pool level is high, a manual transfer to the suppression pool water source ensures a water supply for continuous operation of the RCIC System. The steam supply to the turbine is piped from a'ain steam line upstream of the associated inboard main steam line isolation valve.

The RCIC System is designed to provide core cooling for a wide range of reactor pressures (150 psig to 1174 psig).

Upon receipt of an initiation signal, the RCIC turbine accelerates until a 600 gpm flow rate (design flow) is achieved. As the RCIC turbine flow varies, the turbine control valve is automatically adjusted to maintain design flow. Exhaust steam from the RCIC turbine is discharged to the suppression pool. A full flow test line is provided to route water from and to the CST to allow testing of the RCIC (continued)

B 3.5-24 Amendment BFN-UNIT 2

RCIC System 8 3.5.3 SURVEILLANCE SR 3.5.3.2 (continued)

REQUIREMENTS steam flow path for the turbine and the flow controller position.

The 31 day Frequency of this SR,was derived from the Inservice Testing Program requirements for performing valve testing at least once every 92 days. The Frequency of 31 days is further justified because the valves are operated under procedural control and because improper valve position would affect only the RCIC System. This Frequency has been shown to be acceptable through operating experience.

SR 3.5.3.3 and SR 3.5.3.4 The RCIC pump flow rates ensure that the system can maintain reactor coolant inventory during pressurized conditions with the RPV isolated. The flow tests for the RCIC System are performed at two different pressure ranges such that system capability to provide rated flow is tested both at the higher and lower operating ranges of the system.

Additionally, adequate steam flow must be passing through the main turbine or turbine bypass valves to continue to control reactor pressure when the RCIC System diverts steam flow. Reactor steam pressure must be a 950 psig to perform SR 3.5.3.3 and a 150 psig to perform SR 3.5.3.4. Therefore, sufficient time is allowed after adequate pressure is achieved to perform these SRs. Reactor startup is allowed prior to performing the low pressure Surveillance because the reactor pressure is low and the time allowed to satisfactorily perform the Surveillance is short.

Alternately, the low pressure Surveillance test may be performed prior to startup using an auxiliary steam supply.

The reactor pressure is allowed to be increased to normal operating pressure since it is assumed that the low pressure Surveillance has been satisfactorily completed and there is no indication or reason to believe that RCIC is inoperable.

Therefore, these SRs are modified by Notes that state the Surveillances are not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after .the reactor steam pressure and flow are adequate to perform the test.

(continued)

BFN-UNIT 2 B 3.5-28 Amendment

Primary Containment 8 3.6.1.1

'PPLICABLE The safety design basis for the primary containment is that SAFETY ANALYSES it must withstand the pressures and temperatures of the exceeding the design leakage rate.

limiting OBA without The DBA that postulates the maximum release of radioactive material within primary containment is a LOCA. In the analysis of this 'accident, it is assumed that primary containment is OPERABLE such that release of fission products to the environment is controlled by the rate of primary containment leakage.

Analytical methods and assumptions involving the primary containment are presented in References 1 and 2. The safety analyses assume a nonmechanistic fission product release following a DBA, which forms the basis for determination of offsite doses. The fission product release is, in turn, based on an assumed leakage rate from the primary containment. OPERABILITY of the primary containment ensures that the leakage rate assumed in the safety analyses is not =

exceeded.

The maximum allowable leakage rate for the primary containment (L,) is 2.0% by weight of the containment air per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at the design basis LOCA maximum peak containment pressure (P.) of 50.3 psig (Ref. 1).

Primary containment satisfies Criterion 3 of the NRC Policy Statement (Ref. 6).

LCO Primary containment OPERABILITY is maintained by limiting leakage to ~ 1.0 L., except prior to the first startup after performing a required Primary Containment Leakage Rate Testing Program leakage test. At this time, applicable leakage limits must be"met. Compliance with this LCO will ensure a primary containment configuration, including equipment hatches, that is structurally sound and that will limit leakage to those leakage rates assumed in the safety analyses.

Individual leakage rates specified for the primary containment air lock are addressed in LCO 3.6. 1.2.

(continued)

BFN-UNIT 2 8 3.6-2 Amendment

Primary Containment Air Lock B 3.6.

1.2 BACKGROUND

and leak tightness are essential for maintaining primary (continued) containment leakage rate to within limits in the event of a DBA. Not maintaining air lock integrity or leak tightness may result in a leakage rate in excess of that assumed in the unit safety analysis.

APPLICABLE The DBA that postulates the maximum release of radioactive SAFETY ANALYSES material within primary containment is a LOCA. In the analysis of this accident, it is assumed that primary containment is OPERABLE, such that release of fission products to the environment is controlled by the rate of primary containment leakage. The primary containment is designed with a maximum allowable leakage rate (L,) of 2.05 by weight of the containment air per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at the calculated maximum peak containment pressure (P.) of 50.3 psig (Ref. 3). This allowable leakage rate forms the basis for the acceptance criteria imposed on the SRs associated with the air lock.

Primary containment air lock OPERABILITY is also required to minimize the amount of fission product gases that may escape primary containment through the air lock and contaminate and pressurize the secondary containment.

The primary containment air lock satisfies Criterion 3 of the NRC Policy Statement (Ref. 4).

LCO As part of primary containment, the air lock's safety function is related to control of containment leakage rates following a DBA. Thus, the air lock's structural integrity and leak tightness are essential to the successful mitigation of such an event.

The primary containment air lock is required to be OPERABLE.

For the air lock to be considered OPERABLE, the air lock interlock mechanism must be OPERABLE, the air lock must be in compliance with the Type B air lock leakage test, and both air lock doors must be OPERABLE. The interlock allows only one air lock door to be opened at a time. This provision ensures that a gross breach of primary containment does not exist when primary containment is required to be (continued)

BFN-UNIT 2 B 3.6-7 Amendment

0' RHRSM System and UHS B 3.7.1 e B B

3.7 3.7.1 PLANT SYSTENS Residual Heat Removal Service Water (RHRSM) System BASES BACKGROUND The RHRSM System is designed to provide cooling water for the Residual Heat Removal (RHR) System heat exchangers, required for a safe reactor shutdown following a Design Basis Accident (DBA) or transient. The RHRSM System is operated whenever the RHR heat exchangers are required to operate in the shutdown cooling mode or in the suppression pool cooling or spray mode of the RHR System.

The RHRSM System is common to the three BFN units and consists of the UHS and four independent and redundant loops, each of which feeds one RHR heat exchanger in each unit. Each loop is made up of a header, two 4500 gpm pumps, a suction source, valves, piping, and associated instrumentation. One loop with one pump operating is capable of providing 5N'f the required cooling capacity to maintain safe shutdown conditions for one unit following a design basis accident. However, one pump is capable of providing sufficient cooling capacity to maintain a safe shutdown condition for each of the non-accident units. As such, a subsystem consists of a loop with one or two OPERABLE pumps dependent upon the number of fueled units.

The RHRSW System is designed with sufficient redundancy so that no single active component failure can prevent it from achieving its design function. The RHRSW System is described in the FSAR, Section 10.9 (Ref. 1).

Cooling water is pumped by the RHRSM pumps from the Wheeler Reservoir through the tube side of the RHR heat exchangers, and discharged back to the Mheeler Reservoir.

The system is initiated manually from each of the three units control rooms. If operating during a loss of coolant accident (LOCA), the system is automatically tripped on degraded bus voltage to allow the diesel generators to automatically power only that equipment necessary to reflood the core. The system can be manually started any time the degraded bus voltage signal is manually overridden or clears, and is assumed to be manually started within 10 minutes after the LOCA.

(continued)

BFN-UNIT 2 B 3.7-1 Amendment

0 RHRSM System and UHS B 3.7.1 APPLICABLE The RHRSM System removes heat from the suppression pool to SAFETY ANALYSES limit the suppression pool temperature and primary containment pressure following a LOCA. This ensures that the primary containment can perform its function of limiting the release of radioactive materials to the environment following a LOCA. The ability of the RHRSM System to support long term cooling of the reactor or primary containment is discussed in the FSAR, Chapters 5 and 14 (Refs. 2 and 3, respectively). These analyses explicitly assume that the RHRSM System will provide adequate cooling support to the equipment required for safe shutdown. These analyses include the evaluation of the long term primary containment response after a design basis LOCA.

The safety analyses for long term cooling were performed for various combinations of RHR System failures and considers the number of units fueled. Mith one unit fueled, the worst case single failure that would affect the performance of the RHRSM System is any failure that would disable two subsystems or pumps of the RHRSM System (e.g, the failure of an RHR Suppression Pool Cooling/Spray return line valve which effectively disables two RHRSM subsystems or pumps).

Mith two and three units fueled, a worst case single failure could also include the loss of two RHRSM pumps caused by losing a 4 kV shutdown board since there are certain alignment configurations that allow two RHRSM pumps to be powered from the same 4 kV shutdown board. As discussed in the FSAR, Section 14.6.3.3.2 (Ref. 4) for these analyses, manual initiation of the OPERABLE RHRSM subsystems and the associated RHR System is assumed to occur 10 minutes after a DBA. The RHRSM flow assumed in the analyses is 4000 gpm per pump with two pumps operating in one loop. In this case, the maximum suppression chamber water temperature and pressure are 177'F (as reported in Reference 3) and 50.3 psig, respectively, well below the design temperature of 281'F and maximum allowable pressure of 62 psig. This is also below the 200'F limit imposed by Design Criteria BFN-50-7064A (Ref. 5) for all plant transients involving SRV operations.

The RHRSM System, together with the UHS, satisfies Criterion 3 of the NRC Policy Statement (Ref 6).

(continued)

BFN-UNIT 2 B 3.7-2 Amendment

RHRSM System and UHS B 3.7.1 BASES (continued)

LCO Four RHRSW subsystems are required to be OPERABLE to provide the required redundancy to ensure that the system functions to remove post accident heat loads, assuming the worst case single active failure occurs coincident with the loss of offsite power.

An OPERABLE RHRSM subsystem consists of:

a. The required number of OPERABLE RHRSM pumps dependent upon the number of units fueled; and

b. An OPERABLE flow path capable of taking suction from the intake structure and transferring the water to the required RHR heat exchangers at the assumed flow rate.

The LCO is modified by two Notes. Note 1 specifies that when 1 or 2 units are fueled, there must be at least one OPERABLE pump per RHRSM subsystem. Note 2 specifies that when 3 units are fueled, two of the RHRSW subsystems must have two OPERABLE RHRSM pumps.

e The OPERABILITY of the UHS for RHRSM is based on having a maximum water temperature within the limits specified in Figure 3.7.1-1.

APPLICABILITY In NODES 1, 2, and 3, the RHRSM System and UHS are required to be OPERABLE to support the OPERABILITY of the RHR System for. primary containment cooling (LCO 3.6.2.3, "Residual Heat Removal (RHR) Suppression Pool Cooling," and LCO 3.6.2.4, "Residual Heat Removal (RHR) Suppression Pool Spray" ) and decay heat removal (LCO 3.4.7, "Residual Heat Removal (RHR)

Shutdown Cooling System-Hot Shutdown" ). The Applicability is therefore consistent with the requirements of these systems.

In MODES 4 and 5, the OPERABILITY requirements of the RHRSM System and UHS are determined by the systems they support.

ACTIONS The Actions are modified by a Note indicating that the applicable Conditions of LCO 3.4.7 be entered and Required Actions taken if the inoperable RHRSM subsystem results in inoperable RHR shutdown cooling. This is an exception to (continued)

BFN-UNIT 2 B 3.7-3 Amendment

RHRSM System and UHS B 3.7.1 ACTIONS LCO 3.0.6 and ensures the proper actions are taken for these (continued) components.

With one RHRSM subsystem or required pump inoperable, the inoperable RHRSM subsystem or required pump must be restored to OPERABLE status within 30 days. With the unit in this condition," the remaining OPERABLE RHRSW subsystems are adequate to perform the RHRSW heat removal function.

However, the overall reliability is reduced because a single failure could result in reduced primary containment cooling capability. The 30 day Completion Time is based on the availability of equipment in excess of normal redundancy requirements and the low probability of an event occurring requiring RHRSM during this period.

B. B. a d B.3 Required Action B.l requires verification that at least three OPERABLE RHRS'W pumps are associated with the two OPERABLE RHRSM subsystems. The Required Action is modified by a Note indicating that the required action is applicable only when two units are fueled. Required Action B.2 requires verification that at least four OPERABLE RHRSM pumps are associated with the two OPERABLE RHRSM subsystems.

The Required Action is modified by a Note indicating that the required action is applicable only when three units are fueled.

Required Action B.3 requires that with two RHRSW subsystems inoperabl'e, one inoperable RHRSW subsystem be restored to OPERABLE status within 7 days. Mith the unit(s) in this condition, the remaining OPERABLE RHRSW subsystems are adequate to perform the RHRSM heat removal function.

However, the overall reliability is reduced because a single failure in the OPERABLE RHRSM subsystems could result in loss of RHRSW function. The 7 day Completion Time is based on the redundant RHRSW capabilities afforded by the OPERABLE subsystems and the low probability of an event occurring requiring RHRSM during this period.

(continued)

BFN-UNIT 2 8 3.7-4 Amendment

RHRSW System and UHS B 3.7.1 BASES ACTIONS C.1 (continued)

With three or more RHRSW subsystems inoperable, the RHRSW System is not capable of performing its intended function.

The requisite number of subsystems and pumps must be restored to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . The 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Completion Time is based on the Completion Times provided for the RHR suppression pool cooling and spray functions.

D.1 and D.2 If the RHRSW subsystems cannot be restored to OPERABLE status within the associated Completion Times or the UHS is determined inoperable, the unit must be placed in a NODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least NODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in NODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

SURVEILLANCE S 3 7.1.

REQUIREMENTS Verifying the correct alignment for each manual and power operated valve in each RHRSW subsystem flow path provides assurance that the proper flow paths will exist for RHRS'W oper ation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these valves are verified to be in the correct position prior to locking, sealing, or securing. A valve is also allowed to be in the nonaccident position, and yet considered in the correct position, provided it can be realigned to its accident position. This is acceptable because the RHRSW System is a manually initiated system.

This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves.

(continued)

BFN-UNIT .2 B 3.7-5 Amendment

RHRSW System and UHS B 3.7.1 SURVEILLANCE RE(UIR EVENTS

~ll ... ( tt dl The 31 day Frequency is based on engineering judgment, is consistent with the procedural controls governing valve operation, and ensures correct valve positions.

JII9. 7 .. R Verification of the UHS temperature ensures that the heat removal capability of the RHRSW System is within the assumptions of the DBA analysis. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months frequencies are based on operating experience relating to trending of the parameter variations during the applicable NODES.

REFERENCES 1. FSAR, Section 10.9.

2. FSAR, Chapter 5.

3. FSAR, Chapter 14.

4. FSAR, Section 14.6.3.3.2.

5. Design Criteria BFN-50-7064A, Primary Containment Systems - Units 2 and 3

6. NRC No.93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.

BFN-UNIT 2 B 3.7-6 Amendment

EECM System and UHS B 3.7.2 B 3.7 PLANT SYSTEMS 8 3.7.2 Emergency Equipment Cooling Mater (EECM) System and Ultimate Heat Sink (UHS)

BASES BACKGROUND The EECW System is designed to provide cooling water for the removal of heat from equipment, such as the diesel generators (DGs), residual heat removal (RHR) pump coolers, and room coolers for other Emergency Core Cooling System equipment, required for a safe reactor shutdown following a Design Basis Accident (DBA) or transient. The EECW System also provides cooling to unit components, as required, during normal operation. Upon receipt of a loss of offsite power or loss of coolant accident (LOCA) signal, the essential loads are provided cooling water by automatically starting RHRSW pumps aligned to EECW headers.

The EECW System, which is common to the three BFN units, consists of the UHS and two independent and redundant loops with each loop consisting of a header, two 4500 gpm pumps, a suction source, valves, piping and associated instrumentation. Two EECM pumps (one per loop or both on one loop) are capable of providing the required cooling capacity to support the required systems. The two loops are separated from each other so failure of one loop will not affect the OPERABILITY of the other. The EECW System is described in the FSAR, Section 10. 10 (Ref. 3)

Cooling water is pumped from the Wheeler Reservoir by the EECW pumps to the essential components through the two main headers. After removing heat from the components, the water is discharged back to the Wheeler Reservoir.

APPLICABLE Sufficient water inventory is available for all EECW System SAFETY ANALYSES post LOCA cooling requirements for a 30 day period with no additional makeup water source available. The ability of the EECW System to support long term cooling of the reactor containment is assumed in evaluations of the equipment required for safe reactor shutdown presented in the FSAR, Chapters 5 and 14 (Refs. 2 and 3, respectively). These analyses include the evaluation of the long term primary containment response after a design basis LOCA.

(continued)

BFN-UNIT 2 8 3.7-7 Amendment

EECW System and UHS B 3.7.2 APPLICABLE The ability of the EECW System to provide adequate cooling SAFETY ANALYSES to the identified safety equipment is an implicit assumption (continued) for the safety analyses evaluated in References 1 and 2.

The ability to provide onsite emergency AC power is dependent on the ability of the EECW System to cool the DGs.

The long term cooling capability of the RHR and core spray pumps is also dependent on the cooling provided by the EECW System.

The EECW System, together with the UHS, satisfy Criterion 3 of the NRC Policy Statement (Ref. 4).

LCO The EECW loops are independent of each other to the degree that each has separate controls, power supplies, and the operation of one does not depend on the other. In the event of a DBA, two EECW pumps are required to provide the minimum heat removal capability assumed in the safety analysis for the system to which it supplies cooling water. To ensure this requirement is met, three EECW pumps must be OPERABLE.

At least two pumps will operate if the worst single active failure occurs coincident with the loss of offsite power."

The EECW System is considered OPERABLE when it has an OPERABLE UHS, three OPERABLE pumps, and two OPERABLE flow paths capable of taking suction from the intake structure and transferring the water to the appropriate equipment.

The OPERABILITY of the UHS for EECW is based on having a maximum water temperature of 95'F. Additional requirements for UHS temperatures are provided in SR 3.7. 1.2.

The isolation of the EECW System to components or systems may render those components or systems inoperable, but does not affect the OPERABILITY of the EECW System.

APPLICABILITY In MODES 1, 2, and 3, the EECW System and UHS are required to be OPERABLE to support OPERABILITY of the equipment serviced by the EECW System. Therefore, the EECW System and UHS are required to be OPERABLE in these MODES.

In MODES 4 and 5, the OPERABILITY requirements of the EECW System and UHS are determined by the systems they support.

(continued)

BFN-UNIT 2 B 3.7-8 Amendment

EECW System and UHS B 3.7.2 BASES (continued)

ACTIONS A.1 With one required EECW pump inoperable, the required EECW pump must be restored to OPERABLE status within 7 days.

With the system in this condition, the remaining OPERABLE EECW pumps are adequate to perform the heat removal function. However, the overall reliability is reduced because a single failure in the EECW System could result in loss of EECW function.

The 7 day Completion Time is based on the redundant EECW System capabilities afforded by the remaining OPERABLE pumps, the low probability of an accident occurring during this time period and is consistent with the allowed Completion Time for restoring an inoperable DG.

B.l and B.2 If the required EECW pump cannot be restored to OPERABLE status within the associated Completion Time, or two or more EECW pumps are inoperable or the UHS is determined inoperable, the unit must be placed in a HODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least HODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in HODE 4 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

SURVEILLANCE SR 3.7.2.1 RE(UIREHENTS Verification of the UHS temper ature ensures that the heat removal capability of the EECW System is within the assumptions of the DBA analysis. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency is based on operating experience related to trending of the parameter variations during the applicable HODES.

SR 3.7.2.2 Verifying the correct alignment for each manual and power operated valve in the EECW System flow paths provide assurance that the proper flow paths will exist for EECW (continued)

BFN-UNIT 2 B 3.7-9 Amendment

EECW System and UHS B 3.7.2 BASES SURVEILLANCE SR 3.7.2.2 (continued)

RE(UIREHENTS operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these valves were verified to be in the correct position prior to locking, sealing, or securing. A valve is also allowed to be in the nonaccident position, and yet considered in the correct 'position, provided it can be automatically realigned to its accident position within the required time. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves.

This SR is modified by a Note indicating that isolation of the EECW System to components or systems may render those components or 'systems inoperable, but does not affect the OPERABILITY of the EECW System. As such, when required EECW pumps, valves, and piping are OPERABLE, but a branch connection off the main header is isolated, the EECW System is still OPERABLE.

The 31 day Frequency is based on engineering judgment, is consistent with the procedural controls governing valve operation, and ensures correct valve positions.

SR 3.7.2.3 This SR verifies that the EECW System pumps will automatically start to provide cooling water to the required safety related equipment during an accident event. This is demonstrated by the use of an actual or simulated initiation signal.

Operating experience has shown that these components will usually pass the SR when performed at the 18 month Frequency. Therefore, this Frequency is concluded to be acceptable from a reliability standpoint.

(continued)

BFN-UNIT 2 B 3.7-10 Amendment

EECW System and UHS B 3.7.2 BASES (continued)

REFERENCES 1. FSAR, Chapter 5.

2. FSAR, Chapter 14.

3. FSAR, Section 10. 10.

4. NRC No.93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.

B 3.7-11 Amendment

~ '

ATMS-RPT Instrumentation 3.3.4.2 SURVEILLANCE REQUIREMENTS continued SURVEILLANCE FREQUENCY SR 3.3.4.2.2 Perform CHANNEL FUNCTIONAL TEST. 92 days SR 3.3.4.2.3 Perform CHANNEL CALIBRATION. The 18 months

. Allowable Values shall be:

a. Reactor Vessel Mater Level Low Low, Level 2: a 471.52 inches above vessel zero; and

b. Reactor Steam Dome Pressure High:

a 1175.0 psig.

SR 3.3.4.2.4 Perform LOGIC SYSTEH FUNCTIONAL TEST 18 months including breaker actuation.

BFN-UNIT 2 3.3-34 Amendment

RPS Instrumentation 3.3.1.1 Table 3.3.1.1-1 (page 2 of 3)

Reactor Protection System Instruaentation APPLICABLE COND IT IOIS HODES OR REQUIRED REFERENCED OTHER CHANNELS FROH SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLNIABLE FUNCTION COND IT I ONS SYSTEH ACTION D.1 REQJIREHENTS VALUE

2. Average Pouer Range Honi tors (cont inued)

d. Dosnscale SR 3.3.1.1.7 h 3X RTP SR 3.3.1.1.8 SR 3.3.1.1.14

e. Inop 1,2 SR 3.3.1 1.7

~

SR 3.3.1.1.8 SR 3.3.1.1.14

3. Reactor Vessel Stes¹ 1,2 SR 3.3.1.1 1~ 5 1090 psfg Dome Pressure -High SR 3.3.1.1.8 SR 3.3.1 1.10

~

SR 3.3.1.1.'14

4. Reactor Vessel Mater 1,2 SR 3.3.1.1.1 > 538 inches Level -Lcm, Level 3 SR 3.3.1.1.8 above vessel SR 3.3.1.1.13 Zero SR 3.3 1.1.14

~

5. Hain Stoa¹ Isolation SR 3.3.1.1.8 S 10K closed Valve -Closure SR 3.3.1.1.13 SR 3.3.1.1.14

6. Drys l l Pressure - High 1,2 SR SR 3.3.1 1.8

~

3.3.1.1.13 S 2.5 psig SR 3.3.1.1.14

7. Berm Discharge Voiuae Mater l.evel -High

a. Res i stance Tea@era ture 1,2 SR 3.3.1.1.8 s 50 gallons Detector SR 3.3.1.1.13 SR 3.3.F 1.14 5(a) SR 3.3.1.1.8 5 50 gallons SR 3.3.1.1.13 SR 3.3.1.1.14

b. Float Sui tch 1,2 SR 3.3.1.1.8 5 50 gallons SR 3.3.1.1.13 SR 3.3.1.1.14 5(a) SR 3.3.1.1.8 S 50 gallons SR 3.3.1 1.13

~

SR 3.3.1.1.14 (cont inued)

(a) Mith any control rod Nithdraun from a core cell containing one or more fuel assemblies.

BFN-UNIT 2 3.3-7 Amendment

RPS Instrumentation 3.3.l.l Table 3.3.1.1-1 (page 2 of 3)

Reactor Protection System Instruaentatfon APPLICABLE COND I T IONS NODES OR REQUIRED REFERENCED OTHER CHANNELS FROI SPECIFIED PER TRIP REOUI RED SURVEILLANCE ALLOMASLE FUNCTION CONDITIONS SYSTEN ACTION D.1 REQUIRENENTS VALUE

2. Average Pouer Range Noni tors (cont inued)

d. Downscale. SR 3.3.1.1.7 2 3X RTP SR 3.3.1.1.8 SR 3.3.1.1.14

e. Inop 1,2 SR 3.3.1.1.7 SR 3.3.1.1.8 SR 3.3.1.1 ~ 14

3. Reactor Vessel Steam 1,2 SR 3.3.1.1.1 5 1090 psig Dome Pressure -High SR 3.3.1.'1.8 SR 3.3.1.1.10 SR 3.3 '.1.14

4. Reactor Vessel Mater 1,2 SR 3.3.1.1.1 ? 538 inches Level -Lou, Level 3 SR 3.3.1.1.8 above vessel SR 3.3.1.1.13 zero SR 3.3.1.1.14

5. Hain Stems Isolation SR 3.3.1.1.8 5 10K closed Valve -Closure SR 3.3 1.1.13

~

SR 3.3.1.1.14

6. Dryuell Pressure -High 1,2 SR 3.3.1.1.8 5 2.5 psig SR 3.3.1.1.13 SR 3.3.1.1.14

7. Scram Discharge Voiune Mater Level -High

a. Resistance Temperature 1,2 SR 3.3.1 1.8

~ 5 50 gallons Detector SR 3.3.1.1.13 SR 3.3.1.1.14 5(a) SR 3.3.1.1.8 < 50 gallons SR 3.3.1.1.13 SR 3.3.1.1.14

b. Float Sgitch 1,2 SR 3.3.1.1.8 5 50 gallons SR 3.3.1.1 13 SR 3.3.1.1.14 5(a) SR 3.3.1.1.8 < 50 gallons SR 3.3.1.1.13 SR 3.3.1.1.14 (continued)

(a) Mith any control rod withdram from a core cell containing one or more fuel assemblies.

BFN-UNIT 3 3.3-7 Amendment

0 ATMS-RPT Instrumentation 3.3.4.2 SURVEILLANCE RE UIREHENTS continued SURVEILLANCE FREQUENCY SR 3.3.4.2.2 Perform CHANNEL FUNCTIONAL TEST. 92 days SR 3.3.4.2.3 Perform CHANNEL CAL'IBRATION. The 18 months Allowable Val,ues shall be:

a. Reactor Vessel Mater Level Low Low, Level 2: a 471.52 inches above vessel zero; and

b. Reactor Steam Dome Pressure High:

a 1175.0 psig.

SR 3.3.4.2.4 Perform LOGIC SYSTEM FUNCTIONAL TEST 18 months including breaker actuation.

BFN-UNIT 3 .3.3-34 Amendment

JULY 08, 1997 (3) Pursuant to the Act and 10 CFR Parts 30, 40 and 70 to receive/

possess and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as requiredg (4) Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive/

possess and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form for sample analysis or instrument calibration or associated with radioactive apparatus or componentsy (5) Pursuant to the Act and 10 CFR Parts 30 and 70, to possess but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C. This license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter Is Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70'nd is sub)ect to all applicable provisions of the Act and to.,

the rules, regulations, and orders of the Commission now or hereafter in effect; and is sub)ect to the additional conditions specified or incorporated below:

(1) Maximu ower Leve The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 3458 megawatts thermal.

(2) Tech ica S ecificatio s The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 208 are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

(3) Deleted.

BFN Unit 3

0 Definitions 1.1

1. 1 Definitions (continued)

PHYSICS TESTS PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation.

These tests are:

a. Described in Section 13.10, Refueling Test Program; of the FSAR;

b. Authorized under the provisions of 10 CFR 50.59; or

c. Otherwise approved by the Nuclear Regulatory Commission.

RATED THERMAL POWER RTP shall be a total reactor core heat transfer (RTP) rate to the reactor coolant of 3458 HWt.

SHUTDOWN MARGIN (SDH) SDH shall be the amount of reactivity by which the reactor is subcritical or would be subcritical assuming that:

a. The reactor is xenon free;

b. The moderator temperature is 68'F; and

c. All control rods are fully inserted except for the single control rod of highest reactivity worth, which is assumed to be fully withdrawn.

With control rods not capable of being fully inserted, the reactivity worth of these control rods must be accounted for in the determination of SDM.

STAGGERED TEST BASIS A STAGGERED TEST BASIS shall consist of the testing of one of the systems, subsystems, channels, or other designated components during the interval specified by the Surveillance Frequency, so that all systems, subsystems, .

channels, or other designated components are tested during n Surveillance Frequency intervals, where n is the total number of systems, subsystems, channels, or other designated components in the associated function.

(cont>nued)

BFN-UNIT 3 1.1-5 Amendment

SLC System 3.1.7 SURVEILLANCE REQUIREMENTS continued SURVEILLANCE FREQUENCY SR 3. 1.7.5 Verify the SLC conditions satisfy the following equation: 31 days AND (13 wt. X) (86 gpm) (19.8 atomX)

Once within 24 where, hours after water or boron C = sodium pentaborate solution is added to the concentration (weight percent) solution.

g = pump flow rate (gpm)

E - Boron-10 enrichment (atom percent Boron-10) 0'R 3.1.7.6

)Verify a

39 1325 each gpm at psig.

pump a

develops a flow rate discharge pressure 18 months SR 3.1.7.7 Verify flow through one SLC subsystem from 18 months on a pump into reactor pressure vessel. STAGGERED TEST BASIS SR 3. 1.7.8 Verify all piping between storage tank and 18 months pump suction is unblocked.

(continued)

BFN-UNIT 3 3.1-23 Amendment

RPS Instrumentation 3.3.1.1 Table 3.3.1.1-1 (page 2 of 3)

Reactor Protection Systos Instruaentat ion APPLICABLE COND IT IONS NODES OR REQUIRED REFERENCED OTHER CHANNELS FROH SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOMABLE FUNCT ION COND I T I ONS SYSTEN . ACTION D.1 REQUIREHENTS VALUE

2. Average Power Range Noni tors (cont inued)

d. Downscale SR 3.3.1.1.7 2 3X RTP SR 3.3.1.1.8 SR 3.3.1.1.14

e. Inop 1,2 SR 3.3.1.1.7 NA SR 3.3.1.1.8 SR 3.3.1.1.14

3. Reactor Vessel Steam 1,2 SR 3.3 ~ 1 ~ 1.1 S 1090 psig Dome Pressure -High SR 3.3.1.1.8 SR 3.3.1.1.10 SR 3.3.1.1.14

4. Reactor Vessel Mater 1,2 SR 3.3.1.1.1 2 538 inches Level -Low, Level 3 SR 3.3.1.1.8 above vessel SR 3.3.1.1.13 zero SR 3.3.1.1.14

5. Hain Steaa Isolation SR 3.3.1.1.8 < 10X closed Valve Closure SR 3.3.1.1.13 3.3.1.1.14 SR

6. Drywall Pressure -High 1,2 SR 3.3.1.1.8 S 2.5 psig SR 3.3.1.1.13 SR 3.3.1.1 '4

7. Scram Discharge Volune Mater Level -High

a. Resistance Tespelatule 1,2 SR 3.3.1.1.8 5 50 gallons Detector SR 3.3 ~ 1.1.13 SR 3.3.1.1.14 5(a) SR 3.3.1.1.8 S 50 gallons SR 3.3.1.1.13 SR 3.3.1.1.14

b. Float Switch 1,2 SR 3.3.1.1.8 5 50 gallons SR 3.3.1.1 13 ~

SR 3.3.1.1.14 5(a) SR 3.3.1 1.8

~

< 50 gallons SR 3.3.1.1.13 SR 3.3.1.1.14 (cont inued)

(a) Mith any control rod withdrawn from a core cell containing one or more fuel assemblies.

BFN-UNIT 3 303 7 Amendment

ATMS-RPT Instrumentation 3.3.4.2 SURVEILLANCE REQUIREMENTS continued SURVEILLANCE FREQUENCY SR 3.3.4.2.2 Perform CHANNEL FUNCTIONAL TEST. 92 days SR 3.3.4.2.3 Perform CHANNEL CALIBRATION. The 18 months

~ - Allowable Values shall be:

a. Reactor Vessel Mater Level Low Low, Level 2: a 471.52 inches above vessel zero; and

b. Reactor Steam Oome Pressure High:

a 1175.0 psig.

SR 3.3.4.2. 4 Perform LOGIC SYSTEM FUNCTIONAL TEST 18 months including breaker actuation.

BFN-UNIT 3 3.3-34 Amendment

c: IZ

a. n Pt BFN Power/Flow Stability Regions Region Region I II 10 2.9% R od Lin 100.0% Rod Line O

0 mzfll'

~ ~O o

70.00 95.2'6

/i Rod Line 0 K il 2% Rod Line I (Q 6o.oo U C O

~ 0o O.

Note: Operation

~ m~ so.oo Not Permitted In I g This Region

~

(

Ill m I-m+0) 'g 40.00

~ C z00)

0) CL 30.00 Natural 20.00 Circulation Line 10.00 0.00 0 5 10 15 20 25 30 35 40 45 50 55 60 65 70 75 80 85 90 95 100 105 Core Flow (% of rated)

S/RVs 3.4.3 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.3 Safety/Relief Valves (S/RVs)

LCO 3.4.3 The safety function of 12 S/RVs shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIHE A. One or more required A.l Be in HODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months S/RVs inoperable.

AND in t

A.2 Be MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months SURVEILLANCE REQUIREMENTS SURVEILLANCE , FREQUENCY SR 3.4.3.1 Verify the safety function lift settings of In accordance the required S/RVs are within setpoint as follows:

i 3% of the with the Inservice Testing Program Number of Setpoint

~S'Vs 4 1135 4 1145 5 1155 Following testing, lift settings shall be within i I'X.

(continued)

BFN-UNIT 3 3.4-7 Amendment

0 ECCS-Operating 3.5.1 SURVEILLANCE RE(UIREMENTS continued SURVEILLANCE FREQUENCY SR 3.5.1.5 NOTE Only required to be performed prior to entering MODE 2 from MODE 3 or 4, when in MODE 4 ) 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

Verify each recirculation pump discharge 31 days valve cycles through one complete cycle of full travel.

SR 3.5.1.6 Verify the following ECCS pumps develop the In accordance specified flow rate against a system head with the corresponding to the specified reactor Inservice pressure. Testing SYSTEM, HEAD Program CORRESPONDING TO A VESSEL TO NO. TORUS OF DIFFERENTIAL SYSTEM FLOM RATE PUMPS PRESSURE OF Core Spray z 6250 gpm 2 a 105 psid LPCI ~ 19,200 gpm 2 a 20 psid LPCI a 10,450 gpm 1 > 20 psid SR 3.5.1.7 -NOTE-Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure and flow are adequate to perform the test.

Verify, with reactor pressure z 1040 and 92 days a 950 psig, the HPCI pump can develop a flow rate a 5000 gpm against a system head corresponding to reactor pressure.

(continued}

BFN-'UNIT 3 3 '-5 Amendment

RCIC System 3.5.3 SURVEILLANCE RE(UIREHENTS SURVEILLANCE FREQUENCY SR 3.5.3.1 Verify the.RCIC System piping is filled 31 days with water from the pump discharge valve to the injection valve.

SR 3.5.3. 2 Verify each RCIC System manual, power 31 days operated, and automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position.

SR 3.5.3.3 -NOTE-Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure and flow are adequate to perform the test.

~ 'erify, and with reactor pressure z 1040 psig

~ 950 psig, the RCIC pump can develop flow rate a 600 gpm against a system head a

92 days corresponding to reactor pressure.

SR 3.5.3.4 NOTE-Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure and flow are adequate to perform the test.

Verify, with reactor pressure x 165 psig, 18 months the RCIC pump can develop a flow rate a 600 gpm against a system head corresponding to reactor pressure.

(continued)

BFN-UNIT 3 3.5-13 Amendment

RHRSM System and UHS 3.7.1 3.7 PLANT SYSTEMS 3.7. 1 Residual Heat Removal Service Mater {RHRSM) System and Ultimate Heat Sink (UHS)

LCO 3.7.1 NOTES-

1. Mith 1 or 2 units fueled, each subsystem must have at least one OPERABLE RHRSW pump.

2. Mith 3 units fueled, two RHRSW subsystems must have two OPERABLE RHRSM pumps.

Four RHRSM subsystems and UHS shall be OPERABLE.

APPLICABILITY: NODES 1, 2, and 3.

t ACTIONS N OTE-------------------------------------

Enter applicable Conditions and Required Actions of LCO 3.4.7, "Residual Heat Removal (RHR) Shutdown Cooling System-Hot Shutdown," for RHR shutdown cooling made inoperable by RHRSM System.

CONDITION REQUIRED ACTION COMPLETION TINE A. One RHRSW subsystem or A.l Restore RHRSM 30 days required pump subsystem or required inoperable. pump to OPERABLE status.

(continued)

BFN-UNIT 3 3.7-1 Amendment

RHRSM System and UHS p 3.7.1 CONDITION REQUIRED ACTION COMPLETION TINE B. Two RHRSW subsystems B.l --------NOTE--------

inoperable. Only applicable when two units are fueled.

Verify at least three 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months OPERABLE RHRSW pumps are associated with two OPERABLE RHRSM subsystems.

B.2 --------NOTE--------

Only applicable when three units are fueled.

Verify at least four 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months OPERABLE RHRSM pumps 0 are associated with two OPERABLE RHRSM subsystems.

B.3 Restore one 7 days inoperable RHRSM subsystem to OPERABLE status.

C. Three or more RHRSM C.1 Restore the required 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months subsystems inoperable. RHRSM subsystems to OPERABLE status.

Required Action B.l or B.2 and associated Completion Time not met.

(continued)

BFN-UNIT 3 3.7-2 Amendment

RHRSW System and UHS 3.7.1 CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action A.l, D. I Be in MODE 3. 12 hours 8.3, or C.l and associated Completion AND Time not met.

D.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months OR UHS inoperable.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.1.1 Verify each RHRSW manual and power operated 31 days valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position or can be aligned to the correct position.

SR 3.7.1.2 Verify the average water temperature of UHS 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months UHS is within the limits specified in Figure temperature (

3.7.1-1. 91'F 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months UHS temperature >

91'F BFN-UNIT 3 307 3 Amendment

RHRSW System and UHS 3.7.1 Figure 8.7.1-1 Reactor Thermal Power Versus Ultimate Heat Sink Temperature Limit Unacceptable Acceptable L tthraate Heat Slntt Terat>>rature (degrees F)

Amendment

t EECM System and UHS 3.7.2 3.7 PLANT SYSTEMS 3.7.2 Emergency Equipment Cooling Mater (EECM) System and Ultimate Heat Sink (UHS)

LCO 3.7.2 The EECM System with three pumps and UHS shall be OPERABLE.

APPLICABILITY: NODES I, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TINE A. One required EECW pump A. I Restore the required 7 days inoperable. EECW pump to OPERABLE status.

B. Required Action and B.l Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition" A AND not met.

B.2 Be in NODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months OR Two or more required EECM pumps inoperable.

OR UHS inoperable.

BFN-UNIT 3 3.7-4 Amendment

EECW System and UHS 3.7.2 SURVEILLANCE REQUIREHENTS SURVE ILLANCE FREQUENCY

'SR 3.7.2.1 NOTE Refer to SR 3.7. 1.2 for additional UHS requirements.

Verify the average water temperature of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months UHS is x 95'F.

SR 3.7.2.2 'NOTE Isolation of flow to individual components does not render EECW System inoperable.

Verify each EECW subsystem manual and power 31 days operated valve in the flow paths servicing safety related systems or components, that is not locked, sealed, or otherwise secured in position, is in the correct position.

SR 3.7.2.3 Verify each required EECW pump actuates on 18 months an actual or simulated initiation signal.

BFN-UNIT 3 3.7-5 Amendment

~ ~

SLC System B 3.1.7 BASES SURVEILLANCE SR 3.1.7.4 REQUIREMENTS (continued) This Surveillance requires the amount of Boron-10 in the SLC solution tank to be determined every 31 days. The enriched sodium pentaborate solution is made by combining stoichiometric quantities of borax and boric acid in demineralized water. Since the chemicals used have known Boron-10 quantities, the Boron-10 quantity in the sodium pentaborate solution formed can be calculated. This parameter is used as input to determine the volume requirements for SR 3. 1.7. 1. The 31 day Frequency of this Surveillance is appropriate because of the relatively slow variation of boron concentration between surveillances.

SR 3.1.7.6 Oemonstrating that each SLC System pump develops a flow rate z 39 gpm at a discharge pressure a 1325 psig ensures that pump performance has not 'degraded during the fuel cycle.

This minimum pump flow rate requirement ensures that, when combined with the sodium pentaborate solution concentration and enrichment requirements, the rate of negative reactivity insertion from the SLC System will adequately compensate for the positive reactivity effects encountered during power

'reduction, cooldown of the moderator, and xenon decay. This test confirms one point on the pump design curve and is indicative of overall performance. The 18 month Frequency is acceptable since inservice testing of the pumps, performed every 92 days, will detect any adverse trends in pump performance. I SR 3.1.7.7 and SR 3.1.7.8 These Surveillances ensure that there is a functioning flow path from the boron solution storage tank to the RPV, including the firing of an explosive valve. The replacement charge for the explosive valve shall be from the same manufactured batch as the one fired or from another batch that has been certified by having one of that batch successfully fired. The pump and explosive valve tested should be alternated such that both complete flow paths are tested every 36 months at alternating 18 month intervals.

The Surveillance may be performed in separate steps to (continued)

BFN-UNIT 3 B 3.1-44 Amendment

0 Recirculation Loops Operating B 3.4.1 APPLICABLE Safety analyses performed for FSAR Chapter 14 implicitly SAFETY ANALYSES assume core conditions are stable. However, at the high (continued) power/low flow corner of the power/flow map, an increased probability for limit cycle oscillations exists (Ref. 3) depending on combinations of operating conditions (e.g.,

power shape, bundle power, and bundle flow). Generic

'valuations indicate that when regional power oscillations become detectable on the APRMs, the safety margin may be insufficient under some operating conditions to ensure actions taken to respond to the APRMs signals would prevent violation of the MCPR Safety Limit (Ref. 4). NRC Generic Letter 86-02 (Ref. 5) addressed stability calculation methodology and stated that due to uncertainties, 10 CFR 50, Appendix A, General Design Criteria (GDC) 10 and 12 could not be met using analytic procedures on a BWR 4 design.

However, Reference 5 concluded that operating limitations which provide for the detection (by monitoring neutron flux noise levels) and suppression of flux oscillations in operating regions of potential instability consistent with ..

the recommendations of Reference 3 are acceptable to demonstrate compliance with GDC 10 and 12. The NRC concluded that regions of potential instability could occur at calculated decay ratios of 0.8 or greater by the General Electric methodology.

Stability tests at operating BWRs were reviewed to determine a generic region of the power/flow map in which surveillance of neutron flux noise levels should be performed. A conservative decay ratio was chosen as the basis for determining the generic region for surveillance to account for the plant to plant variability of decay ratio with core and fuel designs. This decay ratio also helps ensure sufficient margin to an instability occurrence is maintained. The generic region has been determined to be bounded by the 76.2X rod line and the 45K core flow line.

.BFN conservatively implements this generic region with the "Operation Not Permitted" Region and Regions I and II of Figure 3.4. 1-1. This conforms to Reference 3 recommendations. .Operation is permitted in Region II provided neutron flux noise levels are verified to be within limits. The reactor mode switch must be placed in the shutdown position (an immediate scram is required) if Region I is entered.

Recirculation loops operating satisfies Criterion 2 of the NRC Policy Statement (Ref. 6).

(continued)

BFN-UNIT 3 B 3 '-3 Amendment I

ECCS -Operating B 3.5.1 BASES BACKGROUND The HPCI System is designed to provide core cooling for a (continued) wide range of reactor pressures (150 psig to 1174 psig).

Upon receipt of an initiation signal, the HPCI turbine stop valve and turbine control valve open and the turb'ine accelerates to a specified speed. As the HPCI flow increases, the turbine governor valve is automatically adjusted to maintain design flow. Exhaust steam from the HPCI turbine is discharged to the suppression pool. A full flow test line is provided to route water from and to the CST to allow testing of the HPCI System during normal operation without injecting water into the RPV.

The ECCS pumps are provided with minimum flow bypass lines, which discharge to the suppression pool. The valves in these lines automatically open (for CS and RHR they are already open) to prevent pump damage due to overheating when other discharge line valves are closed. To ensure rapid delivery of water to the RPV and to minimize water hammer effects, all ECCS pump discharge lines are filled with water. The LPCI and CS System discharge lines are kept full of water using the pressure suppression chamber head tank or condensate head tank. The HPCI System is normally aligned to the CST. The height of water in the CST is sufficient to maintairi the piping full of water up to the first isolation valve. The relative height of the feedwater line connection for HPCI is such that the water in the feedwater lines keeps the remaining portion of the HPCI discharge line full of water.

The ADS (Ref. 4) consists of 6 of the 13 S/RVs. It is designed to provide depressurization of the RCS during a small break LOCA if HPCI fails or is unable to maintain required water level in the RPV. ADS operation reduces the RPV pressure to within the operating pressure range of the low pressure ECCS subsystems (CS and LPCI), so that these subsystems can provide coolant inventory makeup. Each of the S/RVs used for automatic depressurization is equipped with one air accumulator and associated inlet check valves.

The accumulator provides the pneumatic power to actuate the valves.

APPLICABLE The ECCS performance is evaluated for the entire spectrum of SAFETY ANALYSES break sizes for a postulated LOCA. The accidents for which ECCS operation is required are presented in References 5 (continued)

BFN-UNIT 3 B 3.5-4 Amendment

ECCS -Operating B 3.5.1 BASES SURVEILLANCE SR 3.5. 1.5 (continued)

RE(UIREHENTS The specified Frequency is once per 31 days. However, this SR is modified by a Note that states the Surveillance is only required to be performed prior to entering MODE 2 from MODE 3 or 4, when in HODE 4 > 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . Verification during or following each entry into MODE 4 > 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to entering MODE 2 from MODE 3 or 4 is an exception to the normal Inservice Testing Program generic valve cycling Frequency of 92 days, but is considered acceptable due to the demonstrated reliability of these valves. The 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months is intended to indicate an outage of sufficient duration to allow for scheduling and proper performance of the Surveillance. If the valve is inoperable and in the open position, the associated LPCI subsystem must be declared inoperable.

SR 3.5.1.6 SR 3.5.1.7 and SR 3.5.1.8 The performance requirements of the low pressure ECCS pumps are determined through application of the 10 CFR 50, Appendix K criteria (Ref. 7). This periodic Surveillance is performed (in accordance with the ASHE Code,Section XI, requirements for the ECCS pumps) to verify that the ECCS pumps will develop the flow rates required by the respective analyses. The low pressure ECCS pump flow rates ensure that adequate core cooling is provided to satisfy the acceptance criteria of Reference 13. The pump flow rates are verified against a system head equivalent to the RPV pressure expected during a LOCA. The total system pump outlet pressure is adequate to overcome the elevation head pressure between the pump suction and the vessel discharge, the piping friction losses, and RPV pressure present during a LOCA. These values may be established by testing or analysis or during preoperational testing.

The flow tests for the HPCI System are performed at two different pressure ranges such that system capability to provide rated flow is tested at both the higher and lower operating ranges of the system. Additionally, adequate steam flow must be passing through the main turbine or turbine bypass valves to continue to control reactor pressure when the HPCI System diverts steam flow. 'eactor steam pressure must be a 950 psig to perform SR 3.5. 1.7 and

) 150 psig to perform SR 3.5. 1.8. Adequate steam flow is (continued)

BFN-UNIT 3 B 3.5-12 Amendment

RCIC System B 3.5.3 B 3.5 EMERGENCY CORE COOLING SYSTEHS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEH B 3.5.3 RCIC System BASES BACKGROUND The RCIC System is not part of the ECCS; however, the RCIC System is included with the ECCS section because of their similar functions.

The RCIC System is designed to operate either automatically or manually following reactor pressure vessel (RPV) isolation accompanied by a loss of coolant flow from the feedwater system to provide adequate core cooling and control of the RPV water level. Under these conditions, the High Pressure Coolant Injection (HPCI) and RCIC systems perform similar functions. The RCIC System design requirements ensure that the criteria of Reference I are satisfied.

The RCIC System (Ref. 2) consists of a steam driven turbine pump unit, piping, and valves to provide steam to the turbine, as well as piping and valves to transfer water from the suction source to the core via the feedwater system line, wher e the coolant is distributed within the RPV through the feedwater sparger. Suction piping is provided from the condensate storage tank (CST) and the suppression pool. Pump suction is normally aligned to the CST to minimize injection of suppression pool water into the RPV.

However, if the CST water supply is low, or the suppression pool level is high, a manual transfer to the suppression pool water source ensures a water supply for continuous operation of the RCIC System. The steam supply to the turbine is piped from a main steam line upstream of the associated inboard main steam line isolation valve.

The RCIC System is designed to provide core cooling for a wide range of reactor pressures (150 psig to ll74 psig).

Upon receipt of an initiation signal, the RCIC turbine accelerates until a 600 gpm flow rate (design flow) is achieved. As the RCIC turbine flow varies, the turbine control valve is-automatically adjusted to maintain design flow. Exhaust steam from the RCIC turbine is discharged to the suppression pool. A full flow test line is provided to route water from and to the CST to allow testing of the RCIC (continued)

BFN-UNIT 3 B 3.5-24 Amendment

g 4 RCIC System B 3.5.3 SURVEILLANCE SR 3.5.3.2 (continued)

REQUIREMENTS steam flow path for the turbine and the flow controller position.

The 31 day Frequency of this SR was derived from the Inservice Testing Program requirements for performing valve testing at least once every 92 days. The Frequency of 31 days is further justified because the valves are operated under procedural control and because improper valve position would affect only the RCIC System. This Frequency has been shown to be acceptable through operating experience.

SR 3.5.3.3 and SR 3.5.3.4 The RCIC pump flow rates ensure that the system can maintain reactor coolant inventory during pressurized conditions with the RPV isolated. The flow tests for,, the RCIC System are performed at two different pressure ranges such that system capability to provide rated flow is tested both at the higher and lower operating ranges of the system.

Additionally, adequate steam flow must be passing through the main turbine or turbine bypass valves to continue to control reactor pressure when the RCIC System diverts steam flow. Reactor steam pressure must be z 950 psig to perform SR 3.5.3.3 and a 150 psig to perform SR 3.5.3.4. Therefore, sufficient time is allowed after adequate pressure is achieved to perform these SRs. Reactor startup is allowed prior to performing the low pressure Surveillance because the reactor pressure is low and the time allowed to satisfactorily perform the Surveillance is short.

Alternately, the low pressure Surveillance test may be performed prior to startup using an auxiliary steam supply.

The reactor pressure is allowed to be increased to normal operating pressure since it is assumed that the low pressure Surveillance has been satisfactorily completed and there is no indication or reason to believe that RCIC is inoperable.

Therefore, these SRs are modified by Notes that state the Surveillances are not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after the reactor steam pressure and flow are adequate to perform the test.

(continued)

BFN-UNIT 3 B 3.5-28 Amendment

Primary Containment B 3.6.1.1 BASES (continued)

APPLICABLE The safety design basis for the primary containment is that SAFETY ANALYSES it must withstand the pressures theand temperatures ofrate. the I

limiting DBA without exceeding design leakage The DBA that postulates the maximum release of radioactive material within primary containment is a LOCA. In the analysis of this accident, it is assumed that primary containment is OPERABLE such that release of fission products to the environment is controlled by the rate of primary containment leakage.

Analytical methods and assumptions involving the primary containment are presented in References 1 and 2. The safety analyses assume a nonmechanistic fission product release following a DBA, which forms the basis for determination of offsite doses. The fission product release is, in turn, based on an assumed leakage rate from the primary containment. OPERABILITY of the primary containment ensures that the leakage rate assumed in the safety analyses is not-exceeded.

allowable leakage rate for the primary

~,

The maximum containment (L,) is 2.0'X by weight of the containment air per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at the design basis LOCA maximum peak

containment pressure (P.) of 50.3 psig (Ref. 1).

Primary containment satisfies Criterion 3 of the NRC Policy Statement (Ref. 6).

LCO Primary containment OPERABILITY is maintained by limiting leakage to x 1.0 L., except prior to the first startup after performing a required Primary Containment Leakage Rate Testing Program leakage test. At this time, applicable leakage limits must be met. Compliance with this LCO will ensure a primary containment configuration, including equipment hatches, that is structurally sound and that will limit leakage to those leakage rates assumed in the safety analyses.

Individual leakage rates specified for the primary containment air lock are addressed in LCO 3.6. 1.2.

(continued)

BFN-UNIT 3 B 3.6-2 Amendment

Primary Containment Air Lock B 3.6.

1.2 BACKGROUND

and leak tightness are essential 'for maintaining primary (continued) containment leakage rate to within limits in the event of a DBA. Not maintaining air lock integrity or leak tightness may result in a leakage rate in excess of that assumed in the unit safety analysis.

APPLICABLE The DBA that postulates the maximum release of radioactive SAFETY ANALYSES material within primary containment is a LOCA. In the analysis of this accident, it is assumed that primary containment is OPERABLE, such that release of fission products to the environment is controlled by the rate of primary containment leakage. 'The primary containment is designed with a maximum allowable leakage rate (L.) of 2.05 by weight of the containment air per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at the calculated maximum peak containment pressure (P.) of 50.3 psig (Ref. 3). This allowable leakage rate forms the basis for the acceptance criteria imposed on the SRs associated with the air lock.

Primary containment air lock OPERABILITY is also required to minimize the amount of fission product gases that may escape primary containment through the air lock and contaminate and pressurize the secondary containment.

The primary containment air lock satisfies Criterion 3 of the NRC Policy Statement (Ref. 4).

LCO As part of primary containment, the air lock's safety function is related to control of containment leakage rates following a DBA. Thus, the air lock's structural integrity and leak tightness are essential to the successful mitigation of such an event.

The primary containment air lock-is required to be OPERABLE.

For the air lock to be considered OPERABLE, the air lock interlock mechanism must be OPERABLE, the air lock must be in compliance with the Type 8 air lock leakage test, and both air lock doors must be OPERABLE. The interlock allows only one air lock door to be opened at a time. This provision ensures that a gross breach of primary containment does not exist when primary containment is required to be (continued)

BFN-UNIT 3 B 3.6-7 Amendment

~,

RHRSW System B 3.7.1 B 3.7 PLANT SYSTEHS B 3.7. 1 Residual Heat Removal Service Mater (RHRSM) System BASES BACKGROUND The RHRSW System is designed to provide cooling water for the Residual Heat Removal (RHR) System heat exchangers, required for a safe reactor shutdown following a Design

- Basis Accident (DBA) or, transient. The RHRSW System is operated whenever the RHR heat exchangers are required to operate in the shutdown cooling mode or in the suppression pool cooling or spray mode of the RHR System.

The RHRSW System is common to the three BFN units and consists of the UHS and four independent and redundant, loops, each of which feeds one RHR heat exchanger in each unit. Each loop is made up of a header, two 4500 gpm pumps, a suction source, valves, piping, and associated instrumentation. One loop with one pump operating is capable of providing 50% of the required cooling capacity to maintain safe shutdown conditions for one unit following a design basis accident. However, one pump is capable of providing sufficient cooling capacity to maintain a safe 0 shutdown condition for each of the non-accident units. As such, a subsystem consists of a loop with one or two OPERABLE pumps dependent upon the number of fueled units.

The RHRSW System is designed with sufficient redundancy so that no single active component failure can prevent it from achieving its design function. The RHRSW System is described in 'the FSAR, Section 10.9 (Ref. I).

Cooling water is pumped by the RHRSW pumps from the Wheeler Reservoir through the tube side of the RHR heat exchangers, and discharged back to the Wheeler Reservoir.

The system is initiated manually from each of the three units control rooms. If operating during a loss of coolant accident (LOCA), the system is automatically tripped on degraded bus voltage to allow the diesel generators to automatically power only that equipment necessary to reflood the core. The system can be manually started any time the degraded bus voltage signal. is manually overridden or clears, and is assumed to b'e manually started within 10 minutes after the LOCA.

(continued)

BFN-UNIT 3 B 3.7-1 Amendment

RHRSM System B 3.7.1 BASES (continued)

APPLICABLE The RHRSW System removes heat from the suppression pool to SAFETY ANALYSES limit the suppression pool temperature and primary containment pressure following a LOCA. This ensures that the primary containment can perform its function of limiting the release of radioactive materials to the environment following a LOCA. The ability of the RHRSW System to support long term cooling of the reactor or primary containment is discussed in the FSAR, Chapters 5 and 14 (Refs. 2 and 3, respectively). These analyses explicitly assume that the RHRSM System will provide adequate cooling support to the equipment required for safe shutdown. These analyses include the evaluation of the long term primary containment response'fter a design basis LOCA.

The safety analyses for long term cooling were performed for various combinations of RHR System failures and considers the number of units fueled. With one unit fueled, the worst case single failure that would affect the performance of the RHRSW System is any failure that would disable two subsystems or pumps of the RHRSW System (e.g, the failure of an RHR Suppression Pool Cooling/Spray return line valve which effectively disables two RHRSW subsystems or pumps).

With two and three units fueled, a worst case single failure could also include the loss of two RHRSW pumps caused by losing a 4 kV shutdown board since there are certain alignment configurations that allow two RHRSW pumps to be powered from the same 4 kV shutdown board. As discussed in the FSAR, Section 14.6.3.3.2 (Ref. 4) for, these analyses, manual initiation of the OPERABLE RHRSW subsystems and the associated RHR System is assumed to occur 10 minutes after a DBA. The RHRSW flow assumed in the analyses is 4000 gpm per pump with two pumps operating in one loop. In this case, the maximum suppression chamber water temperature and pressure are 177'F (as reported in Reference 3) and 50.3 psig, respectively, well below the design temperature of 281'F and maximum allowable pressure of 62 psig. This is also below the 200'F limit imposed by Design Criteria BFN-50-7064A (Ref. 5) for all plant transients involving SRV operations.

The RHRSM System, together with the UHS, satisfies Criterion 3 of the NRC Policy Statement (Ref 6).

(continued)

BFN-UNIT 3 B 3.7-2 Amendment

RHRSW System B 3.7.1 BASES (continued)

LCO Four RHRSW subsystems are required to be OPERABLE to provide the required redundancy to ensure that the system functions to remove post accident heat loads, assuming the worst case single active failure occurs coincident with the loss of offsite power.

An OPERABLE RHRSW subsystem consists of:

a. The required number of OPERABLE RHRSW pumps dependent upon the number of units fueled; and

b. An OPERABLE flow path capable of taking suction from the intake structure and transferring the water to the required RHR heat exchangers at the assumed flow rate.

The LCO is modified by two Notes. Note 1 specifies that when 1 or 2 units are fueled, there must be at least one OPERABLE pump per RHRSW subsystem. Note 2 specifies that when 3 units are fueled, two of the RHRSW subsystems, must have two OPERABLE RHRSW pumps.

The OPERABILITY of the UHS for RHRSW is based on having a maximum water temperature within the limits specified in-Figure 3.7.1-1.

APPLICABILITY In MODES 1, 2, and 3, the RHRSW System and UHS are required to be OPERABLE to support the OPERABILITY of the RHR System for primary containment cooling (LCO 3.6.2.3, "Residual Heat Removal (RHR) Suppression Pool Cooling," and LCO 3.6.2.4, "Residual Heat Removal (RHR) Suppression Pool Spray" ) and decay heat removal (LCO 3.4.7, "Residual Heat Removal (RHR)

Shutdown Cooling System- Hot Shutdown" ). The Applicability is therefore consistent with the requirements of these systems.

In MODES 4 and 5, the OPERABILITY requirements of the RHRSW System and UHS are determined by the systems they support.

ACTIONS The Actions are modified by a Note indicating that the applicable Conditions of LCO 3.4.7 be entered and Required Actions taken if the inoperable RHRSW subsystem results in inoperable RHR shutdown cooling. This is an exception to (continued)

BFN-UNIT 3 3.7-3 Amendment

RHRSW System B 3.7.1 BASES ACTIONS LCO 3.0.6 and ensures the proper actions are taken for these

'continued) components.

A.1 Mith one RHRSM subsystem or required pump inoperable, the inoperable RHRSM subsystem or required pump must be restored to OPERABLE status within 30 days. Mith the unit in this condition, the remaining OPERABLE RHRSM subsystems are adequate to perform the RHRSW heat removal'function.

However, the overall reliability is reduced because a single failure could result in'educed primary containment cooling capability. The 30 day Completion Time is based on the availability of equipment in excess of normal redundancy requirements and the low probability of an event occurring requiring RHRSW during this period.

C.

B. 1 8.2 and B.3 Required Action B. 1'requires verification that at least three OPERABLE RHRSW pumps are associated with the two OPERABLE RHRSM subsystems. The Required Action is modified by a Note indicating that the required action is applicable only when two units are fueled. Required Action 8.2 requires verification that at least four OPERABLE RHRSW pumps are associated with the two OPERABLE RHRSW subsystems.

The Required Action is modified by a Note indicating that the required action is applicable only when three units are fueled.

Required Action B.3 requires that with two RHRSM subsystems inoperable, one inoperable RHRSW subsystem be restored to OPERABLE status within 7 days. With the unit(s) in this condition, the remaining OPERABLE RHRSW subsystems are adequate to perform the RHRSW heat removal function.

However, the overall reliability is reduced because a single failure in the OPERABLE RHRSW subsystems could result in loss of RHRSW function. The 7 day Completion Time is based on the redundant RHRSW capabilities afforded by the OPERABLE subsystems and the low probability of an event occurring requiring RHRSW during this period.

(continued)

BFN-UNIT 3 B 3.7-4 Amendment

0 1

4

RHRSW System B 3.7.1 BASES ACTIONS C.1 (continued)

Mith three or more RHRSM subsystems inoperable, the RHRSW System is not capable of performing its intended function.

The requisite number of subsystems and pumps must be restored to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . The 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Completion Time is based on the Completion Times provided for the RHR suppression pool cooling and spray functions.

D. 1 and 0.2 If the RHRSW subsystems cannot be restored to OPERABLE status within the associated Completion Times or the UHS is determined inoperable, the unit must be placed in a NODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3.within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in NODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the t

required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

SURVEILLANCE SR 3".7.1.1 REQUIREMENTS Verifying the correct alignment for each manual and power operated valve in each RHRSM subsystem flow. path provides assurance that the proper flow paths will exist for RHRSW operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these valves are verified to be in the correct position prior to locking, sealing, or securing. A valve is also allowed to be in the nonaccident position, and yet considered in the correct position, provided it can be realigned to its accident position. This is acceptable because the RHRSW System is a manually initiated system.

This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves.

(continued)

BFN-UNIT 3 B 3.7-5 Amendment

RHRSW System B 3.7.1 SURVEILLANCE SR 3.7.1.1 (continued)

RE(UIREHENTS The 31 day Frequency is based on engineering judgment, is consistent with the procedural controls governing valve operation, and ensures correct valve positions.

I I ~

SR 3.7.1.2 Verification of the UHS temperature ensures that the heat removal capability of the RHRSW System is within the assumptions of the DBA analysis. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months frequencies are based on operating experience relating to trending of the parameter variations during the applicable NODES.

REFERENCES 1. FSAR, Section 10.9.

2. FSAR, Chapter 5.

3. FSAR, Chapter 14.

4. FSAR, Section 14.6.3.3.2.

5. Design Criteria BFN-50-7064A, Primary Containment Systems - Units 2 and 3

6. NRC No.93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.

BFN-UNIT 3 B 3.7-6 Amendment

0 EECM System and UHS B 3.7.2 B 3.7 PLANT SYSTEMS B 3.7.2 Emergency Equipment Cooling Mater (EECM) System and Ultimate Heat Sink (UHS)

BASES BACKGROUND The EECW System is designed to provide cooling water for the removal of heat from equipment, such as the diesel generators (DGs), residual heat removal (RHR) pump coolers, and room coolers for other Emergency Core Cooling System equipment, required for a safe reactor shutdown. following a Design Basis Accident (DBA) or transient. The EECM System also provides cooling to unit components, as required, during normal operation. Upon receipt of a loss of offsite power or loss of coolant accident (LOCA) signal, the essential loads are provided cooling water by automatically starting RHRSW pumps aligned to EECM headers.

The EECW System, which is common to the three BFN units, consists of the UHS and two independent and redundant loops with each loop consisting of a header, two 4500 gpm pumps, a suction source, valves, piping and associated instrumentation. Two EECM pumps (one per loop or both on one loop) are capable of providing the required cooling capacity to support the required systems. The two loops are separated from each other so failure of one loop will not affect the OPERABILITY of the other. The EECM System is described in the FSAR, Section 10. 10 .(Ref. 3)

Cooling water is pumped from the Mheeler Reservoir by the EECW pumps to the essential components through the two main headers. After removing heat from the components, the water is discharged back to the Wheeler Reservoir.

APPLICABLE Sufficient water inventory is available for all EECW System SAFETY ANALYSES post LOCA cooling requirements for a 30 day period with no additional makeup water source available. The ability of the EECW System to support long term cooling of the reactor containment is assumed in evaluations of the equipment required for safe reactor shutdown presented in the FSAR, Chapters 5 and 14 (Refs. 2 and 3, respectively). These analyses include the evaluation of the long term primary containment response after a design basis LOCA.

(continued)

BFN-UNIT 3 8 3.7-7 - Amendment

EECM System and UHS B 3.7.2 BASES APPLICABLE The ability of the EECM System to provide adequate cooling SAFETY ANALYSES to the identified safety equipment is an implicit assumption (continued) for the safety analyses evaluated in References 1 and 2.

The ability to provide onsite emergency AC power is dependent on the ability of the EECW System to cool the DGs.

The long term cooling capability of the RHR and core spray pumps is also dependent on the cooling provided by the EECW System.

The EECW System, together with the UHS, satisfy Criterion 3, of the NRC Policy Statement (Ref. 4).

h LCO The EECW loops are independent of each other to the degree that each has separate controls, power supplies, and the operation of one does not depend on the other. In the event of a DBA, two EECW pumps are required to provide the minimum heat removal capability assumed in the safety analysis for the system to which it supplies cooling water. To ensure this requirement is met, three EECW pumps must be OPERABLE.

At least two pumps will operate if the worst single active failure occurs coincident with the loss of offsite power.

The EECW System is considered OPERABLE when it has an OPERABLE UHS, three OPERABLE pumps, and two OPERABLE flow paths capable of taking suction from the intake structure and transferring the water to the appropriate equipment.

The OPERABILITY of the UHS for EECW is based on having a maximum water temperature of 95'F. Additional requirements'or UHS temperatures are'rovided in SR 3.7. 1.2.

The isolation of the EECW System to components or systems may render'those components or systems inoperable, but does not affect the OPERABILITY of the EECW System.

APPLICABILITY In NODES 1, 2, and 3, the EECW System and UHS are required to be OPERABLE to support OPERABILITY of the equipment serviced by the EECW System. Therefore, the EECW System and UHS are required to be OPERABLE in these MODES.

In MODES 4 and 5, the OPERABILITY requirements of the EECW System and UHS are determined by the systems they support.

(continued)

BFN-UNIT 3 B 3.7-8 Amendment

EECW System and UHS B 3.7.2 ACTIONS A. 1 With one required EECW pump inoperable, the required EECW pump must be restored to OPERABLE status within 7 days.

With the system in this condition, the remaining OPERABLE EECW pumps are adequate to perform the heat removal function. However, the overall reliability is reduced because a single failure in the EECW System could result in loss of EECW function. e The 7 day Completion Time is based on the redundant EECW System capabilities afforded by the remaining OPERABLE pumps, the low probability of an accident occurring during this time pe'riod and is consistent with the allowed Completion Time for restoring an inoperable DG; B.l and B.2 If the required EECW pump cannot be restored to OPERABLE status within the associated Completion Time, or two or more EECW pumps are inoperable or the UHS is determined inoperable, the unit must be placed in a NODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least HODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in NODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

SURVEILLANCE SR 3.7.2.1 RE(UIREHENTS Verification of the UHS temperature ensures that the heat

. removal capability of the EECW System is within the assumptions of the DBA analysis. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency is based on operating experience related to trending of the parameter variations during the applicable NODES.

SR 3.7.2.2 Verifying the correct alignment for each manual and power operated valve in the EECW System flow paths provide assurance that the proper flow paths will exist for EECW (continued)

BFN-UNIT 3 B 3.7-9 Amendment

0 EECW System and UHS B 3.7.2 BASES SURVEILLANCE 2 2 ( i""d)

REQUIREHENTS operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these valves were verified to be in the correct position prior to locking, sealing, or securing. A valve is also allowed to be'in the nonaccident position, and yet considered

" automatically in the correct position, provided it can be realigned to its accident position within the required time. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves.

This SR is modified by a Note indicating that isolation of the EECW System to components or systems may render those components or systems inoperable, but does not affect the OPERABILITY of the EECW System. As such, when required EECW pumps, valves, and piping are OPERABLE, but a branch connection off the main header is isolated, the EECW System is still OPERABLE.

The 31 day Frequency is based on engineering judgment, is consistent with the procedural controls governing valve operation, and ensures correct valve positions.

SR 3.7.2.3 This SR verifies that the EECW System pumps will automatically start to provide cooling water to the required safety related equipment during an accident event. This is demonstrated by the use of an actual or simulated initiation signal.

Operating experience has shown that these components will usually pass the SR when performed at the 18 month Frequency. Therefore, this Frequency is concluded to be

'cceptable from a reliability standpoint.

(continued)

BFN-UNIT 3 B 3.7-10 Amendment

WZ EECW System and UHS B 3.7.2 BASES (continued)

REFERENCES , 1. FSAR, Chapter 5.

2. FSAR, Chapter 14.

3. FSAR, Section 10. 10.

4. NRC No.93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.

BFN-UNIT 3 B 3.7-11 Amendment

0

ML18038B961

Contents

Text

3.4.3 APPLICABILITY

5.1 BACKGROUND

1.2 BACKGROUND

1.2 BACKGROUND

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ML18038B961

Text

3.4.3 APPLICABILITY

5.1 BACKGROUND

1.2 BACKGROUND

1.2 BACKGROUND

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