Text

Amends 159,155 & 130 to Licenses DPR-33,DPR-52 & DPR-68, Respectively,Adding ASME Section XI Pump & Valve Program Definition & Changing Applicable Testing Frequencies to Ref Definition
	ML20196B320
Person / Time
Site:	Browns Ferry
Issue date:	11/22/1988
From:	Black S; NRC OFFICE OF SPECIAL PROJECTS
To:	;
Shared Package
ML20196B329	List: ML20196B320; ML20196B376;
References
	TAC-00250, TAC-00251, TAC-00252, TAC-250, TAC-251, TAC-252, NUDOCS 8812060243
	Download: ML20196B320 (113)
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UNITED STATES

! ",,. [ g NUCLEAR REGULATORY COMMISSION g

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WASHINGTON, D. C. 20555 t

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TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-259 BR0kNS FERRY NUCLEAR PLANT, UNIT 1 AMEN 0 PENT TO FACILITY OPERATING LICENSE Amendment No.157 License ho, DPR-33 1.

The Nuclear Regulatory Comission (the Comission) has found that:

A.

The application for amendment by Tennessee Valley Authority (the licensee) dated October 27, 1987, coeplies with the standards and requirerrents of the Atomic Energy Act of 1954, as amended (the Act),

and the Comission's rules and regulations set forth in 10 CFP Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and r39ulations of the Commission; C.

There is reasonable assurance (1) that the activities authorized by this a m ndment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.

The issuance of this amendrent will not be inimical to the co w.on defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Coernission's regulations and all applicable requirements have been satisfied.

8312060243 881122 POR ADOCK 05000259 P

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s to 2-Accordingly, the license is amended by changes to the Technical 2.

Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No. OPR-33 is hereby arended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.159, are hereby incorporated in the license.

The licensee shall operate the Q :ility 'n accordance with the Technical Specifications.

This license amendment is ef fective as of its date of issuance and shall 3.

be implemented within 60 days from the date of issuance.

FOR THE NUCLEAR REGULATORY COMISSION L C.k tl % Q Suzanne Black, Assistant Director j

for Projects j

TVA Projects Division l

Office of Special Projects l

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Attachment:

Changes to the Technical l

6 Specifications

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Date of Issuance: Novembei 22, 1983 i

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s ATTACHMENTTOLICENSEAMENDMENTNO.1YI FACILITY OPERATING LICENSE NO. OPR-33, DOCKET NO. 50-259 Pevise the Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages. The revised paqes are identified by the captioned amendment number and contain margina?

lines indicating the area of chance. Overlesf pages* are provided to maintain document completeness.

REMOVE IN9EPT 1.0-11 1.0-11*

1,0-12 1.0-12 1.0-13 3.3/4.3-11 3.3/4.3-11*

3.3/4.3-12 3.3/4.3-12 3.3/4.4-1 3.3/4.4-1 3.3/4.4-2 3.3/4.4-2*

3.5/4.5-1 3.5/4.5-1 3.5/4.5-2 3.5/4.5-2 3.5/4.5-3 3.5/4.5-3*

3.5/4.5-4 3.5/4.5-4 3.5/4.5-7 3.5/4.5-7 3.5/4.5-8 3.5/4.5-8*

3.5/4.5-9 3.5/4.5-9 3.5/4.5-10 3.5/4.5-10*

3.5.4.5-12 3.5/4.5-12*

as 3.5/4.5-13 3.5/4.5-13 3.5/4.5-14 3.5/4.5-14 3.5/4.5-15 3.5/4.5-15*

l 3.5/4.5-34 3.5/4.5 34*

3.5/4.5-35 3.5/4.5-35 3.6/4.6-9 3.6/4.6-9*

3.6/4.6-10 3.6/4.6-10 3.6/4.6-30 3.6/4.6-30 3.6/4.6-31 3/6/4.6-31*

3.7/4.7-9 3.7/4.7-9 3.7/4.7 10 3.1/4.7-10 3.7/4.7-11 3.7/4.7-11 3.7/4.7-12 3.7/4.7-17 3.7.4.7-17 3.7/4.7-17 l

3.7.4.7 18 3.7/4.7-18 3.7/4.7-21 3.7/4.7-21*

3.7/4.7-22 3.7/4.7-2?

3.7/4.7-49 3.7/4.7-49*

3.7/4.7-50 3.7/4.7-50 i

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.0 DEFINITIONS (Cont'd)

GG.

Site Boundary - Shall be that line beyond which the land is not owned, leased, or otherwise controlled by TVA.

HH.

Unrestricted Area - Any area at or beyond the site boundary to which access is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials or any area within the site boundary used for industrial, commercial, institutional, or recreational i

purposes.

l II.

Dese Eauivalent I-Lil - The D09E EQUIVALENT I-131 shall be the concentration of I-121 (in Ci/gm) which ' lone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132. I-133, I-134, and I-135 actually present. The thyroid dose conversion factor used for this calculation shall be those listed in Table III of TID-14844 "Calculation of Distance Factors for Power and Test Peactor Sites",

i JJ.

Gaseous Waste Treatment System - The charcoal adsorber vessels installed on the discharge of the steam jet air ejector to provide delay to a unit's offgas activity prior to release.

KK.

Members of the Public - Shall include all individuals who by virtue of their occupational status have no formal association l

vith the plant. This category shall include non-employees of i

the licensee who are permitted to use portions of the site for recreational, occupational, or other purposes not associated with plant functions. This category shall n21 include non-employees such as vending machine servicemen or postmen l

vho, as part of their formal job function, occasionally enter restricted areas.

j LL.

Surveillance - Surveillance Requirements shall be met during the OPERATIONAL CONDITIONS or other conditions specified for individual limiting conditions for operation unless otherwise i

stated.in an individual Surveillance Requirements. Each j

surveillance Requirement shall be performed within the I

spr,cified time interval with, l

(1)

A maximum allevable extention not to exceed 25% of the surveillance interval, but j

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i (2)

The combined time entered for any 3 consecutive

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surveillance intervals shall not exceed 3.25 times the specified surveillance interval Performance of a Survelliance Requirement within the specified

)

time interval shall constitute compliance and OPERABILITY requirements for a limiting condition for operation and asscciated action statements unless otherwise required by these I

specifications. Surveillance requirements do not have to be i

performed on inoperable equipment.

ITM 1.0-11 Unit 1 s

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1.0 DETINITION3 (Ccnt'd)

MM. Surveillance reauirements for ASME Section XI Pumo and Valve Fronram - Surveillance requirements for Inservice Testing of ASME Code Class 1, 2, and 3 components shall be applicable as follows:

1.

Inservice testing of ASME Code Class 1, 2, and 3 pumps and valves shall be performed in.accordance with Section XI of the.'4SME Boiler and Pressure Code and applicable Addenda as required by 10 CTR 50, Section 50.55a(s), except where specific written relief has been granted by the Commission pursuant to 10 CTR 50, Section 50.55a(g)(6)(1).

2.

Surveillance intervals specified in Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda for the inservice testing activities required by the ASME Boilar and Pressure Vessel Code and applicable Addenda shall be applicable as follows in these technical specifications:

ASME Boiler and Pressure Vessel Requir!d frequencies Code and applicable Addenda for performing inservice t e rminolog*f for inservice testina activities testine activities Weekly At least once per 7 days Monthly At least once per 31 days Quarterly or every 3 months At least once per 92 dayr Semiannually or every 6 months At least once per 184 day.

Every 9 months At least once per 276 dtys Yearly or annually At least once per 366 days 3.

The provisions of Specification 1.0 LL are applicable to the above required frequencies for performing inservice 4

testing activities.

4.

Performance of the above inservice testing activities shall be in addition to other specified surveillance requirem.nts.

5.

Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any technical specification.

j BFN 1.0-12 A~end en' No. 15?

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se Table 1.1 SURVIIf.fAMCE TREQUENCY NOTATION

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NOTATION FRIOUENCY S

(Shift)

At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

D (Daily)

At least once per normal calendar 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months day (midnight to midnight).

W (Weekly)

At least once per 7 days.

M (Monthly)

At least once pct 31 days.

j Q

(Quarterly)

At least once per 3 months or 92 days.

EA (Semi-Annually)

At least once per 6 months or 184 days.

I Y

(Yearly)

At least once per year or 366 days.

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R (Refuelins)

At least once per operating cycle.

l S/U (Start fjp)

Prior to each reactor startup.

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N.A.

Not applicable.

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(Prior)

Completed prior to each release.

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J.3/4.3 RIACTIVITY CONTROL L;MITIt0 CONDITIONS FOR OPERATION SURVEILLANCE RIQUIREMENTS 3.3.C.

Scram Insertion Times 4.3.C.

Scram Insertion Times 2.

The average of the scram inter-

2. At 16-veek intervals, 10%

tion times for the three fastest of the operable control operable control rods of all rod drives shall be scram-groups of four control rods in timed above 300 pais, a two-by-two array shall be.so Whenever such scram time greater thans measurements are made, an evaluation shall be made

% Inserted Fron Avg. Scram Inser-to provide reasonable ru11v Withdrawn tion Times (see) casurance that proper control rod drive 5

0.398 performance is being 20 0.954 maintained.

50 2.120 90 3.800 a.

The maximum scram insertion time for 90% insertion of any operable control rod shall not l

exceed 7.00 seconds.

a D.

Reactivity Ancmalies D.

Reactivity Anomalies The reactivity equivalent of During the startup test

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the difference between the program and startup following actual critical rod refueling outages, the I

configuration and the expected critical rod configurations j

configuration during power vill be compared to the l

operation shall not exceed 1%

k.

expected configurations at 4

i If this limit is exceeded, the selected operating conditions.

reactor vill be shut down These comparisons will be

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j until the cause has been used as base data for determined and corrective reactivity monitoring during i

i actions have been taken as subsequent power operation l

i appropriate, throughout the fuel cycle.

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1 At specific power operating l

conditions, the critical rod configuration vill be compared to the configuration l

expected based upon appropriately corrected past data. This comparison vill be made at least every full l

power month.

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3.3/A.3 REACT *VITY CONTROL JIMITING CONDITIONS PA: OPERATION SURVEILLANCE REQUIREMENTS 3.3.E.

If Specifications 3.3.C and.D 4.3.E.

Surveillance requirements are above cannot be met, an orderly as specified in 4.3.C and.D shutdown shall be initiated and

above, the reactor shall be in the SHUTDOWN CONDITION within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

T.

deram Discharme Volume (SDV)

F.

Erram Discharme Volume (SDV}

1.

The scram discharge volume 1.a. The scran discharge drain tnd veat valves shall volume drain and vent be OPERABLE any time that valves shall be verified the reactor protection open prior to each system !s required to be startup and monthly CPERABL% except as thereafter. The valves specified in 3.3.T.2.

may be closed intermittently for testing not to exceed i

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months in any 24-hour i

period during operation.

1.b. The scram discharge volume drain and vent valves shall be demonstrated CPERABLE in accordance with Specification 1.0.MM.

l 2.

In the event any SDV drain 2.

When it is determined or vent valve becomes that any SDV drain or inoperable, REACTOR POWER vent valve is inoperable, OPERATION may continue the redundant drain or provided the redundant vent valve shall be drain or vent valve is demonstrated OPERABLE OPE RABLE.

immediately and weekly thereafter.

3.'

If redundant drain or vent 3.

He additional valves bscome inoperable, surveillance required.

the reactor shall be in HOT STANDBY CONDITION within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

BTN 3.3/4.3-12 A end ent i;o.133,1!?

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3.4/4.4 STANDBY L M 11.D CONTROL SYSTEM LIMITING CONDITION! FOR OPERATI E SURVEILLANCE REQUIREMENTS 3.4 STANLBY LIQUID M NTROL SY$1 M 4.4 STANDBY Lf0UID CONTROL SYSTg Acelicability Aeolicability Applies to the operating status Applies to the surveillance j

of the Standby Liquid Control requiressnts of the Standby System.

Ligtid Control System.

Qpiective Obiective To assure the availability fa To verify the operability of the system with the capability to Standby Liquid Control System.

shut devn the reactor and maintain the shutdown condition without the use of centrol rods.

Seecificatien Seecificati33 A.

Normal System Availability A.

Normal System Availability 1.

Except as specified in The operability of the Standby 3.4.B.1, the Standby Liquid Liquid Control System shall be Control System shall be verif ted by the performance OPERABLE at all times when of the following testat there is fur' in the reactor vessel and the tear'.r is not 1.

Verify pump OPERABILITY in a shutdov. cwast ica with in accordance with Specificatics 3.3.A.1 Specification 1.0.MM.

satisfied.

2.

At least once during each operating cycle::

a.

Check that the setting of the system relief j

valves is 1,425 1 75 l

psig.

b.

Manually initiate the system, except explo-sive valves. Visually verify flow by pumping boren solution through the recirculation path and back to the Standby Liquid Control Solution Tank. After purping boron solution, the system shall be flushed with demineralized vater. Verify minimum ETN 3.4/4.0 1 e n t e n '. *. 0. 136. 154. 159 Unit 1 l'

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o 3.4/4.4 STANDBY Lf oUfD CONTROL SYSTEM LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 4.4.A Normal Svatem Aeolicability 4.4.A.2.b. (Cont'd) pump flow rate of 39 sps against a system head of 12/5 pais by pumping domineralized water from the Standby Liquid Control Test Tank.

c. Manually initiate one of the Standby Liquid Control System loops and pump domineralized water into the reactor vessel.

This test checks explosion of the charge associated with the tested loop, proper operation of the valves, and pump operability.

Replacement charges shall be selected such that the age of charge in service shall not exceed five years from the manufacturer's assembly date.

d.

Both systems, including both explosive valves, shall be tested in the course of two operating cycles.

BTN 3.4/4,4-2

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'3.5/4.5 CORE AND CONTAINMENT COOLING SYSTEMS i

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5 CORE AND CONTAINMENT C0OLING 4.5 CORE AND CONTAINMENT COOLING SYSTEMS SYSTEMS Arelicability Acelicability Applies to the operational Applies to the surveillance status of the core and requirements of the core and containment cooling systems.

containment cooling systems when the correspoeding limiting condi-tion for operation is in effect.

Obiective Oblective To assure the operability of To verify the OPERABILITY of the the core and containment cooling core and containment cooling systems under all conditions for systems under all conditions for which this cooling capability is which this cooling capabili*y is t

an essentisl response to plant an essential response to plant abnormalities.

abnormalities.

Seecification Snecification A.

Core Serav System (CSS)

A.

Core Sorav System (CSS) 1.

The CSS shall be OPERABLE:

1.

Core Spray System Testing (1) PRIOR TO STARTUP 111E Frecuency from a COLD CONDITION, or a.

Simulated once/

Automatic Operating (2) when there is irradiated Actuation Cycle fuel in the vessel test and when the reactor vessel pressure b.

Pump Opera-Per Specif1-is greater than bility cation 1.0.MM atmospheric pressure, except as specified c.

Motor Per Specifi-in Specification Operated cation 1.0.MM 3.5.A.2.

Valve Operability d.

System flov Once/3 rate: Each months loop shall 9

deliver at i

least 6250 l

gpm against a system l

bead corres-

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L ponding to a

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ET3 3.5/4.5-1 d er$ ment '.0.

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e 3.5/4.5 CORE AND CONTAINMENT COOLING SYSTEMS L.MITIN0 CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.A core Serav System (CSS) 4.5.A Core Serav system (CSS) l 4.5.A.1.d (Cont'd) 105 psi differential l

pressure between the reactor vessel and the primary i

I containment.

e.

Check Valve Per Specification 1.0.MM.

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If one CSS loop is inoperable, 2.

When it is deterLined that one the reactor may remain in core spray loop is inoperable, operation for a period not to at a time when operability is exceed 7 days providing required, the other core spray all active components in loop and the RHRS (LPCI mode) the other CSS toop tad the shall be demonstrated to be RHR system (LPCI mode)

OPERABLE immediately. The and the diesel generators OPERABLE core spray loop shall are OPERABLE.

be demonstrated to be OPERABLE daily thereafter.

l 3.

If Specification 3.5.A.1 or Specification 3.5.A.2 cannot be met, the reactor shall be in the COLD SHUTDOWN CONDITION vithin 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

4.

When the reactor

.ssel pressure is atmospheric and irradiated fuel is in the reactor vessel at least one I

core spray loop with one OPERABLE pump and associated l

diesel generator shall be OPERABLE, except with the I

reactor vessel head removed as specified in 3.5.A.5 or PRIOR TO STARTUP as l

specified in 3.5.A.1.

l BPN 3.5/4.5-2 Anendment ho. 153, 159 l

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3.5/4.5 CORE AND CONTAINMINT COOLING SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVIILLANCE REQUIREMENTS i

3.5.A Core Sorav System (CSS) 5.

When irradiated fuel is in the reactor vessel and the reactor vessel head is removed, core spray is not required provided work is not in progress which has the potential to drain the vessel, provided the fuel pool gates are open and the fuel pool is maintained above the lov I

level alarm point, and provided one RHR$W pump and associated valves I

supplying the standby l

l coolant supply are OPERABLE.

1 ETN 3.5/4.5-3 Unit 1

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3.5/4.$

CORE AND CONTA?NMINT C00L2NG SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.B Residual Heat Removal System 4.5.B. Residual Heat Removal Sys123 I

(RHRS) (LPCI and Containment (RHRS) (LPCI and Containment Cooling)

Cooling) 1.

The RNRS shall be OPERABLE:

1. a.

Simulated once/

Automatic Operating (1) PRIOR TO STARTUP from Actuation Cycle a COLD CONDITION, or Test (2) when there is b.

Pump Opera-Per Specifi-irradiated fuel in j

bility cation 1.0 f04 the reactor vessel i

I and when the reactor vessel pressure is c.

Motor Opera-Per Specifi-greater than ted valve cation 1.0.MM atmospheric, except as operability specified in Specifications 3.5.B.2,,

d.

Pump T1ov Once/3 months through 3.5.B.7.

Rate f

e.

Test Check Per Specifi-I Valve cation 1.0.MM I

Each LPCt pump shall deliver i

9000 spm against an indicated l

system pressure of 125 psig.

l Two LPCI pumps in the same t

loop shall deliver 12000 spm against an indicated system pressure of 250 psig.

2.

With the reactor vessel 2.

An air test on the dryvell pressure less than 105 psis, and torus headers and nozzles the RHRS may be removed shall be conducted once/5 from service (except that tve years. A water test may be RHR pumps-containment cooling performed on the torus header mode and associated heat in lieu of the air test.

exchangers must remain OPERABLE) for a period not to exceed 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months while being drained of suppression chamber quality water and filled with primary coolant quality water, provided that during cooldown two loops with one pump per loop or one loop l

vith two pumps, and associated diesel get.erators, in the core spray system are OPERABLE.

ETN 3.5/4.5-4

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3.5/A.5 CORE AND CONTAINMINT COOLING SYSTEMS LIMITING CONDITIONS FOR OTLRATION SURVEILLANCE REQUIREMINTS i

3.5.3 Residual Heat Removal System 4.5.B Residual Heat Removal System (RHRS) (LPCI and Containment (RMRS) (LPCI and Containment Cooling)

Cooling)

8. If Specifications 3.5.B.1 8.

No additional surveillance through 3.5.B.7 are not met, required.

I an orderly shutdown,shall be initiated and the reactor shall be shutdown and f

placed in the COLD SKUTDOWN CONDITION within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

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When the reactor vessel 9.

When the reactor vessel I

pressure is atmospheric and pressure is atmospheric, irradiated fuel is in the the RHR pumps and valves f

reactor vessel, at least one RHR that are required to be i

loop with two pumps or two loops OPERABLE shall be l

vith one pump per loop shall demonstrated to be OPERABLE f

be OPERABLE. The pumps' per Specification 1.0.MP..

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associated diesel generators must also be OPERABLE.

10. If the conditions of 10.

No additional surveillance i

Specification 3.5.A.5 are met, required.

LPCI and containment cooling are not required.

11. When there is irradiated fuel

11. The RHR pumps on the in the reactor and the reactor adjacent units which supply vessel pressure is greater than cross-connect capability atmospheric, 2 RHR pumps and shall be demonstrated to be associated heat exchangers and OPERABLE monthly when the l

valves on an adjacent unit must cross-connect capability be OPERABLE and capable of is required.

I supplying cross-cortnect l

capability except as specified in l

Specificatien 3.5.B.12 below.

(Note: Because cross-connect capability is not a short-term requirement, a component is not considered inoperable if cross-connect capability can be i

restored to service within 5 l

l hours.)

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3.5/4.5 CORE AND CONTAINMINT COOLING SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIRIMENTS 3,$,5 Residual Heat Removal Svatem 4.5.B Residual Heat Removal System t

(RHRS) (LPCI and Containment (RMRS) (LPCI and Containment Cooling)

Cooling)

12. If one RHR pump or associated

12. When it is determined i

heat exchanger located that one RMR pump or

(

en the unit cross-connection associated heat exchanger in the adjacent unit is located on the unit INOPERABLE for any reason cross-connection in the (including valve inoperability, adjacent unit is f

pipe break, etc.), the reactor INOPERABLE at a time when may remain in operation operability is required, t

for a period not to exceed the remaining RER pump rnd f

30 days provided the remaining cssociated heat exchanger i

RHR pump and associated diesel on the unit cross-connection l

generator are OPERABLE, shall be demonstrated to be OPERABLE immediately and every 15 days thereaf ter until the INOPERABLE pump and associated heat exchanger are returned to normal l

service.

l 13.

If RHR cross-connection flow or 13.

No additional surveillance heat removal capability is loat, required.

the unit may remain in operation l

for a period not to exceed 10 i

davs unless such capability is i

restored.

I 14 All recircolation pump 14 All recirculation pump discharge valves shall discharge valves shall be OPERABLE prior to be tested for operability reactor startup (or during any period of closed if permitted reactor Cold Shutdevn i

elsewhere in these exceeding 48 hcurs, if I

l specifications).

operability tests have i

i not been perforred during the preceding i

31 days.

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o 6 1.5/4.5 CORE AND CONTAINMENT COOLING SYSTEMS LIMITING CONDIT!0NT, Fnt OPERATION SURVEILLANCE REQUIREMENTS 3.5.C RHR Service Water and Emeraency 4.5.C RHR Service Water and Emereenev Ecuirment Coolina Water Systems 7,au t ement Coolina Water Systeta (IECVS)

(EECVS) 1.

PRIOR TO STARTUP from 1.

a.

Each of the RHR$W pumps a COLD CONDITION, 9 RHRSW normally assigned to pumps must be OPERABLE, with automatic service on 7 pumps (including pump D1 the EECW headers will or D2) assigned to RHR$W be tested service and 2 automatically automatically each time starting pumps assigned to the diesel generators EECW service, are tested. Each of the RRR8W pumps and all associated essential control valves for the EECW headers and RHR heat i

exchanger headers ahall be demonstrated to be OPERABLE in accordance with l

Specification 1.0.MM.

l b.

Annually each RHRSW pump shall be flow-rate tested. To

[

be considered OPERABLE, each pump shall pump at least l

4500 spa through its normally a 11gned flow path.

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" sndment No. 159 i

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H 3.5/t. 5 CORE AND C0!rrAINMENT C00LINC JYSTEMS LIMITINO CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.C RHR Service Water and_Emeraenev 4.5.C EER Service Water and Emerzency Ecuirment Coolint Water Syste~s Ecuirment Coolint Water Syste?s (EECWS) (Centinued)

(EECVS) (Continueil 2.

During reactor power

2. a.

If no more than two operation, RHR$W pu ps RHR$W pumps are must be OPERABLE and INOPERABLE, increased assigned to service as surveillance is not indicated in Table 3.5-1 required.

for the specified time

limits, b.

When three RHR$W pumps are IMOPERABLE, the remaining pu=ps and associated essential

{

control valves shall be operated daily.

c.

When four RHR$W pumps are INOPERABLE, the l

remaining pumps and associated essential I

control valves shall be l

operated daily.

1 1

3.

During power operation,

3. Routine surveillance for both RHR$W pu=ps D1 and these pumps is specified l

D2 normally or alternately in 4.5.C.1.

I assigned to the RHR heat exchanger header supplying I

the standby coolant supply connection cust be l

OPERABLE except as i

specified in 3.5.C.4 I

and 3.5.C.5 belev.

i u

l

.5/~.5-;;

r: tr t :, ', :

'J n i t

6 0

1,5/4.5 CORE AND CONTAINMENT COOLING SYSTEMS I

L LIMITING CONDITIONS FOR OPERATION SURVIILLARCE REQUIRIMENTS 3,$,c RMR Service Water and Emertenev 4.5.C RHR Service Water and Emertenev l

Eggj ement Coolint Water Systems Eauirment Coolina Water Systems l

{EECVS) (Continued)

(EECWS) (Continued) l 4.

One of the D1 or D2 RHRSW

4. When it la determined that I

pumps assigned to the RHR one of the RRR5V pumps heat exchanger supplying supplying standby coolant j

the standby coolant supply is IN0FERABLE at a time t

connection may be when operability is t

INOPERABLE for a period required, the OPERABLE l

not to exceed 30 days RHR$W pump on the same j

provided the OPERABLE pump header and the RNR heat is aligned to supply the exchanger header and l

RHR heat exchanger beader associated essential control f

and the associated diesel valves shall be demonstrated generator and essential to be OPERABLE immediately l

and every 15 days l

thereafter.

5.

The standby coolant supply capability may be INOPERABLE l

for a period not to exceed l

I 10 days.

I 6.

If Specifications 3.5.C.2 through 3.5.C.5 are not met, an orderly shutdown shall be initiated and the unit placed in the Cold Shutdown condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

[

7.

There shall be at least L

2 RHRSW pumps, associated with the selected RHR pumps, aligned for RER heat exchanger service for each reacter vessel

[

containing irradiated fuel.

j i

l i

l r

5F::

3.5/4.5-12

+c:~ent

'0.

153

'Cn i t 1

,s t

i 3.5/4.5 CORE AND CONTAINMENT COOLING SYSTEMS l

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.D Eaulement Area Coolers 4.5.D Eculement Area Coolers 1.

The equipment area cooler

1. Each equipment area cooler

[

associated with each RHR is operated in conjunction pump and the equipment with the equipment served l

area cooler associated by that particular coolers i

with each set of core therefore, the equipment

(

spray pumps (A and C area coolers are tested at l

or B and D) must be the same frequency as the OPERABLE at all times pamps which they serve.

I I

when the pump or pumps served by that specific j

cooler is considered to t

be CPERABLE.

I 2.

When an equipment area l

cooler is not OPERABLE, l

the pump (s) served by that cooler must be considered inoperable for technical l

specification purposes.

E.

Hinh Pressure Coolant Iniection j E. Himh Pressure Coolant System (HPCIS)

Iniection System (HPCIS) t 1.

The HPCI system shall be 1.

HPCI Subsystem testing f

CPEiABLE:

shall be performed as i

follows 6

a.

Simulated once/

(1) PRIOR TO STARTUP free a COLD CONDITION, or l

Automatic operating Actuation cycle Test

[

t e

l I

b.

Pump Per

[

(2) whenever there i, i

irradiated fuel in the Opera-Specification f

reactor vessel and the bility

1. 0. m i

reactor vessel pressure issteaterthan122psig,l, c.

Motor Oper-Per i

except as specified in attd Valve Specification l

Specification 3.5.E.2.

Operability

1. 0. m d.

Tiov Rate at once/3 t

normel months l

reactor vessel operating l

pressure i

EPN 3.5/4.5-13 bcn hent *o.

159 Unit 1 L

t f

l i

I

i 3.5/4.5 CORE AND CONTAINMEffr C00 LINO SYSTEMS I

LIw!!!NO CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.E Minh Pressure Coolant Iniection 4.5.E Hiah Pressure Coolant Iniection L

System (HPCIS)

Svatem (HPCIS)

I 4.5.E.1 (Cont'd) e.

Tiow Rate at once/

150 pais operating cycle The NPCI pump shall deliver at least 5000 spa during each flow rate test.

2.

If the HPCI system is

2. When it is determined that

[

inoperable, the reactor may the NPCIS is inoperable, the remain in operation for a ADS actuation logic, the i

period not to exceed 7 daya, RCICS, the RHR$ (LPCI), and provided the ADS, CSS, RNRS the CSS shall be (LPCI), and RCICS are demonstrated to be OPIRABLE OPERABLE.

immediately. The RCICS and ADS logic shall brir denonstrated to be OPERABLE daily thereafter.

3.

If Specifications 3.5.E.1 I

or 3.S.E.2 are net met, an orderly shutdown shall be initiated and the i

reacter vessel pressure shall be reduted to 122 psis or Issa within 24 l

hours.

T.

Reacter Core isolatien Coolina T.

Reacter Core Iselation Coolint System (RCICS)

SystegL[RCICS) t 1.

The RCICS shall be OPERABLE:

1. Rf!C Subsystem testing shall be performed as follows:

I (1) FRIOR 70 STARTUP from a COLD CONDITION, or

a. Simulated Auto-Once/

i matic Actuation operating (2) whenever there is Test cycle irradiated fuel in the

[

reactor vessel and the

b. Pump Per Specif!-

t l

reactor vessel pressure Operability catien 1.0.tM is above 122 psis, except as specified in i

c. Motor-Operated Fer 3.5.T.2.

l Valve Specificatien Cretability 1.0.PM i

1 BPN 3.5/4.5-14 A enPert ',o.159 t

Unit 1 l

l l

i i

e.

I 3,S/4.S CORE AND CONTAINMENT COOLING SYSTEMS l

LIMITING CON t", IONS FOR OPERATION SURVEILLANCE REQUIREMENTS

(

3.5.7 Reactor Core Isolation Coolina 4.5.T Reactor Core Isolation Coolina System (RCICS)

System (RCICS)

(

4.5.F.1 (Cont'd) d.

Tiov Rate at once/3 normal reactor months vessel operating j

pressure l

e.

Flow Rate at once/

150 pais operatint cycle i

l The RCIC pump shall deliver at least 600 spm during each flev test.

2.

If the RCICS is INOPERABLE,

2. When it is determined that the reactor may remain in the RCICS is IN0FERABLE, the l

operation for a period not HPCIS shall be demonstrated to exceed 7 days if the to be OPERABLE immediately.

HPCIS is CPERABLE during

[

such time.

3.

If Specifications 3.5.T.1 f

or 3.5.F.2 are not met, an orderly shutdown shall be initiated and the reactor

[

shall be depressurized to i

less than 122 psig within

[

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

G.

Automatie Deeressurization G. Auta== tie Deereasurization

[

System (ADS)

System (ADS) 1.

Tour of the six valves of 1.

During each operating j

the Automatic cycle the following i

Depressurization Systen tests shall be performed j

shall be OPERABLE:

on the ADS:

l s

l (1) prior to a STARTUP a.

A simulated autematic i

from a Cold Condition, actuation test shall

[

or, be perfereed prior to l

STARTUP after each f

t

!T::

3. 5 / t.. ! -; 5 Unit i

l t

t

i e,

j y

3.5.M. Refereneet 1.

"Tusi Densification Effects ora General Electric Boiling Water Reactor Tuel," Supplements A, 7, and 8, NEIM-10735, August 1973.

2.

Supplement 1 to Technical Report on Densification of General Electric Reactor Tuels, December 14, 1974 (USA Regulatory Staff).

3.

Cennunicati u V. A. Moore to I. S. Mitchell, "Modified GE Model for Tuel Densification," Docket 50-321, March 27, 1974 4.

Generic Reload Tuel Application, Licensing Topical Report, NIDE-24011-P-A and Addenda.

5.

Letter from R. H. Buchholz (GE) to P. S. Check (NRC), "Response to KRC Request Tor Information on CDYN Computer Model," September 5, 1980.

t I

r i

l l

I

" n: v t 'a.

1-t 3r::

3,3n.3 32 Unit 1

[

j I

r 1

I I

e.

4,)

kV1E Shu WVHkSAAMCHk W W w A a sm W.JhLmJ wwA. L A A A gia w g A AtwwgugAga The testing interval for the core and containment cooling systems is based on industry practice, quantitative reliability analysis, judgment, and practicality. The core cooling systems have not been designed to be fully testable during operation. For example, in the case of the HPCI, automatie initiation during power operation vould result in pumping cold water into the reactor vessel which is not desirable. Complete ADS testing during power operation causes an undesirable loss-of-coolant inventory. To increase the availability of the core and containment cooling system, the components which make up the system, i.e.,

instrumentation, pumps, valves, etc., are tested frequently. The pumps and motor operated injection valvis are also tested in accordance with Specification 1.0.MM to assure their operability. A simulated autcmatic l

actuation test once each cycle combined with testing of pumps and inje: tion valves in accordance with Specification 1.0.MM is deemed to be l

adequate testing of these systems.

When ecmponents and subsystems are out-of-service, overall core and containment cooling reliability is maintained by demonstrating the operability of the remaining equipment. The degrad of operability to be demenstrated depends on the nature of the reason for the out-of-service equipment. For routine out-of-service periods caused by preventiva.

maintenance, etc., the pump and valve operability checks will be performed to demonstrate operability of the remaining components.

However, if a failure, design dettetency, cause the outage, then the demonstration of operability should be thorough enough to assure that a

generic problem does not exist. For example, if an out-of-service period was caused by failure of a pu=p to deliver rated capacity due to a design deficiency, the other pumps of this type might be subjected te a flow rate test in addition to the operability cheeks.

Whenever a CSCS system or loop is made inoperable because of a required test et calibration, the other CSCS systems or loops that art required to i

be operable shall be censidered operable if tt y are within the required surveillance testing frequency and there is no reason to suspect they are inoperable. If the function, system, or loop under test, or calibration is found inoperable or exceeds the trip level setting, the LCC and the required surveillance testing for the system or loop shall apply.

Redundant operable components are subjected to increased testing during equiteent out-of-service times. This adds further censervatism and increases assurance that adequate cooling is available should the need arise.

Maximum Averste Planar LHGR. LHOR. and MOP 2 The KAPLHGR, LHGR, and MCPR shall be checked daily to d<termine if f'tel burnup or control rod movement has caused changes in power distribution.

Since changes due to burnup are slov, and only a few control rods are moved daily, a daily check of power distribution is adequate.

ETN 3.5/4.5-35

g r: en.

.,7, 15g Unit 1 m.

a' j

e f

3.6/4.6 PRIMARY SYSTEM BOUNDARY LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.6.C.

coelant Leakare 4.6.C.

Coolant ' eakane 2

1. a.

Any time irradiated 1.

Reactor coolant fuel is in the systes leakage shall reactor vessel and be checked by the l

reactor coolant sump and air sampling temperature is above system and recorded 212'T, reactor coolant at least once per i

leakage into the 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

l prirary containment from unidentified sources i

shall not exceed r

5 spm. In addition, i

the total reactor i

coolant system I

leakake into the primary containment I

shall not exceed 25 spm.

I

(

b.

Anytime the reactor is in j

RUN mode, reactor coolant l

1eakage into the primary containment from i

unidentified sources i

shall not inareese by i

more than 2 spa averaged l

over any 24-hour period

(

l

)

in which the reactor j

is in the RUN mode

[

l except as defined in j

3.6.C.1.c below.

c.

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

in the RUN mode following i

STARTUP, an increase in t

reactor coelant leakage into the primary

[

containment of >2 spm i

i is acceptable as long as the requirements i

of 3.6.C.1.a are met.

i I

I i

I ETN 3.6/4.6-9 A-^"

~E t '; 0 137 Unit 1

l e

l

~

1 6/4.6 PRIMARY SYSTEM BOUNDARY l

LIMITING CONDITIONS T0R LPIRATION SURVEILLANCE REQUIP.EMENTS i

e 3.6.C coolant Leakatt 4.6.C CoolatiLkeakate 2.

Both the sump and air sampling

2. With the air sampling systems shall be OPERABLE during systes inoperable, grab I

REACTOR POWER CPTRATION. Trea samples shall be obtained and after the datt that one of and analyzed at least once

[

these systems is made or found every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

1 l

to be.fnoperable for any rsason.

[

j REACTJK POVER OPERATION is l

perr.issible only during the succeeding 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for the sunp system or 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months for the air i

sarpling system.

I i

l The air sampling system may i

l be removed from service for a period of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for l

calibration, function testing, 1

and maintenance without I

providing a temporary

[

menttor.

l I

3.

If the condition in 1 or 2 above cannot be met, an orderly shutdown shall be initiated l

j and the reactor shall be i

i shutdown in the COLD SNUTDOWN t

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

l t

3.6.D.

Relief Valves 4.6.D.

Relief Valves i

1.

Approximately one-half of i

1.

When more than one relief valves t are known to be failed, an all relief valves shall I

orderly shutdown shall be l

be bench-checked or initiated and the reactor repinced with a depressurized to legs than 105 bench-checked valve l

paig within 24 houra.

each operating cycle.

J All 13 valves vill have been checked or replaced i

I

'f uprn the completten of l

every second cycle.

l l

2.

In accordance with l

Specification 1.0.MM.

I l

each relief valve shall be I

manually opened until i

l thermocouples and l

acoustic monitors I

I downstream of the valve f

indicats steam is flowing from the valve.

j t

f ETN 3.6/4.6-10

+.q e q.g, g,

,9 LT.i t 1

{

l l

,,o a.v.w,,.w.w sw.u. -,

reasonably in a matter of a few hours utilizing the available leakage detection schemes, and if the origin cannot be determined in a reasonably short time, the unit should be shut down to allow further investigation and corrective action.

The two spa limit for coolant leakage ratr increase over any 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period is a limit specified by NRC (Reference 2). This limit applies only during the RUN mo6e to avoid being penalized for the expected coolant leakage increase during pressurization.

The total leakage rate consists of all leakage, identified and unidentified, which flows to the dryvell floor drain and equipment drain sumps.

The capacity of the dryvell floor sump pump is 50 spa and the capacity of the dryvell equipment sump pump is also 50 spm. Removal of 25 spm from either of these sumps can be accomplished with considerable margin.

ItETIFET, 1.

Nuclear System Leakage Rate Limits (BTNP TSAR Subsection 4.10) 2.

Safety Evaluation Report (SER) en IE Bulletin 82-03 3.6.D/4.6.D Pelief Valvej To meet the safety basis, 13 relief valves have been installed on the unit with a total capacity of 34.1 percent of nuclear botier rated steam flov at a refeeence pressure of (1,105 4 1 percent) pais. The analysis of the vorst ooerpressure transient, (3-second closure of all sain steam line isolation valves) net.lecting the direct scram (valve position scram) results in a maximum vessel pressure which, if a neutron flux scram is assumed considering 12 valves operable, results in adequate margir. to the code allevable overpressure limit of 1,375 psig.

To meet operational design, the analysis of the plant isolation transient (generator load reject with bypass valve failure to open) shows that 12 of the 13 relief valves limit peak system pressure to a value which te vell below the allowed vessel overpressure of 1,375 pais.

Experience in relief valve operation shevs that a testing of 50 percent of the valves ver year is adequate to detect fatiu.*es or deteriorations.

The relief valves are benchtested every second operating cycle to ensure that their setpoints are within the 1 i percent tolerance. The relief valves are tested in place in accordance with Specification 1.0.MM to I

establish that they vill open and pass steam.

i 1

ETN 3.6/4.6-30

sans eng.,o, 159 i

Unit 1 i

i

a o a.o.u,..v.u s w w..e we The requirements established above apply when the nuclear system can be pressurized above ambient conditiens. These requirements are applicable at nuclear system pressures below normal operating pressures because abnoreal operational transients could possibly start at these conditions such that eventual overpressure relief would be needed. However, these transients are much less severe, in terms of pressure, than those starting at rated conditions. The valves need not be functional when the vessel head is removed, since the nuclear system cannot be preesurized.

EIrrRINCES 1.

Nuclear System Pressure Relief System (BTNP TSAR Subsectirn 4.4) 2.

Amendment 22 in response to AEC Question 4.2 of December 6, 1971.

3.

"Protection Against overpressure" (ASMI Boiler and Pressure Vessel Code. Sectien ?!I, Article 9) 4 Brevns Terry Nuclear Plant Design Deficiency Report--Target Rock Safety-Lelief Valves, transmatted by J. E. Gilliland to T. E. Kruesi, August 29, 1973 5.

Generic Reload Tuel Application, Licensing Topical Report, NIDE-24011-P-A and Addenda 3.6.E/4.6.E Jet Pures l

Tailure of a jet pump nozzle 4:sembly holddown mechanism, nozzle assembly and/or riser, vould increase the cross-sectional flow area for blevdown following the design basis double-ended line break. Also, failure of the diffuser vould eliminate the capability to reflood the core to two-thirds height level following a recirculation line break. Therefore, if a failure cecurred, repairs must be made, j

The detection technique is as fo11 ova. With the two recirculatten pumps balanced in speed to within 2 5 percent, the flow rates in both recirculation loops will be verified by control room monitoring l

1 instruments.

If the two flow rate values do not dif f er by more than 10 percent, riser and nozzle assembly integrity has been verified.

r If they do differ by 10 percent r more, the core flow rate measured by J

the jet pump diffuser differential pressure system must be checked 1

against the core flow rate derived from the measured values of loop flov i

to core flow correlation.

If the difference between measured and derived I

)

core flow rate is 10 percent or more (with the derived value higher) i 1

diffuser crasurements vill ba taken to define the locatien within the 4

vessel of failed jet pump nozzle (or riser) and the unit shut dovn for repairs.

If the potential blevdown flow area is increased, the system l

J

!TN 3.t/4.e-31

. :.i t 1 i

l l

i 3.7/4.7 C0fffAINMENT SYSTEMS LIMITINC CONDITIONS TOR OPERATION SUPVEILLANCE RIQUIREMI;!rIS 4.7.A.

Prir.ary Centainment 4.7.A.2. (Cont'd)

J.

Continuous Leak Rate Monitor I

When the primary containment is inerted the containset.t shall be continuously monitored i

for gross leakage by L

review of the inerting system sakeup t

requirements. This monitoring systes may be taken out of service for paintenance but shall be returned to service as l

soon as practicable.

(

t K.

Drvve11 and Terus

(

Surfaces l

The interior surfaces of i

the dryvell and torus above the level one foot l

below the normal vater i

line and outside j

surfaces of the torus t

below the water line l

shall be visually i

inspected each operating I

cycle for deterioration f

and any signs of

(

structural damage with

[

particular attention to piping connections and supports and for signs l

l of distress or

}

I displacement.

l l

l

\\

I 1

f

[

tru

3. 7 n.7-9 l

1 l'ni t 1

e.

r 3.7/4.7 C0ffTAINME!TT SYSTEM!

r l

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.7.A PRIMARY C0!T*Af?M ?rf 4.7. A ?RIMARY CONTAf ?M?rf 3.

Pressure Suerrassion Chamber -

3. Pressure Sueeression Char.ber-Reacter Buildina Vacuum Breakers Remeter Buildina Vacuum Breakers I

a. Except as specified in

a. The pressure suppression 3.7.A.3.b below, two pressure chamber-reactor building i

suppression chamber-reactor vacuum breakers shall be

(

building vacuum breakers shall exercised in accordance with be OPERABLE at all times when Specification 1.0. Pet, and the l

i primary centatraent integrity associated instrumentation is required. The setpoint includina tetPoint shall be of the differential pressure functional.ly tested for proper instrumentation which actuates operstien each three months.

l the pressure suppression chamber-reseter building l

vacuum breakers shall be 7

j 0.5 paid.

l 1

1

b. From and af ter the date

b. A visual examination and j

that one of the pressure determination that the i

i suppression chamber-reactor force required to open each

[

building vacuum breakers is vacuum breake- (check valve) l made or found to be inoperable does not exceed 0.5 paid for any reason, reactor vill be made each refueling operation is permissible only

outage, i

during the succeeding seven days, provided that the I

repair procedure does not j

i violate primary contairaent I

integrity.

4 D.Iyve11-Pressure Sumerenaien l

4. Drvvell-Pressure _sumeression i

Chavber Vacuum Breakers Chamber Vacuum Bt.tahtra i

I 1

a. When prirary containment is

a. Each dryvell-suppression

{

required, all dryve11 chamber vacuum breaker j

suppression chamber vacuum j

shall be tested in accordance I

breakers shall be OPERAILi I

with Specification 1.0.MM.

I and positioned in the fu'

{

l closed positten (excep' J

3 j

b. When it is determined that j

during testing) excep, s

specified in 3.7.A.4.b and tvo vacuum breakers are 1

3.7.A.4.c., below.

inoperable for opening at a 1

tire when crerability is

b. One dryvell-suppressicn required, all other vacuum

(

1 1

chameer vacuum breaker may breaker valves shall be

[

be nenfully closed so long exercised imediately and t

as it is detereined to be not every 15 days thereaf ter until

(

4 core than 3' open as indicated the inoperable valve has been f

j by the resitien itshts, returned to normal service.

i 3

I

)

l

! T."

3.7/4.7-10

e e e n.,, ; g, g ;

j

)

1l nit 1 i

I

3. n..,

w ainimuni sis - s LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.7.A Primary centainment 4.7.A Primary containment 3.7.A.4 (Cont'd) 4.7.A.4 (Cont'd)

c. Two dryvell-suppression

c. Each vacuum breaker valve chamber vacuum breakers shall be inspected for may be determined to be proper operation of the inoperable fir opening.

vulve and limit switches in accordance with Specification 1.0.fm.

I

d. If Specifications 3.7. A.4.a.

d. A leak test of the dryvell 3.7.A.4.b. or 3.7.A.4.c.

to supp ession chamber cannot be zet, the structure shall be conducted unit shall be placed in a during each operating cycle.

COLD SHUTDOWN CONDITION in Acceptable leak rate is I

0.09 lb/see of primary an orderly canner within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

containment att.osphere with j

1 pst differential.

5.

orvaan ceneentratien

5. Orvaan ceneantratien a.

Containment atmosphere shall be

a. The primary containment I

reduced to less than 4% oxygen oxygen concentration shall with nitrogen gas during reactor I

be measured and recorded daily. The oxygen power operation with reactor coolant pressure above 100 pais, measurement shall be adjusted I

except as specified in 3.7.t.5.b.

to account for the uncertainty of the method used by adding a predetermined error function, b.

Within the 24-hour period I

b. The methods ised to measure the primary eintainment subsequent to placing the reactor,

in the RUN MODE following a shut-oxygen concer 4 ration shall down, the containment atmosphere be calibrated once every oxygen concentration shall be refuelina cycle, reduced to less than 4% by volume and maintained in this condition.

Deinerting zey consence 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to a shutdown.

t If plant control air is being used ;

c. The control air supply valve c.

to supply the pneumatic control for the pneumatic control system inside primary containment, system inside the primary the reactor shall not be started,

" ntainment shall be verified or if at power, the reactor shall ssed prior to reactor startu;-

be brought to a COLD $NUIDO*=3 and monthly thereafter.

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

d.

If Specificaticn 3.7.A.5.a and 3.7.A.S.b cannot te ret, an i

orderly shutdevn shall be l

initiated and the reactor shall te in a COLD $MUTOC'3 CO* 0!!ICS within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

!T::

3.7/4.7 11 2 faP+et '.0.

'. ! '. '. 5 3 Unit 1

3.7/4.7 t'0!CAf fMI?C SYSTEMS LIMITING CONDIT!CNS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.7.A.

Pri-ary Centsin-ent 4.7.A.

Primary Containment 6.

Dryvell-Sueeression Chamber

6. prvvell-Suenression Chamber pirrerential PreatuI1 Differentini Pressure

a. Differential pressure between

a. The pressure differential the dryvell and suppression between the dryvell and chamber shall be maintained suppression chamber shall at equal to or greater than be recorded at least once 1.1 paid except as specified each shift.

in (1) and (1) below (1) This differential shal' te estat11shed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of achieving eterating terperature and pressure. The differential pressure may be reduced to less than 1.1 paid 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to a scheduled shutd6vn.

(2) This differential may l

be decreased to less than 1.1 paid for a etximus of fcur hours during required l

operability testing of the HPCI system, RCIC system and the dryvell-pressure suppressicn chtsber vacuum breakers,

b. If the differential pressure of Specification 3.7.A.6.a cannot be maintained and the differential pressure cannot be restored within the subsequent six-hour peried, an orderly shutdown shall be initiated and the reactor shall be in the COLD SWJTDOWN CONDIT!0N vithin 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

l l

i ETN 3.7/4.'-12 Unit 1 I

3,7/4.7 Colf?iTImRNT SYSTEM 3 L:M: TIN 3 CONDITIONS FOR OPERATION SURVIILLANCE REQUIREMENTS 3.7.C, seeendary Centainment 4 If refueling tone secondary containment cannot be maintained, the following conditions shall be met

s. Handling of spent fuel and all operatsons over spent fuel pools and open reactor l

wells containing fuel shall be prohibited.

b. The standby gas treatment

[

system suction to the refueling zone will be I

blocked except for a I

controlled leakage area t

sized to assure the i

achievins of a vacuum of at least 1/4-inch ot' vater i

i and not over 3 inches of l

l water in all three reactor i

tones.

1 l

D.

Primary Containment Isolation Valves D. Primary Containment feelatten Valves i

l l

1. When primary containment

1. The primary containment l

integrity is required, all isolation valves isolation valves listed in Table surveillance shall be i

3.7.A and all reactor coolant performed as follows:

l system instrument line flow check i

valves shall be OFIRABLE except

a. At least once per

{

as specified in 3.7.D.2.

operatina cycle, the OPERABLE isolation valves l

l that are power operated and j

l automatically initiated shall be tested for simulated automatic l

initiation, and in i

accordance with i

Specification 1.0.MM.

I i

tested for closure times.

t ETN 3.7/4.7-17 A ecd ent No. 1 5. 159 l'ni t 1 l

e 3.9/4.9 CONTAINMENT SYSTEMS LIMITINO CONDITION 3 TOR OTERATION SURVIILLANCE REQUIRIMENTS I

3.7.D. Primary contain. tent Isolation 4.7.D.

Primary Containment 7 solation ValYet Valvis 4.7.D.1 (Cont'd)

b. In accordance with Specification 1.0.MM, all normally open power operated isolation valves shall be functionally tested, i

c. (Deleted)

d. At least once per operating I

cycle, the operability of the reactor coolant system instrument line flow chet-valves shall be verified.

2. In the event any isolatten valve

2. Whenever an isoletion valve specified in Table 3.7.A becomes listed in Table 3.7.A is inoperable, reactor power inoperable, the posiNion of at operation may continue provided least one other valve in each at least one valve in each line line having an inoparable valve having an inoperable valve is shall be recorded tally.

i CPERABLE and within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months l

either i

I

a. The inoperable valve is restored to CPERABLE l

l status, or

b. Each affected line is isolated by use of at least l

one deactivated containment i

iselatten valve secured in the isolated position.

3. If Specification 3.7.D.1 and i

3.7.D.2 cannot be met, an i

orderly shutdown shall be initiated and the reactor shall i

be in the COLD SKUTDOWN CONDITION l

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

l E TN 3.7/4.7-18 A~end ent NO. 145. 1*f. 159

[

Unit 1 i

l

),7/4.7 CONTAINMENT SYSTIMS LIMITING CONDITIONS FOR OPERATION SURVE!LLANCE RIQUIREMINTS l

3.7.T. Prim a ry Centain-ent Purte 4.7.T.

Primary Centainment Puttg I

System tyg, tim

1. The primary containment shall

1. At least once every 18 menths, be nores11y vented and purged the pressure drop across the through the primary containment combined HEPA filters and purge system. The standby gas charcoal adsorber banks shall treatment system may be used be demonstrated to be less than l

vhen primary containment purge 4.5 inehts of water at system system is INOPERABLE.

design flev rate (2 10%).

2. a. The results of the in elace

2. a. The tests and sampic cold D0P and halogensted analysis of Specification hydrocarben tests at design 3.7.T.2 shall be performed flows en KEFA filters ans at least once per operating charcoal adsorber banks shall cycle or once every thov 199% DOP removal and 18 months, whichever occurs 199% halogenated hydrocarbon first or after 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months removal when tested in of system operatien and accordance with following significant ANSI N510-1975.

painting, fire, or chemical release in any ventilation I

tone com=unicating with I

the system.

i

b. The results of laboratory

b. Cold DOP testing shall be carbon sample analysis performed af ter each shall show 185% tadioactive complete or psrtial methyl iodide removal when replacement of the HIPA tested in accordance with filter bank er Of ter any ASTM D3803 structural maintenance on (130'C 95% R.H.).

the system housing.

c. System flow rate shall be

c. Halogenated hydrocarbon shevn te te within 21C% ef testing shall be perfereed design flov vhen tested in after each complete or accerdance with ANSI N510 partial replacement of the

1975, charcoal adsorber bank or after any structural esintenance Cn the syseem housing.

[

l ETN 3.7/4.7-21 E~ e n :~ f u t

'4. '.*3.

li:

l'n i t 1

(

.o,

~

J.7/4.7 CONTAf10ENT SYSTEMS LIMITIES CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS

(

3.7.G.

Centainment Attenthere 4.7.G.

Containment Atmesehere l

Dilution Svatan (CAB)

Dilution System (CAD)

(

1.

The Containment Atmosphere 1.

svatan Omerability Dilution (CAD) Systen shall l

be CPERABLE withs i

a.

Two independent a.

Cycle each solenoid systems capable of operated air /nitrosen supplying nitrosen valve through at l

1 to the dryvell and least one complete I

torus, cycle of full travel I

in accordance with I

Specification 1.0.MM..

and at least once per sonth verify that each manual valve in the flow path is cren.

[

i b.

A minimum supply of b.

Verify that the CAD l

2,500 sallons of System contains a liquid nitrosen per minimum supply of system.

2,500 gallons of liquid nitresen twice per week.

2.

The Containment Atmosphere When TCV s4-8B is Dilution (CAD) Systen shall inoperable, each be OPERABLE whenever the solenoid operated

+

reactor is in the RUN MODE.

air / nitrogen valve of Systes 5 shall be l

cycled through at least

[

one complete cycle of full travel and each manual valve in the flow path of Systen B shall be verified l

open at least on:e per

veek, t

t 3.

If one system is inoperable.

l the reactor may remain in j

operation for a period of 30 days provided all active i

components in the other f

system are CTERAILE.

l 6

i ETN 3.7/4.7-22 A e n dme n t *.0.

159 l

t' nit 1

[

l I

s

4 o

3.//4./

DADL3 (bonc*Q) with doors closed and fan in :peration, DOP aerosol shall be sprayed I

externally along the full linear periphery of each respective door to check the gasket seal. Any detection of DOP in the fan exhaust shall be considered an unacce; table test result and the gaskets repaired and test repeated.

If significant painting, fire or chemical release occurs such that the HEPA filter or charcoal adsorbar could beceme contaminated from the fumes, chemicals or foreign material, the same tests and sample analysis shall be performed as required for operational use. The determination of significant I

shall be made by the operator on duty at the time of the incident.

Knowledgeable staff members should be censulted prior to making this l

p determination.

Demonstration of the automatic initiation capability and operability al filter cooling is necessary to assure system performance cspability. If one stanJby l

l gas treatment system is inopera' ele, the other systems must be tested daily.

l This substantiates the availability of the operable systems and thus rear:or I

operation and refueling operation can continue for a limited period of time.

3.7.D/4.7 D Primary Containment Isolation Valves f

Double isolation valves are provided on lines penetrating

.e primary l

containment and open to ti.e free space of the containment. Closure of one of I

the valves in each line would be sufficient to maintain the integrity of the l

pressure suppression system. Automatic initiation is required to tdaimize the l

potential leakage paths from the containment in the event of a LOCA.

l l

Groue 1 - Process lines are isolated by reactor vessel low water li,tel (378")

in order to allow for removal of decay heat subsequent to a scram, yr. isolate in time for proper operatica of the core standby cooling systems. It valves in Group 1, except the reactor water sample line valves, are also c).osec when process instrumentation detects excessive main steam line flow, high radiation, low pressure, or main stets space high temperature. The reactor water sample line valves isolate only on reactor low water level at

S" or main steam line high radiation.

Grouo 2 - Isolation valves are closed by reactor vessel low water level (538")

or high dryvell pressure. The Group 2 isolation signal also "isolates" the reactor building and startr, the standby gas treatment system. It is not desirable to actuate the Group 1 isolation signal by a transient or spurious signal.

l Greue 3 - Process lines are normally in use, and it is therefore not desirable to cause spurious isolation due to high dryvell pressure resulting from i

nonsafet) related causes. To protect the reactor from a possible pipe treak in tne syste=, isolation is provided by high temperature in the cleanup system area or high flov through the inlet to the cleanup systeu. Also, since the vessel could potentially be drained through the cleanup system, a low-level isolation is provided.

BTN 3.7/4.7-49 Unit 1

e l.

J. s / a.. #

M/ tat.kg. Wwas up Groues 4 and 5 - Process lines are designed to remain operable and mitigate the consequences of an accident which results in the isolation of other process lines. The signals which initiate isolation of Groups 4 and 5 prrcess lines are therefore indicative of a condition which would render them inoperable.

Greue 6 - Lines are connected to the primary containment but not directly to the reactor vessel. These valves are isolated on reactor lov vater level (538"), high dryvell pressure, or reactor building ventilation high radiation which vould indicate a possible accident and necessitate primary containment isolation.

Groue 7 - Process lines are closed only on the respective turbir.: strem supply valve not fully closed. This 6ssures that the valves are not open when HPCI or RCIC action is required.

Grour 8 - Line (traveling in-core probe) is isolated on high dryvell pressure or reactor low water level (538"). This is to assure that this line does not provide a leakage path when containment pressure or reactor water level indicates a possible accident condition.

The maximum closure time for the automatic isolation valves of the primary containment and reactor vessel isolation control system havt

en selected in consideration of the design intent to prevent core uncoverint aollowing pipe breaks outside the primary containment and the need to contait. released fission products following pipe breaks inside the primary containment.

In satisfying this design intent, an additional margin has been included in specifying maximum closure.imes.

Thic margin permits identification of degraded valve performance prior to exceeding the design closure times.

In order to assure that the doses that may result from a steam line break do not exceed the 10 CTR 100 guidelines, it is necessary that no fuel rod perforation resulting from the accident occur prior to closure of the main steam line isolation valves. Analyses indicate that fuel rod cladding perforations vould be avoided for main steam valve closure timas, including instrument delay, as long as 10.5 seconds.

These valves are highly reliable, have low service requirements, and most are normally closed. The initiating sensors and associated trip logic are also checked to demonstrate the capability for automatic isolation. The test interval of once per operating cycle for automatic initiation results in a failure probability of 1.1 x 10-' that a line vill not isolate. More I

frequent testing for valve operability in accordance with Specification 1.0.MM results in a greater assurance that the valve vill be operable when needed.

The main steam line isolation valves are functionally tested per Specification 1.0.MM to establish a high degree of reliability.

The primary containment is penetrated by several small diameter instrument lines connected to the reactor coolant system. Each instrument line contains a 0.25-inch restricting orifice inside the primary containment and an excess flow check valve outside the primary containment.

BTN 3.7/4.7-50 A,end ent ';o.159 Unit 1

c-

[p ostmo, UNITED STATES NUCLEAR REGULATORY COMMISSION y*

.,. f) je WASHINGTON, D. C. 20555 3-

. p ENNESSEEVALLEYAUTHORITY DOCKET N0, 50-260 BROWNS FERRY NUCLEAR PLANT, UNIT 2 AMENDMENT TO FACILITY CFERATING LICENSE Amendment No.155 Licens? No. DPR-52 1.

The Nuclear Regulatory Co7eission (the Comission) has founo that:

A.

The application for amendment by Tennessee Valley Authority (the licensee) dated October 27, 1987, cc,rplies with the standards and recuirements of the Atomic Energy Act of 1954, as amended (the Act),

and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of th9 Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the acti'.ities authorized by this arendment can b: conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.

The issuance of this amendment will not be inimical to the comon defense and security or to the bealth and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

2-Accordingly, the license is amended by changes to the Tee'inical 2.

Specifications as indicated in the attachment to this license amendment and paragraph 2.C,(?) of Facility Operating License No. OPR-52 is hereby arended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendices A and R, as revised through Amendrent No.155, are hereby incorporated in the license. The licensee shall operate the facility iri accordance with the Technical Specificatiens.

3.

This license amendment 1s effective as of its date of issuance and shall be irrlemented within 60 days from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION h

%A c

Suzanne Black Assistant Director for Projects TVA Projects Division Office of Special Projects

Attachment:

Changes to the Technical Specifications Date of Issuance:

November 22, 1953

ATTACHMENT TO LICENSE AMENDMENT NO. 155 FACILITY OPERATING LICENSE NO. OPR-52 DOCKET NO. 50-260 Revise the Apoendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages. The revised pages are idantified by the captioned amendment number and contain marginal lines indicating the area of change. Overleaf pages* are provided to maintain document completeness.

REMOVE INSERT 1.0-11 1.0-11*

1.0-1?

1.0-12 3.3/4.3-11 3.3/4.3-11*

3.3/4.3-12 3.3/4.3-12 3.3/4.4-1 3.3/4.4-1 3.3/4.4-2 3.3/4.4-2*

3.5/4.5-1 3.5/4.5-1 3.5/4.5-2 3.5/4.5-2 3.5/4.5-3 3.5/4.5-3*

3.5/4.5-4 3.5/4.5-4 3.5/4.5-7 3.5/4.5-7 3.5/4.5-8 3.5/4.5-8*

3.5/4.5-9 3.5/4.5-9 3.5/4.5-10 3.5/4.5-10*

3.5.4.5-12 3.5/4.5-12*

3.5/4.5-13 3.5/4.5-13 3.5/4.5-14 3.5/4.5-14 3.5/4.5-15 3.5/4.5-15*

3.5/4.5-34 3.5/4.5-32*

3.5/4.5-35 3.5/4.5-33 3.6/4.6-9 3.6/4.6-9*

3.6/4.6-10 3.6/4.6-10 3.6/4.6-30 3.6/4.6-30 3.6/4.6-31 3/6/4.6-31*

3.7/4.7-9 3.7/4.7-9*

3.7/4.7-10 3.7/4.7-10 3.7/4.7-11 3.7/4.7-11 3.7/4.7-12 3.7/4.7-12 3.7.4.7-17 3.7/4.7-17 3.7.4.7-18 3.7/4.7-18 3.7/4.7-21 3.7/4.7-21*

3.7/4.7-22 3.7/4.7-22 3.7/4.7-49 3.7/4.7-49*

3.7/4.7-50 3.7/4.7-50

1.0 DEFINITIONS (Ccnt'd)

MM. Survel11ance Reauirements for ASMI Sectien YI Pume and Valve Proaram - Surveillance requirement lor Inservice Testing of ASME Code Class 1, 2, and 3 components shall be applicable as follows:

1.

Inservice testing of ASMI Code Class 1, 2, and 3 pueps and valves shall be performed in accordance 'with Section XI of the ASMI Boiler and Pressure Vessel Codr. and applicable Addenda as required by 10 CFR 50, Section 50.55a(s), except, where specific written relief has been granted by the Commission pursuant to 10 CFR 50, Section 50.55a(s)(6)(i).

2.

Surveillance intervals specified in Sectica XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda for the inservice testing activities required by the ASME Boiler and Pressure Vessel Code and applicable Addenda shall be applicable as follows in these technical specifications:

ASME Boiler and Pressure Vessel Required frequencies Code and applicable Addenda for performing inservice terminology for inservice testina activities testint activitiec Weekly At least once per 7 days Monthly At least once per 31 days Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days Yearly or annually At least once per 366 days 3.

The provisions of Specification 1.0.LL are applicable to the above required frequencies for performing inservice testing activities.

4 Performance of the above inservice testing activities shall be in addition to other specified Surveillance requirements.

5.

Nothing in the ASME Soiler and Pressure Vessel

ode shall be construed to supersede the requirements of any technical specification.

BTN 1.0-11 Amen:,ent f:o. 155 Unit 2

Table 1.1 SURVEILLANCE FREOUENCY NOTATION NOTATION FREQUENCY t

S (Shift)

At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

D (Daily)

At least one.e per normal calendar 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months day (midnight to midnight).

W (Weekly)

At least once per 7 days.

M (Monthly)

At least once per 31 days.

Q (Quarterly)

At least once per 3 months or 92 days, i

SA (Semi-Annually)

At least once per 6 months or 184 days.

Y (Yearly)

At least once per year or 366 days.

R (Refueling)

At least once per operating cycle.

I S/U (Start-Up)

Prior to each reactor startup.

N.A.

Not applicable.

P (Prior)

Completed prior to each release.

i 4

i l

i i

BTN 1.0-12 l

Amendment No. 155 Unit 2 I

3.3/4.3 PEACTIVITY CONTROL LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.3.C.

Scram Insertion Times 4.3.C. Scram Insertion Times 2.

The average of the scram inser-

2. At 16-week latervals, 10%

tion times for the three fastest of the OPERABLE control OPERABLE control rods of all rod drives shall be scram-groups of four control rods in timed above 800 psig, a two-by-two array shall be no Whenever auch scram time greater than:

measurements are made, an evaluation shall be made

% Inserted From Avg. Scram Inser-to provide reasonable Fully Withdrawn tion Times (sec) assurance that proper control rod drive 5

0.398 performance is being 20 0.954 maintained.

50 2.120 90 3.800 a.

The maximum scram insertion time for 90% insertion of any OPERABLE control rod shall not exceed 7.00 seconds.

D.

E1 activity Anomalies D.

Reactivity Anomalies The reactivity equivalent of During the STARTUP test the difference between the program and STARTUP following actual critical rod refueling outages, the configuration and the expected critical rod configurations configuration during power will be compared to the operation shall not exceed 1%

k.

expected configurations at If this limit is exceeded, the selected operating conditions.

reactor will be shut down These comparisons will be until the cause has been used au base data for determined and corrective reactivity monitoring during actions have teen taken as subsequegt power operation appropriate.

throughout the fuel cycle.

At specific power operating conditions, the critical rod configuration vill be ce= pared to the configuration expected based upon appropriately corrected past data. This esmparison vill be made at least every full power month.

l BFN 3.3/4.3-11 Amendment No. 129 Unit 2

3.3/4,3 REACTIVITY CONTROL LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.3.E.

If Specifications 3.3.C and.D 4.3.E.

Surveillance requirements are above cannot be met, an orderly as specified in 4.3.C and.D shutdown shall be initiated and above.

the reactor shall be in the SHUTDOWN CONDITION within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

F.

Scram Discharme Volume (SDV)

F.

Scram Discharme Volume (SDV) l 1.

The scram discharge volume 1.a. The scram discharge drain and vent valves shall volume drain and vent be OPERABLE any time that valves shall be verified the reactor protection open PRIOR TO STARIUP i

system is required to be and monthly thereafter.

i OPERA 5LE except as The valves sp9eified in 3.3.F.2.

say be closed l

intermittently for l

1 l

testing not to exceed 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months in any 24-hour l

period during operation.

l l

1.b. The scram discharge volume drain and vent I

valves shall be I

demonstrated OPERABL2 I

in accordance with Specification 1.0.MM.

l 2.

In the event any SDV drain 2.

When it is determined l

or vent valve becomes that any SDV drain or inoperable, REACTOR POWER vent valve is inoperable, OrERATION may continue the redundant drain or provided the redundant vent valve shall be drain or vent valve is demonstrated OPERABLE OPERABLE.

immediately and weekly thereafter.

3.

If redundant drain or vent 3.

No additional valves become inoperable, surveillance required.

the reactor shall be in HOT STANDBY CONDITION within l

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

BFN 3.3/4.3-12 Amendment No. 129, 155 Unit 2

a 3.4/4.4 STANDBY LIOUID CONTROL SYSTEM LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3,4 STANDPY LIOUID CONTROL SYSTEM 4.4 STANDBY LIOUID CONTPOL SYSTEM Acelicability Acelicability Applies to the Operating status Applies to the surveillance of the Standby Liquid Control requirement 6 of the Standby System.

Liquid Control System.

Obiective obiective To assure the availability of a To verify the operability of the system with the capability to Standby Liquid Control System, shut devn the reactor arc' maintain the shutd:vr. condition without the use of control rods.

Seecification Snecification A.

Normal System Availability A.

Normal System Availability l

1.

Except as specified in The operability of the Standby 3.4.B.1, the Standby Liquid Liquid Control 3ystem shall be Control System si.all be verified by the performance l

OPERABLE at all times when of the following testst I

there is fuel in the reactor vessel and the reactor is not 1.

Verify pump OPERABILITY in a shutdown condition with in accordance with Specification 3.3.A.1 Specification 1.0.MM.

satisfied.

2.

At least once during each operating cycles a.

Check that the setting of the system relief valves is 1,425 1 75 psig.

b.

Manually initiate the system, except explo-I sive valves. Visually verify flow by pumping boron solution through j

the recirculation path and back to the Standby Liquid Control Solution Tank. After pumping boron solution, the system shall be flushed with demineralized water. Verify minimun ETN 3.4/4.4-1 A~end ent N. 12:. 15^.

'I' Unit 2

+

o 3.4/4.4 STANDBY LIOUID CONTROL SYSTEM LIMITING C0!TDITIONS FOR OPERATION SURVEILLANCE REQUIREMEfrIS 4.4.A Normal System Acelicability 4.4.A.2.b. (Cont'd) pump flow rate of 39 sps against a system head of 1275 psis by pumping desineralized water from the Standby Liquid Control Test Tank.

c. Manually initiate one of the Standby Liquid Control System loops and pump demineralized water into the reactor vessel.

This test checks explosion of the charge associated with the tested loop, proper operation of the valves, and pump operability.

Replacement charges shall be selected such that the age of charge in service shall not exceed five years from the manufacturer's assembly date.

d.

Both systems, including both explosive valves, shall be tested la the course of two operating cycles.

I ETN 3.4/4.4-2 A.endmen* '.o. 132, !!O Unit 2 i

)

3.5/4.5 CORE AND CONTAIRMENT COOLING SYSTEMS LIMITING CONDITIONS FOR OPEPATION SURVEILLANCE REQUIREMINTS 3.5 CORE AND CONTAINMINT COOLING 4.5 CORE AND CONTAINMENT COOLING SYSTEMS SYSTEMS i

Aeolicability Aeolicability Applies to the operational Applies to t.he surveillance l

status of the core and requirements of the core and containment cooling sys".-es.

containment cooling systems when the corresponding limiting condi-tion for operation is in effect.

Objective Objective To assure the opesaDi~. ty of To verify the operability of the the core and contain:Jst cooling core and containment cooling systems under all es 4tions for systems under all conditions for which this cooling c.padility is which this cooling capability is an essential response to plant an essantial response to plant abnormalities, abnormalities.

Seecification Soecification A.

Cere Serav System (CSil A.

Core Serav System (CSS) 1.

The CSS shall be OPERABLE:

1.

Core Spray System Testing.

(1) PRIOR TO STARTUP 1113 Frecuenev from a COLD CONDITION, or a.

Simulated Once/

Automatic Operating (2) when there is irradiated Actuation Cycle fuel in the vessel test and when the reactor l

vessel pressure b.

Pump Opera-Per Specifi-l is greater than bility cation 1.0.MM atmospheric pressure, except as specified c.

Motor Per Specifi-in Specification Operated cation 1.0.MM I 3.5.A.2.

Valve Operability d.

System flov Once/3 rates Each months j

loop shall deliver at least 6250 gpm against a system head corres-ponding to a BTN 3.5/4.5-1 Amendment NO. 155 Unit 2

_;u 3,} (A,5 CORE AND CONTAINMENT COOLING SYSTEMS LIMITINi_ CONDITIONS TOR OPERATION SURVEILLANCE REQUIREMINTS 3.5.A Core Soray System (CSS) 4.5.A Core Sorav System (C111 4.5.A.1.d (Cont'd) 105 psi differential pressure between the reactor vessel and the primary containment.

e.

Check Valve ?er Specification 1.0.MM.

2.

If one CSS icop is inoperable, 2.

When it is determined that one the reactor may remain in core spray loop is inoperable, operation for a period not to at a t.de when operability is exceed 7 days providing required, the other core spray all active components in loop and the RHRS (LPCI mode) the other CSS loop and the shall be demonstrated to be RHR system (LPCI mode)

OPERABLE immediately. The and the diesel generators OPERABLE core spray loop shall are OPERABLE.

be damonstrated to be OPERABLE daily thereafter.

3.

If Specification 3.5. A.1 or Specification 3.5.A.2 cannot be met, the reactor shall be in the COLD SHUTDOWN CONVITION within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

4 When the reactor vessel pressure is atmospheric and irradiated fuel is in the reactor vessel at least one core spray loop with one OPERABLE pump and associated diesel generator shall be OPERABLE, except with the reactor vessel head removed as specified in 3.5.A.5 or PRIOR TO STARTUP as specified in 3.5.A.1.

BPN 3.5/4.5-2 Amendment NO. 149. 155 Unit 2 I

i I

d 3.5/4.5 CORE AND CONTAINMENT COOLING SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.A Core Strav System (CSS) i 5.

When irradiated fuel is in the reactor vessel and the reactor vessel head is removed, core spray is not required pgovided work in not in progress which has the potential to drain the vessel, provided t..

fuel pool gates are open and the fuel pool is raintained above the lov level alarm point, and provided one RHRSW pump and associated valves supplying the standby coolant supply are OPERABLE.

i i

i f

+

I i

i ET!!

3.5/4.5-3 Unit 2 I

n-

,,,a

e 3.5/4.5 CORE AND CONTAINMENT COOLING SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMINTS 3.5.B Residual Heat Removal System 4 5.B. Residual Heat Memoval System (RHRS) (LPCI and Containment (RHRS) (LPCI and Containment Cooling)

Cooling) 1.

The RHRS shall be OPERABLE:

1. a.

Simulated Once/

Automatic Operating (1) PRIOR TO STARTUP Actuation Cycle from a COLD CONDITION, or Test (2) when there is b.

Pump Opera-Per irradiated fuel in bility Specification the reactor vessel 1.0.MM and when the reactor vessel pressure is c.

Motor Opera-Per greater than ted valve Specification atmospheric, except as operability 1.0.MM specified in Specifications 3.5.B.2, d.

Pump Tiov Once/3 through 3.5.B.7.

Rate months e.

Testable Per Check Valve Specification 1.0.MM Each LPCI pump shall deliver 9000 sps against an indicated system pressure of 125 psig.

Two LPCI pumps in the same loop shall deliver 12,000 gpm against an indicated system pressure of 250 pais.

2.

With the reactor vessel

2. An air test on the dryvell pressure less than 105 psig, and torus headers and nozzles the RHR may be removed shall be conducted once/5 from service (except that two years. A water test may be RHR pumps-containment cooling performed on the torus header rode and associated heat in lieu of the air test.

exchangers must remain OPERABLE) for a period not to exceed 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months while being drained of suppression chtsber quality water and filled with primary coolant quality water, provided that during cooldown two loops with tae pump per loop or one loop with two pumps, and associated diesel generators, in the core spray system are OPERABLE.

BTN 3.5/4.5-4 Arendment No. 155 Unit 2

3.5/4.5 CORE AND CONTAINMINT COOLING SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.B Residual Heat Removal System 4.5.B.

Residual Heat Removal System (RHRS) (LPCI and Containment (RHRS) (LPCI and Containment Cooling)

Cooling)

8. If Specifications 3.5.B.1 8.

No additional surveillance through 3.5.B.7 are not met, required.

an orderly shutdown shall be initiated and the reactor shall be shutdown and placed in the COLD SKUTDOWN CONDITION within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

9.

When the reactor vessel 9.

When the reactor vessel pressure is atmospheric and pressure is atmospheric, irradiated fuel is in the

.the EHR pumps and valves reactor vessel, at least one RHR that are required to be loop with two pumps or two loops OPERABLE shall be with one pomp per loop shall demonstrated to be OPERABLE be OPERABLE. The pumps' per Specification 1.0.MM.

l associated diesel generators must also be OPERABLE.

10. If the conditions of 10.

E; additional surveillance Specification 3.5.A.5 are met, required.

LPCI and containment cooling are not required.

l

11. When there is irradiated fuel

11. The RHR pumps on the l

in the reactor and the reactor adjacent units which supply vessel pressure is greater than cross-connect capability l

atmospheric, 2 RHR pumps and shall be demonstrated to be r

l associated heat exchangers and OPERABLE monthly when the valves on an adjacent unit must cross-connect capability be OPERABLE and capable of is required.

supplying cross-connect j

capability except as specified in Specification 3.5.B.12 below.

(Note:

Because cross-connect capability is not a short-term requirement, a component is not considered inoperable if cross-connect capability can be restered to service within 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months .)

BFN 3.5/4.5-7 A.menement No. 155 Unit 2 l

l 3.5/4.5 CORE AND CONTAINMEfff COOLING SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.B Residual Heat Removal System 4.5.B.

Residual Heat Removal System (RHRS) (LPCI and Containment (RHRS) (LPCI and Containment Cooling)

Cooling) 12.

If three RHR pumps or associated

12. When it is determined heat exchangers located that three RHR pumps or on the unit cross-connection associated heat exchangers in the adjacent units are located on the unit INOPERABLE for any reason cross-connection in the (including valve inoperability, adjacent units are pipe break, etc.), the reactor INOPERABLE at a time when may remain in operation operability is required, for a period net to exceed the remaining RHR pump and 30 days provided the remaining associated heat exchanger RHR pump and associated diesel on the Unit cross-connection generator are OPERABLE.

shall be demonstrated to be OPERABLE immediately and r

every 15 days thereafter until the INOPERABLE pump and associated heat exchanger are returned to normal service.

13.

If RHR cross-connection flov or

13. No additional surveillance heat removal capability is lost, required.

the unit may remain in operation for a period not to exceed 10

{

days unless such capability is restored.

14 All recirculation pump 14 All recirculation pump discharge valves shall discharge valves shall be OPERABLE prior to be tested for operability reactor startup (or during any period of closed if permitted reactor Cold Shutdown elsewhere in these exceeding 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , if l

specifications).

operability tests have not been performed during the preceding 31 days.

i i

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1 I

ETt 3.5/4.5-8 e n d e n t ';:. 10. '. ? !

Unit 2 l

r J, 3 # a., 3 bk/ A1,e Alt &# 9 V a t A A A s ta a.w a t A V w w AJ A a t Q VAw&unaw LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.C RHR Service Water and Emerzenev 4.5.C RHR Service Water and Eaulement Coolint Water Systems Emeraency Eouiement (EECWS)

Coolina Water Systems (EECWS) 1.

PRIOR TO STARTUP from a 1.

a.

Each of the RHRSW pumps COLD CONDITION, 9 RHR$W normally assigned to pumps must be OPERABLE, with automatic service on 7 pumps (including one of the EECW headers will pumps D1, D2, B2 or 81) be tested assigned to RHRSW service automatically each time and 2 automatically starting the diesel generators pumps assigned to EECW are tested. Each

service, of the RERSW pumps and all associated essential control valves for the EECW headers and RHR heat exchanger headers shall be demonstrated to be OPERABLE in accordance with Specification 1.0.MM.

l b.

Annually each RHRSW pump shall be flow-rate tested. To be considered OPERABLE, each pump shall pump at least 4500 spm through its normally assigned flow path.

l l

BPN 3.5/4.5-9 A9endment fio 155 Unit 2

3.5/4.5 CORE AND CONTAINMENT COOLING SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.C RHR Service Water and Emermancy 4.5.C. RHR Service Water and Eauienant Coolina Water Systems Emeraenev Eauinment Coolina_

(EECWS) (Cont'd)

Water Systems (EECWS) (Cent'd) 2.

During reactor power

2. a.

If no are than two operation, RHRSW pumps RHRSW pumps are must be OPERABLE and INOPERABLE, increastd assigned to service as surveillance is not indicated in Table 3.5-1 required.

for the specified time

limits, b.

When three RHRSW pumps are INOPERABLE, the remaining purps and associated essential control valvc4 shall be opetated daily.

c.

When four RHRSW pumps are INOPERABLE, the remaining pumps and associated essential control valves shall be operated daily.

3.

During unit 2 power

3. Routine surveillance for operation, any two RHRSW these pumps is specified pumps (D1, D2, B1, and B2) in 4.5.C.1.

normally or alternately assigned to the RHR heat exchanger header supplying the standby coolant supply connection must be OPERABLE except as specified in 3.5.C.4 and 3.5.C.5 below.

ETt 3.5/4.5-10 en e n t.,9, }:9 L' nit 2

E 3,$/4.5 CORE AND CONTAINMENT COOLING SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.C RHR Service Water and Emeraency 4.5.C. RHR Service Water and Emuiement Coolina Water Systems Emerzenev Eauiement Coolina (EECWS) (Continued)

Water Systema (EECWS) (Cont'd) 4.

Three of the D1, D2, B1, B2

4. When it is determined that RHR$W pumps assig_ed to three of the RHRSW pumps the RHR heat exchanger supplying standby coolant supplying the standby are INOPERABLE at a time coolant supply connection when operability is may be INOPERABLE for a required, the OPERABLE period not to exceed 30 days RNRSW pump on the same provided the OPERABLE pump header and the RHR heat is sligned te su; ply the exchanger header and RHR heat exchanger header associated essential control and the essociated diesel valves shall be demonstrated generator and essential to be OPERABLE immediately control valves are OPERABLE, and every 15 days thereafter.

5.

The standby coolant supply capability may be INOPERABLE for a period not to exceed 10 days.

6.

If Specifications 3.5.C.2 through 3.5.C.5 are not met, an orderly shutdown shall be initiated and the unit placed in the Cold Shutdovn condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

7.

There shall be at least 2 RHRSW pumps, associated with the selected LHR pumps, aligned for RHR heat exchanger service for each reactor vessel containing irradiated fuel.

l l

l 2.s 5-12 gnl.

< e,3 e,1.;0. 1 2 1

3.5/4.5 CORE AND CONTAINMINT COOLING SYSTEMS

_ LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIRLENTS 3.5.D.

Eauletent Area Coolers 4.5.D. Eaulement Area Coolert 1

1.

The equipment area cooler

1. Each equipment area cooler i

associated with each RER is operated in conjunction pump and the equipment with the equipment served tsrea cooler associated by that particular cooler; with each set of core therefore, the equipment I

spray pumps (A and C area coolers are tested at or 8 and D) must be the same frequency as the OPERABLE at all times pumps which they serve.

when the pump or pu=pa e

served by that specific cooler is considered to te CTERABLE.

2.

When an equipment area y

cooler is not OPERABLE, the pump (s) served by that L

cooler must be considered I

inoperable for technical specification purposes.

E.

Hith Pressure Coolant Insection E. Hith Pressure Coolant System (HPCIS)

Iniection System (HPCIS)

\\

1.

The HPCI system shall be 1.

HPCI Subsystem testing i

OPERABLE:

shall be performed as i

fcilows:

}

(1)

PRIOR TO STARTUP from a a.

Simulated Once/

COLD CONDITION, or Automatic operating-Actuation cycle I

Test (2) whenever there is b.

Pump Per l

irradiated fuel in the Opera-9pecification reactor vessel and the bility 1.0.MM f

reactor vessel pressure i

c.

Motor Oper-Per l

is greater than 122 psig, except as specified in ated Valve Specification Specification 3.5.E.2.

Operability 1.0.MM d.

Flow Rate at onee/3 I

no rmal months reactor t

vessel operating pressure 4

BTN 3.5/4.5-13 A end ent No.155 i

Unit 2 l

I l

4 l

l

3.5/4.5 CORE AND CONTAINMENT COOLING SYSTEMS LIMITIhG CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.E Hiih Pressure Coolant Iniection 4.5.E. Hizh Pressure Coolant System (HPCIS)

In_tection Svatem (MPCIS) 4.5.E.1 (Cont'd) e.

Flow Rate at once/

150 pais operating cycle The NPCI pump shall deliver at least 5000 spm during each flow rate test.

2.

If the HPCI system is

2. When it is determined that inoperable, the reactor may the HPCIS is inoperable, the remain in operation for a ADS actuation logie, the period not ti exceed 7 days, RCICS, the RHRS (LPCI), and provided the ADS, CSS, RHRS the CSS shall be (LPCI), and RCICS are demonstrated to be OPEEABLE OPERABLE.

immediately. The RCICS and ADS logic shall be demonstrated to be OPERABLE daily thereafter.

3.

If Specifications 3.5.E.1 or 3.5.E.2 are not met, an orderly shutdown shall be initiated and he reactor vessel sure shall be reduct to 122 psis or less vi.hin 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

F.

Reactor Core Is?lation Coolina F.

Reactor Core Isolation Coolina System (RCICl)

System (R11Cil 1.

The RCICS shall be OPERABLE

1. RCIC Subsystem testing shall be performed as follovst (1) PRIOR TO STARTUP from a COLD CONDITION, or

a. Simulated Auto-Once/

matic Actuation operating (2) whenever there is Test cycle irradiated fuel in the reactor vessel and the

b. Pump Per Specifi-reactor vessel prassure Operability cation 1.0.MM is abcVe 122 psig, except as specified in

c. Motor-0perated Per Specifi-3.5.P.2.

Valve cation 1.0.K51 Operability BPN 3.5/4.5-14 A end ent No.

155 Unit 2

l l

3.5/4.5 CORE AND CONTATNMENT COOLING SYSTEh5 LIMITING CCNDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS l

3.5.T.

Reactor Core Isolation Coolina 4.5.F. Reactor Core Isolation Coolina l

System (RCICSS System (RCICS) 4.5.F.1 (Cont'd) d.

Flow Rate at once/3 l

normal reactor months vessel operating pressure l

e.

Flow Rate at once/

i 150 psis operating l

cycle l

l l

The RCIC pump shall deliver at least 600 gpm during each flow test.

2.

If the RCICS is INOPERABLE,

2. When it is determined that the reactor may remain in the RCICS is INOPERABLE, the operation for a period not HPCIS shall be demonstrated

(

to exceed 7 days if the to be OPERABLE immediately.

HPCIS is OPERABLE during such time.

l l

3.

If Specifications 3.5.T.1 l

or 3.5.F.2 are not met, an l

orderly shutdown shall be l

initiated and the reactor i

shall be deptesaurized to f

less than 122 psis within i

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

G.

Automatic Deeressurization G. Automatie Deoressurizatian System (AQ11 System (AD l 1. Tour of the six valves of 1. During each operating the Automatic cycle the following Depressurization System tests shall be performed l shall be OPERABLE: on the ADS l (1) prior to a STARIUP

s.. A simulated automatic l

from a Cold Condition, actuation test shall or, be performed prior to STARTUP af ter each i l l l I EPN 3.5/4.5-15 L'ni t 2 i l l I i

F 3.5 BASES (cont'c) i 3.5.M. References i 1. Loss-of-Coolant Accident Analysis for Browns Ferry Nuclear Plant i Unit 2, NEDO - 24088-1 and Addenda. 2. "BWR Transient Analysis Model Utilizing the RETRAN Program," TVA-TR81-01-A. 3. Getaric Reload Fuel Application, Licensing Topical Report, NEDE - 24011-P-A an.1 Addenda. b i L I L f i r P L 1 P l, I l i i f t I l i l l

eaf-*nt

',0. '. 4 7 ( FTN .t. 5 / 4. 5 - 3 2 Unit 2 i P

.,ge .~.......m.... The testing interval for tae core and containment cooling systems is based on industry practice, quantitative reliability analysis, judgment, and practicality. The core cooling systems have not been designed to be fully testable during operation. For example, in the case of the HPCI, autoratic initiation during power operation would result in pumping cold water into the reactor vessel which is net desirable. Complete ADS testing during power operation causas an undesirable loss-of-coolant inventory. To increase the availability of the care and containment cooling system, the components which make up the system, i.e., instrumentation, pumps, valves, etc., are tested frequently. The pumps and motor operated injection valves are also tested in accordance with l Specification 1.0.MM to assure their operability. A simulated automatic l actuation test once each cycle combined with testing of pumps and injection valves in accordance with Specification 1.0.MM is deemed to be l l adequate testing of these systems. When compenents and subsystems are out-of-service, overall core and containment cooling reliability is maintained by demonstrating the operability of the remaining equipment. The degree of operability to be dem:nstrated depends on the nature of the reason for the out-of-service I equipment. For routine out-of-service perioda caused by preventive maintenance, etc., the pump and valve operability checks will be performed to demonstrate operability of the remalains components. However, if a failure, design deficiency, cause the outage, then the [ demonstration of operability should be thorough enough to assure that a generic problem does not exist. For example, if an out-of-service period vas caused by failure of a pump to deliver rated capacity due to a design deficiene", the other pumps of this type might be subjected to a flov rate test in addition to the operability checks. b Whenever a CSCS system or loop is made inoperable because of a required test or calibration, the other CSCS systems or loops that are required to be operable shall be considered operable if they are within the required surveillance testing frequency and there is no reason to suspect they are inoperable. If the function, system, or loop under test or calibration is found inoperable or exceeds the trip level setting, the LCO and the i required surveillance testing for the system or loop whall apply, f Redundant operable components are subjected to increased testing during equipment out-of-service times. This adds further conservatism and increases assurance that adequate cooling is available should t's need arise. L UAximum Averate Planar LHGR. LHGR. and MCPR The MAPLHGR, LHGR, and MCPR shall be checked daily to determine if fuel I i burnup or control rod movement has caused changes in power distribution. j Since changes due to burnup are slov, and only a few control rods are coved daily, a daily check of power distribution is adequate, j i { BTN 3.5/4.5-33 A end ent No. 155 Unit 2 [ i L i l i f

( 2s /4.6 PRIMARY SYSTEM BOUNDARY l 6 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.6.C. Coolant Leakaae 4.6.C. Coolant Leakane

1. a.

Any time irradiated 1. Reactor coolant [ fuel is in the systen leakage shall reactor vessel and be checked by the reactor coolant sump and air sampling temperature is above system and recorded l 212'T, reactor coolant at least once per leakage into the 4 hours. Primary centainment 6 from unid ': d.fied sources I shall not at: d 5 spm. In af61 tion, i the total reactor coolant system j leakage into the primary containment f sha*1 not exceed 25 spm. { b. Anytime the reactor is in RUN mode, reactor coolant leakage into the primary j containment from [ unidentified sources shall not increase by more than 2 spa averaged i over any 24-hour period j in which the reactor i is in the RUN mode except as defined in 3.6.C.1.c below. I c. During the first 24 hours l in the RUN mode following STARTUP, an increase in reactor coolant leakage. j into the primary containment of >2 spm j is acceptable as j long as the requirements } of 3.6.C.1.a are met. t l L ETN 3.6/4.6-9 benPent M5 133 Unit 2 i I f I l J

4 3.6/4.6 PRIMARY SYSTEM B0UNDARY LIMITING CONDITIONS FOR OPERATION sui'VEILLANCE REQUIREMENTS 3.6.C Coolant Leakane 4.6.C Coolant Leakane 2. Both the sump and air sampling

2. With the air sampling systems shall be OPERABLE during system inoperable, grab REACTOR POWER OPERATION. From samples shall be obtained and after the date that one of and analyzed at least once these systems is made or found every 24 hours.

to be inoperable for any reason, i reactor power operation is per'issible only during the succeeding 24 hours for the sump system or 72 hours for the air sampling system. The air sampling system may be removed from service for a period of 4 hours for calibration, function testing, and maintenance without providing a temporary monitor. 3. If the condition in 1 or 2 i above cannot be met, an orderly I shut'.avn shall be initiated and the reactor shall be l shutdown in the COLD SHUTDOWN CONDITION within 24 hours. 3.6.D. Relief Valves 4.6.D. Relief Valves 1. When more than one relief valves 1. Approximately one-half of are known to be failed, an all relief valves shall orderly shutdovu shall be be banch-checked or initiated and the reactor replaced with a depressurized to less than 105 bench-checked valve psig within 24 hours. each operating cycle. All 13 valves will have been checked or replaced j upon the completion of every second cycle. 2. In accordance with Specification 1.0.MM, l each relief valve shall be ( l manually opened until s thersocouples and acoustic monitors i downstream of the valve indicate steam is flowing from the valve. j i i BrN 3.6/4.6-10 A end ent ':0, 133. 155 l Unit 2

,,i* s.om.o umm 3.6.C/4.6.C (Cont'd) reasonably in a matter of a few hours uti)izins the available leakage detection schemes, and if the origin cannot be determined in a reasonably short time, the unit should be shut down to allow further investigation and corrective action. The 2 spa limit for coolant leakage rate increases over an/ 24-hour period is a limit specified by NRC (Reference 2). This limit applies only during the RUN mode to avoid being penalized for the expected coolant leakage increase during pressurization. The total leakage rate consists of all leakage, identified and unidentified, which flows to the dryvell floor drain and equipment drain sumps. The espacity of the dryvell floor sump pump is 50 spm and the capacity of the dryvell equipment sump pump is also 50 spm. Removal of 25 spm from either of these sumps can be accomplished with considerable margin. REFERENCE 1. Nuclear System Leakage Rate Limits (BTNP TSAR Subsection 4.10) 2. Safety Evaluation Report (SER) on IE Bulletin 82-03 3.6.D/4.6.D Relief Valves To meet the safety basis,13 relief valves have been installed on the unit with a total capacity of 84.1 percent of nuclear boller rated step 7 flow. The analysis of the worst overpressure transient, (3-second closure of all main steam line isolation valves) neglecting the direct scram (valve position scram) results in a maximum vessel pressure which, if a neutron flux scram is assumed considering 12 valves OPERABLE, results in adequate margin to the code allowable overpressure limit of 1,375 psig. To meet operational design, the analysis of the plant isolation transient (generator load reject with bypass valve failure to open) shows that 12 of the 13 relief valves limit peak system pressure to a value which is well below the allowed vessel overpressure of 1,375 psis. Experience in relief valve operation shova that a tett!ng of 50 percent of the valves per yect is adequate to detect failures or deteriorations. The relief valves are benchtested every second operating cycle to ensure that their setpoints are within the i 1 percent tolerance. The relief valves are tested in place in accordance with Specification 1.0.MM to I establish that they will open and pass steam. BTN 3.6/4.6-30 A*end ent ?;o.155 Unit 2 l

o.or*.o vasos 3.6.D/4.6.D (Cont'd) The requirements established above apply when the nuclear system can be pressurized above ambient conditions. These requirements are applicable at nucicar system pressures below normal operating pressures because abnormal operational transf ants could possibly start at these conditions such that eventual overpressure relief would be needed. However, these transients are much less severe, in terms of pressure, than those starting at rated cond$?. ions. The valves need net be functional when the vessel head is removed, since the nuclear system cannot be pressurized. girERENCES 1. Nuclear System Pressure Relief Systes (BFNP FSAR Subst ' ion 4.4) 2. Amendment 22 in response to AEC Question 4.2 of December 6,1971. 3. "Protection Against Overpressure" (ASMI Boiler and Pressure Vessel Code, Section III, Article 9) 4 Browns Ferry Nuclear Plant Design Deficiency Report--Target Rock g Safety-Relief Valves, transmitted by J. E. Gi?. eland to F. E. Kruesi, August 29, 1973 5. Generic Reload Fuel Application, Licensing Topical Report, !GDE-24011-P-A and Addenda 3. 6. 7. / 4. 6. E Jet Pumes F<tlure of a jet pu=p nozzle assembly holduovn mechanism, nozzle assembly and/or riser, would increase the cross-sectional flow area for blevdown follcwing the design basis Cc-

a-ended line break. Also, failure of the diffuser veuld eliminate the capavility to reflood the core to two-thirds height level following a recirculation line break. Therefore, if a l

failure occurred, repairs must be made. l l The detection technique is as follows. With the two recirculation pumps balanced in speed to within t 5 percent, the flow rates in both l recirculatien icops vill be verified by control room monitoring instruments. If the two flow rate values do not differ by more than 10 percent, riser and nozzle assembly integrity had been verified. If they do dif fer by 10 percent e - sre, the core flow rate measured by the jet pump diffuser differential pressure system must be checked Isainst the core flow rate derived frem the measured values of loop flow to core flow correlation. If the difference between measured and derived core flow rate is 10 percent or more (with the derived value higher) diffuser measurements will be taken to define the location within the vessel of failed jwt pump nozzle (or riser) and the unit shut down for repairs. If the potential blevdown flow area is increas6d, the systen YTN 3.6/4.6-31 Unit 2

3.7/4.7 CONTAINMENT SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMINTS 4.7.A. Primary Centainment 4.7.A.2. (Cont'd)

j. Continuous leak Rate Monito;

.n . When the primary containment is inerted i the containmer.t shall be continuously monitored for gross leakage by review of the inerting system makeup requirements. This sonitoring system may be taken out of service for maintenance but shall be returned to service as soon as practicable. I k. Drvve11 and Torus Surfaces The interior surfaces of the drywell and tcrus [ above the level one foot below the normal veter [ line and outside surfaces of the torus I below the water line i shall be visually L inspected each operating cycle for deterioration and any signs of structural damage with ps ticular attentirn to l piping connections and supports and for signs of distress or displacement, i l i l l EfN 3.7/4.7-9 Unit 2 b I

$._ OUY S E UI A OA$hYU$ bn OkO h LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.7.A Primary Contai4Etal 4.7.A Primary Containment 3. Pressure Sueeression Chamber -

3. Pressure Sueeression Chamber-l Reactor Buildina Vacuum Breakers Reactor Buildina Vacuum Breakers I

a. Except as specified in

a. The pressure suppression e two pressure chamber-reactor building 3.7.A.3.b belov supprossion chamber-reactor vacuum breakers shall be building vacuum breakers shall exercised in accordance with be OPERABLE at all times when Specification 1.0.MM, and the l

primary containment integrity associated instrumentation is required. The setpoint including setpoint shall be of the differential preseure functionally tested for proper instrumentation which actuates operation each three months. the pressure suppression chamber-reactor building vacuum breakers shall be 0.5 psid. 1

b. T:

and after the date

b. A visual examination and r

one of the pressure determination that the i ression chamber-reactor force required to open each . ding vacuum breakers is vacuum breaker (check valve) i .e or found to be inoperab e does not exceed 0.5 psid I l for any reason, reactor will be made each refueling l operation is permissible only outage. during the succeeding seven j days, provided that the repair procedure does not l i violate prinary cor.lainment integrity. t e l 4 Drvve11-Prsasure 3ueeression

4. Dryvell-Pressure Sueeression i

Chamber Vacuum Breakers Chamber Vacuum areakers

a. When primary containment is

a. Each dryvell-suppression f

required, all dryvell-chamber vacuum breaker l uuppression chamter vacuum shall be tested in l breakers shall be OPERABLE accordance with l and positioned in the fully Specification 1.0.MM. l closed position (exetit duritig testing) excene av

b. When it is determined that specified in 3.7.A.A.D and two vacuum breakers are i

3.7.A.4.c., below. inoperable for opening at a l tima when operability is J

b. One dryvell-suppression required, all other vacuum chamber vacuum breaker may breaker valves shall be be nonfully closed so long exereired immediately nnd as it is determined to be not every 15 days thereaf ter until more than 3' open as indicated the inoperaDie valve has been by the position lights, returned to normal service.

BTN 3.7/4.7-10 A end ent No. 133. l'5 Unit 2 4 l I

4 3.7/4.7 CONTAINMINT SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.7.A Primary Containment 4.7.A Primary Containment 3.7.A 4 (Cont'd) 4.7.A.4 (Cont'd)

c. Tso dryvell-suppression

c. Each vacuum breaker valve chamber vacuum breakers shall be inspected for may be determined to be proper operation of the inoperable for opening, valve and limit switches

) in accordance with ) Specificatien 1.0.MM. l l

d. If Specifications 3.7.A.4.a,

d. A leak test of the dryvell

.b, or.c cannot be met, the to suppression chamber unit shall be placed in a structure shall be conducted l COLD SHUTDOWN CONDITION in during each operating cycle. I an orderly manner within Acceptable leak rate is 24 hours. 0.09 lb/see of primary containment atmosphere with 1 psi differential. 5. Orvren Ceneentration

5. Orvsen Coneantration a.

Containment atmosphere shall be

a. The primary containment reduced to less than 4% oxygen oxygen concentration shall with nitro 5en gas during reactor be measured and recorded power operation with reactor daily. The oxygen l

coolant pressure above 100/psis, measurement shall be adjusted except as specified in 3.7.A.5.b. to account for the uncertainty of the method used by adding a predetermined error function. I b. Within the 24-hour period

b. The methods used to measure subsequent to placing the reactor the primary ce4.tainment in the RUN MODE folloving a shut-oxygen concet.tration shall i

down, the containmeat atmosphere be calibrated once every I oxygen concentration shall be refuelir.g cycle. reduced to less than 4% by volume and maintained in this condition. Deinerting may commence 24 hours prior to a shutdown. c. If plant control air is being used

c. The control air supply valve to supply the pneumatic control for the pneumatic control system inside primar/ containment.,

system inside the primary the ret: tor shall not be started, containment shall be verified or if 4t power, the reactor shall closed prior to reactor startup be broutht to a COLD SKUTDOWN and monthly thereafter. CONDITION sithin 24 hourc. d. If Specification 3.7.A.5.a and 3.7.A.5.b car.not be met, an orderly shutdown shall be initiated and the reactor shall be I in a COLD SHUTDOWN CONDITION l vithin 24 hours. BFN 3.7/4.7-11 A~ e d ~ s n * '0. 13~. 155 Unit 2

p 3.7/4.7 COTTAINMENT SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREME!CS 3.7.A. Primary Containment 4.7.A. Primary Containment 6. Dryvell-Suoerassion Chamber

6. Drvve11-Sueoression Chamber Differential Pressure Differential Pressure

a. Differential pressure between

a. The pressure differential the dryvell and suppression between the dryvell and f

chamber shall be maintained suppression chamber shall at equal to or greater than be recorded at least once i 1.1 paid except as specified each shift. l in (1) and (2) below: (1) This differential shall be established within 24 hours of achieving operati-. temperature and pres ore. The differential pressure may be reduced to less than 1.1 paid 24 hours prior to a scheduled shutdown. I (2) This differential may 1 be decreased to less than 1.1 psid for a maximum of four hours during required operability testing of the HPCI system, RCIC system and the dryvell-pressure suppression i

hamber vacuum breakers.

b. If the differential pressure of Specification 3.7.A.S.a i

cannot be maintained and the differential pressure cannot be restored within the subsequent six-hour period, I an orderly shutdown shall be l initiated and the reactor shall be in the COLD SKUTDOWN j CONDITION within 24 hours. i BTN 3.7/4.7-12 Unit 2 l t l l 1 i

3.7/4.7 CONTAIMMINT SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.7.C. Secondary containment

4. If refueling zone secondary containment cannot be maintained, the following conditions shall be mett

a. Handling of spent fuel and all operations over spent fuel pools and open reactor wells containing fuel shall be prohibited.

b. The standby gas treatment system suction to the refueling zone vill be blocked except for a controlled leakage area sized to assure the achievina of a vacuus of at least 1/4-inch of water and not over 3 inches of water in all three reactor zones.

D. Primary Centainment Isolation Valves D. Primary Containment Isolation Valves

1. When primary containment

1. The prima;y containment integrity is required, all isolation valves isolation valves listed in surveillance shall be Table 3.7.A and all reactor performed as follows:

coolant system instrument line flow check valves shall be

a. At least once per OPERABLE except as specified operating cycle, the l

in 3.7.D 2. OPERAELE isolation valves that are power operated and automatically initiated shall be tested for simulated e.utomatic initiation, and in accordance with Specification 1.0.MM, l tested for closure times. BTN 3.7/4.7-17 Amend-ent a, 141, 155 Unit 2 ]

3.7/4.7 CONTAINMINT SY$TEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE RZqUIREMENTS 3.7.D. Erima ry Containment Isolation 4.7.D. Primary Containment Isolation Valves Valves i 4.7.D.1 (Cont'd)

b. In accordance with I

Specification 1.0.MM, all normally open power operated isolation valves shall be l functionally tested.

c. (Deleted)

d. At least once per operating cycle, the operability of the p

reactor coolant system instrument line flow check valves shall be verified.

2. In the event any isolation valve

2. Whenever an isolation valve specified in Tabic 3.7.A becomes listed in Table 3.7.A is inoperable, reactor power inoperable, the position of at i

operation may continue provided least one other valve in each l at least one valve in each line line having an inoperable valve having an inoperable valve is shall be recordeo daily. OPERABLE and within 4 hours either:

a. The inoperable valve is restored to OPERABLE status, or

b. Each affected line is isolated by une of at least one deactivated containment isolation valve secured in the 1

isolated position. i

3. If Specification 3.7.D.1 and 3.7.D.2 cannot be met, an orderly shutdown shall be initiated and the reactor shall l

b2 in the COLD SNUTDOWN CONDITIOM within 24 hours. l ET:s 3.7/4.7-18 A end ent '.a. 141. 142. lif Unit 2

3.5/4.5 CORE AND CONTAINMENT COOLING SYSTEMS LIMITING CONDITIONS FOR OPd ATI!M SURVEILLANCE REQUIREMENTS 3.5 COPE AND CONTAffe'ENT COOLING 4.5 CORE AND CONTAINMENT COOLING SYSTEMS EYSTEMS Acelicability Applicahility Applies to the operational Applies to the surveillance status of the Core and requirements of the Core and Containment Coolins Systems. Containment Cooling Systems when the corresponding limiting condi-tion for operation is in effect. Obiective Obinctive To assure the operability of To verify the operability of the the Core and containment Coolins Core and Containment Cooling Systems under all conditions for Systems under all conditions fcr which tats cooling capability is which this cooling capability is an essential response to plant an essential response to plant abno rmali ties, abnormalities. Seecification Seecification A. Core ferav System (CSS) A. Core Serav System (CSS) 1. The CSS shall be OPERABLE: 1. Core Spray System Testing. (1) PRIOR TO STARTUP ILim Ergavency from a COLD CONDITION, or a. Simulated Once/ Automatic Operating (2) when there is irradiated Actuation Cycle fuel in the vessel test and when the reactor vessel pressure b. Pump Opera-Per Specif1-is greater than bility cation 1.0 MM atmospheric pressure, except as specified c. Motor Per Specifi-in Specification 1 Operated cation 1.0.MM l l 3.5.A.2. Valve Operability } d. System flow Once/3 i ratet Each months l loop shall deliver at l 1 east 6250 spm against a system l head corres-pending to a DFN 3.5/4.5-1 Unit 3 Amendment NO. 130 I t

3.5/4.5 CORE AND CONTAINMENT C00 LINO SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.A Core Sorav Svatem (CSS) 4.5.A Core Serav System (CSS) 4.5.A.1.d (Cont'd) 105 psi differential pressure between the reactor vessel and the primary containment. e. Testable Per Check Valve Specificatien 1.0.MM. 2. If one C35 loep is ineperable, 2. When it is determined that ene the reactor may remain in core spray loop is inoperable, operation for a period not to at a time when operability is exceed 7 days providing required, the other core spray all active components in loop and the RHRS (LPCI mode) the other CSS loop and the shall be demonstrated to be RHR system (LPCI mode) OPERABLE immediately. The and the diesel generators OPERABLE core spray loop shall are OPERABLE. be demonstrated to be OPERABLE daily thereafter. 3. If Specification 3.5.A.1 or Specification 3.5.A.2 cannot be met, the reactor shall be in the COLD SHUTDOWN CONDITION within 24 hovrs. 4. When the reactor vessel pressure is atmospheric and irradiated fuel is in the reactor vessel at least one core spray loop with one OPERABLE pump and associated diesel generator shall be OPERABLE, except with the reactor vessel head recoved as specified in 3.5.A.5 or PRIOR TO STARTUP as specified in 3.5.A.1. BrN 3.5/4.5-2 A,mendment No. 124, 130 Unit 3 l

3.5/4.5 CORE AND CONTAINMINT COOLING SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIRIMENTS 3.5.A Core Sorav System (CSS) 5. When irradiated fuel is in the reactor vessel and the reactor vessel head is removed, core spray is not required provided work is not in prugtess which has the potential to drain the vessel, provided the fuel pool gates are open and the fuel pool is maintained above the lov level alarm point, and provided one RHR$W pump and asrociated . 1 vet su; plying the standby coolant supply are OPERABLE. l 1 l ETN Ur.it 3 3.5/4.5-3

.o, 3.5/4.5 CORE AND CONTAlhMLNr LUULING brS11M3 LIMITING CONDITICMS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.B Residual Heat Removal System 4.5.B. Residual Heat Removai System (FHFS) (LPCI and Containment (RMRS) (LPCI and Containment Cooling) Cooling) 1. The RhRS shall be OPERABLE:

1. a.

Simulated Once/ Autonistic Operating (1) PRIOR TO STARTUP Actuation Cycle from a COLD CONDITION, Test or (2) when there is b. Pump Opera-Per Specif1-irradiated fuel in bility tion 1.0.MM the reactor vessel j and when the reactor vessel pressure is e. Motor Opeta-Per Specifi-greater than j ted valve cation 1.0.MM atmospheric, except as operabilit,' specified in l Specifications 3.5 B.2, d. Pump T1ov Once/3 through 3.5.B.7. Rate monthe i i e. Testable Per Check Specification Valve 1.0.MM Each LPCI pump shall i deliver 9000 sps against i I an indicated system j pressure of 125 pais. Two LPCI pumps in the same l loop shall deliver 12000 L l spa slainst an indicated l syste.. pressure of 250 l psis. r 2. With the reactor vessel

2. An air test en the dryvell pressure less than 105 pais, and torus headers and nozzles the RHR may be removed shall be conducted once/5 i

l from service (except that two years. A vater test may be RHR pu: ps-containment cooling performed on the torus header mode and associated hest in lieu of the air test. t exchangers must remain OPERABLE) for a period not to l exceed 24 hours while being t drained of suppression chamber f quality water and filled with primary coolant quality water, provided that during cooldown ( two loops with one pump per [ lt or one loop vith two l PLe l, and associated diesel a I l generators, in the core spray t system are OPERABLE. ETN 3,54.5-4

- W e r t * ;, ID Unit 3 P

l

e .o* 3,5/4.5 CORE AND CONTAINMENT C00 LINO SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMINTS 3.t.B Fesidual Heat Removal System 4.5.B Residual Heat Re-eval Systt; (RHRS) (LPCI and Containment (RHRS) (LPCI and Containment Cooling)(Cont'd) Cooling)(Cont'd)

8. If Specifications 3.5.B.1

8. No additional surveillance through 3.5.B.7 are not met, required.

an orderly shutdown shall be initiated and the reactor shall be shutdown and placed in the COLD SHUTDOWN Condition within 24 hours.

9. When the reactor vessel

9. When the reactor vessel pressure is atmospheric and pressure is atmospheric, irradiated fuel is in the the RHR pumps and valves reactor vessel, at least one RHR that are required to be 1cck with two pumps or two loops OPERABLE shall be with one pump per loop shall demonstrated to be OPERABLE be OPERABLE. The pumps' l

per Specification 1.0.MM. l associated diesel generator: eust also be OPERABLE. i 10. If the cor.ditions of l

10. No additional surveillance Specification 3.5.A.5 are met, required.

LPCI and containment cocling are not required. l 11. When there is irradiated fuel

11. The B and D RHR pumps on in the reactor and the reaccer unit 2 which supply ves'sel pressure le greater than cross-connect capability atmospheric, unit 2 RHR pu=ps B shall be demonstrated and D vith associated heat ex-to be OPERABLE monthly when changers and valves must be the cross-connect capability OPERABLE and capable of f

is required, supplying cross-connect i capability except as specified l in Specification 3.5.B.12 below. (Note: Because cross-connect capability is not a short-term requirement, a aceponent is not considered inoperable if cross-connect capability can be restored to service within 5 hours.) l i i BFN 3.5/4.5-7

end ent ho, 130 Unit 3

3.5/4.5 CORE AND CONTAINMINT COOLING SYSTEMS LIMITIFG CONDITIONS TOR 0?ERATION SURVEILLANCE REQUIRIMENT> 3.5.E Residual Heat Removal Svatem 4.5.B Residual Heat Re) oval System (RHRS) (LPCI and Containment (RMRS) (LPCI and Containment Cooling)(Cont'd) Cooling)(Cont'd) 12. If one RHR pump or associated

12. When it is determined heat exchanger located that one RHR pump or en the unit cross-connection associated heat exchanser in unit 2 is INOPERABLE located on the unit cross-connection in the for any reason (including a

valve inoperability, I adjacent unit is pipe break, etc.), the reactor INOPERABLE at a time when may remain in operation operability is required, for a period not to exceed the remaining RER rump './ l 30 days provided the remaining associated heat exchange: i RHR pump and associated diesel I on the unit cross-connection generator are OPERABLE. shall be demonstrated to be OPERABLE immediately and every 15 days thereaf ter until the INOPERABLE pump and associated heat exchanger are returned to normal t service. 13. If RHR cross-connection flow or

13. No additional surveillance heat removal capability is lost, required.

f the unit may remain in operation for a period not to exceed 10 days unless such capability is [ i restored. 14 All recirculation pump

14. All recirculation pu=p discharge valves shall discharge valves shall i

f be OPERABLE prior to l be tested for operability l reactor startup (or during any period of j closed if permitted reactor Cold Shutdown j elsewhere in these exceeding 48 hours, if specifications). operability tests have J not been performed during the preceding i 31 days. I i i I i I i i i 4 ETN L' nit 3 3.5/4.5-8 A snd ent No. 124 i r i

3.5/4.5 CORE AND CONTAINMENT COOLING SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3,s.C RHP Service Water and Emertenev 4.5.C RMR Service Water and Ecuirment Coolina Water Systems Emermancy Eaulement Coolint' _EECWS) Water Svatems (EECWS) (

1. PRIOR TO STARTUP from a 1.

a. Each of the RHR$W pumps COLD CONDITION, 9 RER$W normally assigned to pumps must be OPERABLE, with automatic service on 7 pumps (including pump B1 the EECW headers will or B2) assigned to RHR$W be tested service and 2 automatically automatically each time starting pumps assigned to the diesel generators EECW service. are tested. Each of the RHISW pumps and all associated essential control valves for the EECW headers and RHR heat exchanger headers shall te demonstrated to be OPERABLE in accordance with Specification 1.0.MM. b. Annually each RHRSW pump shall be flow-rate tested. To be considered OP RABLE, each pump shall pump at least 4500 spm through its normally assigned flow path. l EPN 3.5/4.5-9 end ent No. 130 Unit 3

3,5/4.5 CORE AND CONTAINMENT COOLING SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.0 FHR Service Water and Emerrency 4.5.C RHR Service Water and Eculement Coolint Water Systems Emeraency Eaulement Coolina (EECVS) (Cont'd) Water Systems (EECWS} (Cont'd)

2. During reactor power 2 :.

5. If ns sere than two operation, RHRSW pu=ps RHRSW paaps mie must be OPERABLE and INOPERABLE, increased assigned to service as surveillance is not indicated in Table 3.5-1 required. for the specified time

limits, b.

When three RHRSW pumps are INOPERABLE, the remaining pumps and associated essential control valves shall be o;erated weekly. c. When four RHRSW pumps are INOPERABLE, the remaining pumps and associated essential control valves, shall be operated daily.

3. During power operation, 3.

Routine surveillance for both Rhd!W pu=ps B1 and these pu=ps is specified B2 normally or alternately in 4.5.C.1. assigned to the RHR heat exchanger header supplying l the standby coolant supply connection must be OPERAELE; except as specified in 3.5.C 4 and 3.5.C.5 below. l l 1 1 1 i l l l l l 3.5/4.5-10 en : n *.,;. ';4 i I

9 3,5/4.5 CORE AND CONTAIN'ENT COOLING SYSTEMS LIMITING CONDITION 3 FOR OPERATION SURVEILLANCE REQUIRIMENTS 3.5.C RHR Service Water and Emertenev 4.5.C RHR service Water and Equierent Coolint Water Systers Emertenev Eauitment Coolint (EECWS) (Cont'd) Water.11 stems (EECWS) (Cent'd)

4. One of the B1 or B2 RHRSW 4.

When it is determined that pumps assigned to the RHR the B1 or B2 RHR$W pump heat exchanger supplying is INOPERABLE at a time the standby coolant supply when operability is connection may be required, the OPERABLE INOPERABLE for a period RHR$W pump on the same not to exceed 30 days header and the RER heat provided the OPERABLE pump exchanger header and ir aligned to supply the associated essential centrol RHR heat exchanger header valves shall be demonstrated and the assectated diesel to be OPERABLE immediately generator and essential and every 15 days thereaf ter control valves are OPERABLE.

5. The standby coelant supply espability may be IN0PERABLE for a period not to exceed 10 days.

6. If Specifications 3.5.C.2 through 3.5.C.5 are not cet, an orderly shutdown shall be initiated and the unit placed in the Cold Shutdown condition within 24 hours.

7. There shall be at least 2 RHR$W pumps, associated with the selected RHR pumps, aligned for RHR heat exchanger service for each reactor vessel centaining irradiated fuel.

I I 4 ? = ': 'Jni t 3 3.5/4 5-12

end ent '.. 1: 4, 1:

3.5/4.5 CORE AND C0?ffAIN'Efrf C00L2ftG SYSILMS LIMITINO CONDITIONS FOR OPERATIOP SURVEILLANCE REQUIREMENTS 2.'.D reufr-ent Area Ceelers 4.5.D Eouir-ent Area Ceelers

1. The equipment area cooler 1.

Each equipment area cooler associated with each RHR is operated in conjunction pump and the equipment with the equipment served area cooler associated by that particular cooler: with each set of core therefore, the equipment spray pumps (A and C area coolers are tested at or B and D) must be the same frequency as the OPERABLE at all times pumps which they serve. when the pump or pumps served by that specific cooler is considered to be CPIRABLE.

2. When an equipment area cooler is not CPERABLE, the pump (s) served by that cooler must be considered inoperable for technical specification purposes.

E. Hirh Pressure Coolant Iniecticn E. E3th Pressure Coolant System (HPCIS) Iniectien System (HFC'll 1. The HPCI system shall be 1. HPCI subsystem testing CPERABLE: shall be performed as follovst (1) FRIOR TO STARTUP from a a. Simulated Once/ COLD CORDITION, or Automatic operating Actuation cycle Test (2) whenever there is b. Pu=p Per irradiated fuel in the Opera-Specificatien reactor vessel and the bility 1.0.MM ( reactor vessel pressure is greater than 12J psis, c. Motor Oper-Per except as specified in ated Valve Sp e c i f'.c a t ie n Specification 3.5.E.2. Operability 1.0.KM d. T1ov Rate at Cnce/3 no rmal months reaeter vessel operating pressure STN 3.5/4.5-13 A en: ent ',0. 1:0 Unit 3 e

s' _3. 5 / 4. 5 COPE AND c0NTAINMENT c00 LING SYSUfJ LIMIT!NO CCNDIT!0d5 TOR OPERATION SURVEILLANCE REQUIREMINTS 3.5.E Hith Pressure teclant Inieetien 4.5.E Hith Pressure coeltat. System (HPCIS) Iniection System (HPCL11 4.5.E.1 (Cont'd) e. T1ov Rate at once/ 150 psig operating cycle l The NPCI pump shall deliver at least 5000 grm during each flow rate test. 2. If the HTC! system is 2. When it is determined that inoperable, the reactor Lay the HPCl$ is inopetatie, the re-nin in eteratten for a ADS actuation logic, the period not to exceed 7 days. RCICS, the RERS (LPCI), and provided the ADS, CSS, RHR$ the CSS shall be (LPCI), and RCICS are demonstrated to be CPERABLE OFIRABLE. immediately. The RCICS and ADS logic shall be demonstrated to be OPEFABLE daily thereafter. 3. If Specifications 3.5.E.1 or 3.5 E.2 are not met, an orderly shutdown shall be initiated and the reactor vessel pressure t shall be reduced to 122 psig or less within 24

hours, t

T. Etttter cere Isolatten coolint T. Erneter core Isolatien ceeiine ly11gm (FCICS) System (RCICS) 1. The RCICS shall be CFERABLE: 1. RCIC Subsystem testing shall be perfereed as follovst (1) TRIOR TO STARTUP from a COLD CONDITICN, or

a. Simulated once/

Automatic operating (2) whenever there is ActJation Test cycle irradiated fuel in the reactor vessel and the

b. Pump Fer $teetfi-reactor vessel pressure Operability cation 1.0.MP.

is ateve 10: rsig, t except as specified in j

c. Motor-Operated Fer Valve Specificatten l

3.5.T.2. l q Operability 1.C.?"4 l 3.5/4.5-14 ITN '~'"O~'0I I#' 'L.i t 3

e 3,g/4.5 CORI AND C0fffille8.Ilff C00 LINA _ SYSTEMS I ,!MITING CCNDIT!0NS FOR OPERAT!0N SURVEILLANCE REQUIREME!ff$ j 3,$.T Reseter Core feelatien Ceolier 4.5.T teneter Core Isolatten Ceelir.r Svaten (RcfCS) $vataa.(RCICS) 4.5.T.1 (Cont'd) f l d. Flow Rate at once/3 I normal reactor months vessel operating pressure e. Flev Rate at once/ 150 pois operating cycle [ The RCIC pump shall I deliver at least 600 spm f during each flow test. I t 2. If see RCICS is INOPERABLT., 2. When it is determined that the reactor may remain in the RCICS is INOPERAELE, the ( eperation for a period not HPCIS shall b9 demonstrated to exceed 7 days if the to be OPERABLE immediately i HPCIS is CPERABLE during and weekly thereafter. l such time. 3. If Specifications 3.5,T.1 er 3.5.T.2 are not mete an I' orderly shutdown shall be initiated and the r6 actor shall be depressurized to less than 105 psig within 24 hours. G. Auteratie Deereasurinatien G. Automatie Deereasuritatirn Svates (ADS) Svaten (Ars) 1. Tour of the six valves of 1. During e.sch operating the Automatic cycle the following Depressurination System tests shall be performed shall be OPERABLE: on the ADSt I (1) prior to a STARTUP

a. A simulated autcentie

) from a Cold Conditten, actuation test shall or, be performed prior to i I STARTUP after each l 4 l l i + l E T:; Unit 3 3.5/4.5-15 l t I t I

3.5 EASES (Ctnt'd) 3.5.M References 1. Loss-of-Coolant Accident Analysis for Browns Terry Nuclear Plant Unit 3, NEDO-24194A and Addenda. 2. "BWR Transient Analysis Model Utilizing the RETRAN Progrot," TVA-TR81-01-A. 1 3. Generic Reload Tuel Application, Licensing Topical Report, NIDE-24011-F-A and Addenda, i i i i l t 1 i I I i 1 l ??N i Ur.!: 3 3.5/4.5-35

en: ent *,:, 11:

i I l i l l

4,$ gage and CenQainmeng_foolint SYsQe?S SUPYelllanct frecuenciel The testing interval for the core and containment cooling systems is based on industry practice, quantitative reliability analysis, judgment, and practicality. The core cooling systems have not been designed to be fully testable during operation. For example, in the case of the HPCI, autcmatic initiaticn during power operation would result in pumping cold water into the reactor vessel which is not desirable. Complete ADS testing during power operation causes an undesirable loss-of-coolant invent o ry. To increase the availability of the core and containment cooling system, the coeponents which enke up the syster., i.e., instrumentation, pu=ps, valves, etc., are tested frequintly. The pumps and motor operated injection valves are also tested in accordance with Specification 1.0.MM to assure their operability. A simulated automatic actuation test once each cycle combined with testing of pumps and 4 l injection valves in accordance with specification 1.0.MM is deemed to be l adequate testing of these systems. When eerpenents and subsystems are out-of-service, overall core and centainment cooling reliability is maintained by demonstrating the operability of the remaining equipment. The degree of operability to be det:nstratea deper.ds cr. the nature of the reascn for the out-of-service i equipment. For routine out-of-service periods caused by preventive esintenance, etc., the pump and valve operability checks vill be performed to de :nstrate operability of the remaining ccepenents, i However, if a failure, design deficiency, cause the cutage, then the det:nstratien of operability shculd be thorough enough to assure that a generic problem does not exist. For example, if an out-of-service period was caused by f ailure of a pump to deliver rated caper.ity due to a design deficiency, the other pu=ps of this type might be subjected to a flow rate test in addition to the operability checks. Whenever a CSCS system er loop is =ade inoperable because of a required } test or calibration, the other CSCS systems or loops that are required to be operable shall te censidered CTIRABLE if they are within the required surveillance testing frequency and there is no reason to suspect they are inoperable. If the function, system, or loop under test or calibration I is found inoperable or exceeds the trip level setting, the LCO and the required surveillance testing for the system or loop shall apply. 1 l Redundant CTERABLE corponents are subjected to increased testing during l j equipeent cut-of-service tiers. This adds further eenservatism and increases assurance that adequate ecoling is available should the need

arise, t

j taxi-u-Averjat Flanar LHOR, LHGR _and E*EE l The KATLHGR, LHOR, and M FR shall be checked daily to determine if fuel i l burnup or control rod movement har caused changes in power distribution, l Since changes due to turnup are slov, and only a few control rods are coved daily, a daily check of power distributien is adequate. l l l !TN 3.5/4.5-36 r e n.., ;, ;3: I L'ni t 3 \\ l l I l I

e e 3.6/4.6 PRIMARY SYSTEM 10UNDAtX LIMITING r)MDITIONS TOR OPERATION SURVIILLANCE REQUIREMENTS 3.6.C. coolant Leakane 4.6.C. Coelant Leakana

1. a.

Any time irradiated 1. Reactor coolant fuel is in the system leakage shall reactor vessel and be checked by the reactor coolant sump and air sampling temperature is above systen and recorded 212'r, reactor coolant at least once per leakage into the 4 hours. primary containment from unidentified sources shall not exceed 5 gre. In additten, the total reactor coolant system leakage into the primary centainment shall not exceed 25 spm. b. Anytime the reactor is in RUN mode, reactor coolant leakage into the primary containment from unidentified sources shall not increase by more than 2 grm averaged over any 24-hour period in which the reacter is in the RUN mode except as defined in 3.6.C.1.c below. c. Durins the first 24 hours in the RUN modt following STAxTUP, an increase in reacter coolant leakage into the primary contairment of >2 grm is acceptable as long as the requirements of 3.6.C.1.a are met. l l l 1 E P: l' nit 3 3.6/4.t-9 A~ e n d f a t '.C. 103

l I 3.6/4.6 Pn2MAny SYSTEM nointDARY LIMITING CONDITIONS FOR OptRATION SURVEILLANCE REQUIREMENTS i 3.6.C Coelant Leakant 4.6.C Coolant Leakane l t 2. Both the sump and air sampling

2. With the air samplins systema shall be OPERABLE during systes inoperable, grab l

REACTOR POWER OPERATION. From samples shall be obtained f and after the date that one of and analysed at least once these systems is made or found every 24 hours. to be inoperable for any reason, [ REACTOR POWER OPERATION is s paraissible only during the f succeeding 24 hours for the sump } system or 72 hours for the air sampling system. The air samplins systes n# t be removed from service 1 a period of 4 houra for ( calibration, function testing, ( and maintenance without providing a temporary monitor. 3. If the condition in 1 or 2 j above cannot be met, an orderly i shutdown shall be initiated and the reactor shall be placed l l in the COLD SNUTDOWN CONDITION j vithin 24 hours, j r 3.6.D. Relief Valves 4.6.D. Relief Valves l l 1. When more than one relief valves 1. Approximately one-half cf l l are known to be failed, an all relief valves shall orderly shutdown abs 11 be he bench-enecked or l initiated and the reactor replaced with a [ depressurized to less than 105 bench-checked valve psig within 24 hours. each operating cycle. All 13 valves will have been checked or replaced upon the completion of every second cycle. 2. In accordance with l Specification 1.0.199, l l each relief valve shall te manually opened until l thermocouples and I acoustic monitors i downstream of the valve s indicate steam is flowing from the valve. E T:: 3.6/4.6-10 .+en: e n y, 10?, 12: i Unit 3 l l i l

2.n.vo .v.e r..... .o The requirements established above apply when the nuclear system car. be pressurized above ambient conditions. These requirements are applicable at nuclear system pressures below normal optrating pressures because abnormal eperational transients could possibly start at these conditions such that eventual overpressure relief would be needed. Hevever, these transients are euch less severe, in terms of pressure, than those starting at rated cenditions. The valves need not be functional when the vessel head is re=oved, since the nuclear system cannot be pressurized. Referentti 1. Nuclear System Pressure Relief System (BrNP TSAR Subsection 4.4) 2. "Protection Against Overpressure" (ASMI Boiler and Pressure vessel Code, Section III, Article 9) 3. Brovns Terry Nuclear Plant Design Deficiency Report--Target Rock Safety-Relief Valves, transmitted by J. E. Gilliland to T. E. Kruesi, August 29, 1973 3.6.E/4.6.E Jet Puees Tailure of a jet pump nozzle assembly holddovn mechanism, nozzle assembly and/or riser, vould increase the cross-sectional flow area for blevdown following the design basis double-ended line break. Also, failure of the diffuser would eliminate the capability to reflood the core to two-thirds height level following a recirculation line break. Therefore, if a failure occurred, rer41rs must be made. The detection technique is as fo11 ova. With the two recirculation pu=ps balanced in speed to within i 5 percent, the flow rates in both recirculation loops vill be verified by control room monitoring instruments. If the two 3 flow rate values do not differ by more than 10 percent, riser and nozzle assembly integrity has been verified. If they do differ by 10 percent or more, the core flow rate measured by the jet pump diffuser differential pressure system must be checked against the core flow rate derived from the measured values of loop flov to core flev correlation. If the difference between measured and derived core flow rate is 10 percent or more (with the derived value higher) diffuser measurements vill be taken to define the location within the vessel of failed jet pump nozzle (er riser) and the unit shut dovn for repairs. If the potential blevdovn flev area is increased, the s' tem ETN Crit 3 3.t/4.e-31

J.e.cre.o.w swvu6 we i reasonably in a matter of a sev hours utilizing the available leakage detection schemes, and if the origin cannot be determined in a reasonably short time, the unit should be shut down to allow further investigation and eerrective action. The two spa limit for coolant leakage rate increase over any 24 hour period is a limit specified by NRC (Reference 2). This limit applies only during the RUN mode to avoid being penalized for the expected coolant leakage increase during pressurization. l The total leakage rate consists of all leakage, identified and unidentified, f which flows to the dryvell floor drain and equipment drain sumps. The capacity of the dryvell floor sump pump is 50 sps and the capacity of the dryvell equipment sump pump is also 50 spm. Removal of 25 spa from either 'of these sumps can be accompitshed with considerable margin. l t References i 1. Nucles: System Leakast Rate Limits (BTNP TSAR Subsection 4.10) l 2. Safety Evaluation Report (SER) on IE Bulletin 82-03 l 3.e.D /4. 6.D Relief Valves l To meet the safety basis, 13 telief valves have been installed on the unit l' vith a total capacity of 83.77 percent of nuclear boiler rated steam flow. The analysis of the worst overpressure transient, (3-second closure of all main steam line isolation valves) neglecting the direct scram (valve position scram) results in a maximum vessel pressure which, if a neutron flux scram is t j assumed considering 12 valves CPERABLE, results in adequate margin to the code l allevable overpressure limit of 1,375 psig. To meet operational design, the analysis of the plant isolation transient l (generator load reject with bypass valve failure to open) shows that 12 of the f 13 relief valves limit peak system pressure to a value whir' as tell below the allowed vessel overpressure of 1,375 pais. t Experiente in relief and safety valve operation shows that 6.aating of 50 percent of the valves per year is adequate to detect failures or deteriorations. The relief and safety valves are benchtested every second operating cycle to ensure that their setpoints are within the i 1 percent tolerance. The relief valves are tested in place in accordance with Specification 1.0 MM to estab1!'% that they vill open and pass steam. l j I i i I RTN 3.6/4.6-30 A en: ent '.o. 130 Unit 3 i i 1 1

e 3.7/4.7 C0?MAlfME!C SYSTEMS LIMITING CONDITIO!IS FOR OPERATION SURVEILLANCE REQUIREKr:fjTS 3.7.A PEI"ATY C0!EAI!M?C 4.7. A FFIMARY C0fMAlfME!C 3. Pressure Sureressien Cha-ber -

3. Pressure Sureressien Cha-ter-Reactor Buildina Vacuus Breakers Reactor Buildina Vacuum Breakers

a. Except as specified in

a. The pressure suppression 3.7.A.3.b belov, two pressure chamber-reactor building suppression chamber-reactor vacuum breakers shall be building vacuum breakers shall exercised in accordance with j

be OPERABLE at all times when Specification 1.0.MM and the primary conta'n=ent integrity associated instrumentation is required. The setpoint including setpoin; shall be of the differential pressure functionally tested 'or proper instrumentation which actuates cperation each three tenths. the pressure suppression I chamber-reactor building vacuum breakers shall be j t l 0.5 psid. 1, 1 I

b. Frem and after the date i

b. A visual examination and that one of the pressure determination that the suppression chamber-reactor force required to open each building vacuum breakers is vacuum breaker (check valve) made or found to be inoperable does not exceed 0.5 psid for any reason, reactor vill be made each refueling operation is permissible only l

outage, i

during the succeeding seven days, provided that the repair procedure does not violate primary containment integrity. 4 Drvvell-Pressure Surtressien

4. Dryvell-Pressure Surettaglin

[ht-tgr Vaeue9 Breakers Chamber Vacuw-Breakers

a. When primary contain-ent is

a. Each dryvell-suppression required, all dryvell-cht:ber vacuu: breaker suppression chteber vacuum j

shall be tested in accordance j breakers shall be OTERABLE 'sith Specification 1.0.MM. I and positioned in the fully closed position (except l

b. When it is determined that during testing) except as i

specified in 3.7.A.4.b and two vacuum breakers are 3.7.A.4.c., below. inoperable for opening at a time when operability is

b. One dryvell-suppression required, all other vacuum chamber vacuum breaker may breaker valves shall be be nenfully closed so long exercised it ediately and as it is determined to be not every 15 days thereafter untti core than 3' cren as indicated the inoperable valve has been by the positten lights, returned to normal service.

ETN 3.7/4.7-10

-c-t-e

'0 7, '. Unit 3

ti. +, s I l }.7/4_7 CONTAINMINT SYSTEMS l r LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQt'it?MENTS i i 4.7.A. Primary Centair. ment [ 4.7.A.2. (Cont'd) l

j. Continuous Laak Rate Monitorina f

I When the primary t containment is inerted i the containment shall be continuously monitored [ for gross leakage by r review of the inerting e systes sakeup requirements. This l monitorina systes may be l taken out of service for i maintenance but shall be returned to service as t soon as practicable. } { k. The interior surfaces of the drywell and torus above the level one foot 1 below the normal water l line and outside surfaces of the torus [ below the water line shall be visually } inspected each operatirs j cycle for deterioration and any signs of t structural danase with t particular attention to piping connections and f supper.s and for signs I of distress or [ displacement. t i 1 I 1 L { r i 4 I l I I I t 'ni 3.7/a.7-9 5 Unit 3 i l I

9 i t t 3,y/4.7 coltf Afl0ENT SYSTEMS l LIMITINO CONDITIONS FOR O?!kATION SURVEILLANCE REQUIREh2NTS j 3.7.A Pri-ary Centain-ent 4.7.A Primary Centairsent [ 3.7.A.4 (Cont'd) 4.7.A.4 (Cont'd)

c. Two dryvell-suppression

c. Each vacuum breaker valve chamber vacuuta breakers shall be inspected for i

may be determined to be proper operation of the l inoperable for opening. valve and limit switches in accordance with Specification 1.0. Pet. l

d. If specifications 3.7.A.a.a.

d. A leak test of the dr>vell 3.7.A.4.b, er 3.7.A.4.c.

to suppreestem thaaber j cannot be ret, the structure shall be conducted unit shall be placed during each operating cycle. in a COLD SMUTDOWN Acceptable leak rate is [ ' CONDITION in an orderly 0.09 lb/sec of primary manner within 24 hours. eentainment atmosphere with i 1 psi differential. 5. Orvaan eeneantratien

5. crvaan ceneantration a.

Con *ainment atmosphere shall be

8. The primary containment i

I reduced to less than 4% oxygen orygen concentration shall with nitrogen gas during reactor be sensured and recorded power operation with reactor daily. The oxygen l coolant pressure above 100/psis, measurement shall be adjusted except as specified in 3.7 A.S.b. to account for the uncertainty et the method used by adding a predetermined error function. f b. Within the 24-heur period

b. The methods used to measure l

l suteequent to placing the reactor the primary containment I in the RUN MODE followins a shut-erygen concentration shall f down, the containment atmosphere be calibrated once ev3ry I orygen concentration shall be refue*.ing cycle. l reduced to less than 4% by volume. and maintained in this condition. Deinerting any commence 24 hours i . tier to a shutdown. J i c. If plant control air is being used

c. The control air supply valve to supply the pneumatic control for the pneumatic control systes inside primary containment, system inside the primary the reactor shall not be started, containment shall te verified or if at power, the reactor shall closed prior to reactor startup be brought to a COLD SWTDOVN and zenthly thereafter.

CONDIT!CN within 24 hours, d. If the specifications of 3.7.A.5.a ; through 3.7.A.5.b cannet be met, j an orde,ly shutdown shall be initiatedandtheresetershallbel in a COLD SEUILOWN CONOITICN l vithin 24 h;urs. UN 3.7/4.7-11 . en: en:,:, ;;3, ;;. U.ti t 3 .. I

?.7/4.7. C0fCAINMEffr SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.7.A. Primary C2Dtainment 4.7.A. Primary Containment 6', DIy*' / easion_ Chamber

6. DIyvell-Sueeression Chamber Q1j ;

- dI,tssu re Differential Pressure a. s1 pressure between a'. The pressure differential 1 and suppression between the dryvell and h- 'e snall be maintained suppression chamber shall .. I to or greater than be recorded at least once 1.1 psid except as specified each shift. in (1) and (2) belovs (1) This differential shall be established within 24 hours of achieving operating temperature and pressure. The differee'..ial pressure may be reduced to less than 1.1 paid 24 hours prior to a scheduled s'nutdown. (2) This differential may be decreased to less than 1.1 paid for a maximum of four hours during required operability testing of the HPCI system, RCIC system and the dryvell-pressure suppression chamber vact un breakers.

b. If the differential pressure of Specification 3.7.A.6.a cannot be maintained and the differential pressure cannot be restored within the subsequent six-hour period, an orderly shutdown shall be initiated and the reactor shall be in the COLD SHUTD0'@

CONDITION within 24 hours. I 3FN 3.7/4.7-12 Unit 3 1 i

3,7/4,7 CONTAf f0'ENT SY3TEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.7.C. Secondary Containment

4. If refueling zone secondary containment cannot be maintained the following conditions shall be met:

a. Handling of Opent fuel and all operations over spent fuel pocle and open reactor wells containing fuel shall be prohibited.

.. / + %^3? jt gy{,3(. 4*g

b. The standby gas treatment k,;

system suction to the refueling zone vill be -\\-

h.,i t '-

biccKed except for a controlled leakage area ,~g;ATf-- asi., sized to.ssure the ' ' "M' ' M achieving of a vscuum of at least 1/4-inch of water and not over 3 inches of water in all three reactor zones. D. Primary Containment Isolation Valves D. Prima ry Containment Iselation / Valves

1. When primary containment

1. The primary c, ntain:. ant integrity is required all isolation valves isolation valves listed in Table surveillance shall be

'.7.A and all reactor coolant performed as follows: system instrument line flow check valves shall be OPERABLE

a. At least once per except as specified in 3.7.D.2.

operating evele, the OPERABLE isolation valves that are power operated and automatically initiated shall be tested for simulated automatic initiation, and in accordance with Specification 1.0.MM, I tested for closure times. 1 l l l l BTN 3.7/4.7-17 A e,d ent fio,116.13; Unit 3 1

<e. 3.7/4.7_ CONTAf fNENT SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3,7.D. Primary Centainment Isolation 4.7.D. Primary Containment Valves Isolation Valves 4.7.D.1 (Cont'd)

b. In accordance with Specification 1.0.MM, all normally open power operated isolation valves shall be functionally tested.

c. (Deleted)

d. At least once per operating cycle the operability of the reactor coolant system instrument line flow check valves shall be verified.

2. In the event any isolation valve

2. Whenever an isolation valve specified in Table 3.7.A becomes listed in Table 3.7.A is inoperable, reactor operation inoperable, the position of at may continue provided at least least one other valve in each one valve, in each line having line having an inoperable valve an inoperable valve, is OPERABLE shall be recorded daily, an within 4 hours either:

a. The inoperable valve is restored t <. OPERABLE status, or

b. Each affected line is isolated by use of at least one deactivated containment isolation valve secured in the isolated position.

3. If Specification 3.7.D.1 and 3.7.D.2 cannot be met, an orderly shutdown shall be initiated and the reactor shall be in the COLD S!!UTDOW CONDITION within 24 hours.

BFN 3.7/4.7-18 Amendment No. 116. 117. 130 Unit 3 y.m w

'J. 7 /4. 7 CONTAIi, MENT SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.7.F. Primary Containment Purne 4.7.F. Primary Containment Purne System System

1. The primary containment shall

1. At least once every 18 months, be normally vented and purged the pressure drop across the through the primary containment combined HEPA filters and purge system. The standby gas charcoal adsorber banks shall treatment system may be used be demonstrated to be less than when primary containment purge 8.5 inches of water at system system is INOPERABLE.

design flow rate (i 10%).

2. a. The results of the in-place

2. a. The tests and sample cold DOP and halogenated analysis of Specification hydrocerbon tests at design 3.7.F.2 shall be performed flows on HEFA filters and at least once per operating charcoal adsorber banks shall cycle or once every show 199% DOP removal and 18 months, whichever occurs 199% halogenated hydrocarbon first er after 720 hours removal when tested in of system operation and accordance with following significant ANSI N510-1975.

painting, fire, or chemical release in any ventilation zone communicating with the system.

b. The results of laborstory

b. Coad DOP testing shall be carbon sample analysis performed after each shall show 185% radioactive complete or partial methyl iodide removal when replacement of the HIPA tested in accordance with filter bank or after any ASTM D3803 structural maintenance on (130*C 95% R.H.).

the system housing.

c. System flow rate shall be

c. Halogenated hydrocarbon shown to be within 10% of testing shall be performed design flow when tested in after each complete or accordance with ANSI N510-partial replacement of the 1975.

charcoal adsorber bank or after any structural maintenance on the system i housing. I i ETN 3.7/4 7-21 A end ent 'a,114, ;?! Unit 3

~ 3.7/4 7 C.QNTAINMENT SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.7.G. Containment Atmosphere 4.7.G. Containment Atmosehere Dilution System (CAD) Dilution System (CAD) 1. The Containment Atmosphere 1. System Oeerability Dilution (CAD) System shall be OPERABLE vith: a. Two independent a. Cycle each solenoid systems e sable of operated air / nitrogen supplyint.itrogen valve thro.,h at least to the dryvell and one complace cycle of

torus, full travel in accordance with Specification 1.0.MM, and at least ence per month verify that each manual valve in the flow path is open.

b. A minimum supply of b. Verify that the CAD 2,500 gallons of System contains a liquid nitrogen per minimum supply of i system. 2,500 gallons of liquid nitrogen twice per week. 2. The Containment Atmosphere Dilution (CAD) System shall be OPERABLE whenever the reactor is in the RUN MODE. 3. If one system is inoperable, the reactor may remain in operation for a period of 30 days provided all active components in the other system are OPERABLE. 4 If Specifications 3.7.G.1 and 3.7.G.2, or 3.7.G.3 cannot be met, an orderly shutdown shall be initiated and the reactor shall be in the COLD SHUTDOWN CONDITION within 24 hours. 5. Primary containment pressure shall be limited to a maximum of 30 psig during repressurization following a loss of coolant accident. BTN 3.7/4.7-22 Amendment No. 130 Unit 3

3.7/4.7 C0tCAIfMENT SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE RIQUIREMENTS 3.7.T. Primary Containment Purze 4.7.F. Erimary Containment Purra System System

1. The primary containment shall

1. At least once every 18 months, be normally vented and purged the pressure drop across the through the primary containment combined HEPA filters and purge system. The standby gas charcoal adsorber banks shall treatment system may be used be demonstrated to be less than when primary containment purge 8.5 inches of water at system system is INOPERABLE, design flow rate (i 10%).

l l l

2. a. The results of the in-place

2. a. The tests and sample cold DOP and halogenated analysis of Specification hydrocarbon tests at design 3.7.T.2 shall be performed flows on HEPA filters and at least once per operating charcoal adsorber banks shall cycle or once every show 199% DOP removal and 18 months, whichever occurs 199% halogenated hydrocarbon first or after 720 hours removal when tested in of system operation and accordance with followins significant ANSI N510-1975.

painting, fire, or chemical release in any ventilation zone communicating with the system. l I

b. The results of laboratory

b. Cold DOP testing shall be carbon sample analysis performed after each shall show 185% radioactive complete or partial i

methyl iodide removal when replacement of the HEPA tested in accordance with filter bank or after any l ASTM D3803 structural maintenance on ( (130'C 95% R.H.). the system housing. 1

c. $ stem flow rate shall be

c. Halogenated hydrocarbon 3

shown to be within 110% of testing shall be performed design flow when tested in after each complete or accordance with ANSI N510-partial replactment of the

1975, charcoal adsorber bank or after any structural l

maintenance on the system l housing. l BTU 3.7/4.7-21 Amendment No. 139, 146 Unit 2

e 3.7/4.7 CONTAINMENT SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.7.G. Containment Atmosehere 4.7.G. Containment Atmosehere Dilution System (CAD) Dilution System (CAD) 1. The Containment Atmosphere 1. System Oeerability Dilution (CAD) System shall be OPERABLE with:

a. Two independent a.

Cycle each solenoid systems capable of operated air / nitrogen supplying nitrogen valve through at to the dryvell and least one complete

torus, cycle of full travel in accordance with Specification 1.0.MM, l

and at least once per month verify that each manual valve in the flow path is open,

b. A minimum supply of b.

Verify that the CAD 2,500 gallons of System contains a liquid nitrogen per minimum supply of system. 2,500 gallons of liquid nitrogen twice per va;k. 2. The Containment Atmosphere Dilution (CAD) System shall be OPERABLE whenever the reactor is in the RUN MODE. 3. If one system is inoperable, the reactor may remain in operation for a period of 30 days provided all active components in the other system are OPERABLE. i BrH 3.7/4.7-22 Amendment No. 155 Unit 2

3.7/4.7 BASES (Cont'd) With doors closed and fan in operation, DOP aerosol shall be sprayed externally along the full linear periphery of each respective door to check the gasket seal. Any detection of DOP in the fan exhaust shall be considered an unacceptable test result and the gaskets repaired and test repeated. If significant painting, fire or chemical release or. curs such that the HEPA filter or charcoal adsorber could become contaminate.1 from the fumes, chemicals or foreign material, the same tests and semple analysis shall be performed as required for operational use. The dete>tinee. ion of significance shall be made by the operator en duty at the time at the incident. Knowledgeable staff members should be consulted g4 r to making this determination. Demonstration of the automatic initiation capability and operability of filter cooling is necessary to assure system performance capability. If one Standby Gas Treatment System is inoperable, the other systems must be tested daily. This substantiates the availability of the operable systems and thus reactor operation and refueling operation can continue for a limited period of time. 3.7.D/4.7.D Primary Containment Isolation Valygg Double isolation valves are provided on lines penetrating the primary containment and open to the free space of the containment. Closure of one of the valves in each line would oe sufficient to maintain the integrity of the pressure suppression system. Automatic initiation is required to minimize the potential leakage paths from the containment in the event of a loss-of-coolant accident. Group 1 - Process lines are isolated by reactor vessel low water level (378") in order to allow for removal of decay heat subsequent to a scram, yet isolate in time for proper operation of the Core Standby Cooling Systems. The valves in Group 1, except the reactor Vater sample line valves, are also closed when process instrumentation detects excessive main steam line flow, high radiation, low pressure, or main steam space high temperature. The reactor water sample line valves isolate only on reactor lov vater level at 378" or main steam line high radiation. .Grouc 2 - Isolation valves are closed by reactor vessel lov vater level (538") or high dryvell pressure. The Group 2 isolation signal also "isolates" the reactor building and starts the Standby Gas Treatment System. It is not desirable to actuate the G7oup 2 isolation signal by a transient or spurious signal. Groue 3 - Process lines are nor= ally in use, and it is therefore not desirable to cause spurious isolation due to high dryvell pressure resulting from nonsafety related causes. To protect the reactor from a possible pipe break in the system, isolation is provided by high temperature in the cleanup system area or high flev through the inlet to the cleanup system. Also, since the vessel could potentially be drained through the cleanup system, a low-level isolation is previded. ETN 3.7/4.7-47 Unit 3

w

s. sin.

una-a r~-... -o Greues 4 and 5 - Process lines are designed to remain operable and mitigate the consequences of an accident which results in the isolation of other process lines. The signals which initiate isolation of Groups 4 and 5 process lines are therefore indicative of a condition which would render them inoperable. Greue 6 - Lines are connected to the primary containment but not directly to the reactor vessel. These valves are isolated on reactor low water level (538"), high dryvell pressure, or reactor building ventilation high radiation which would indicate a possible accident and necessitate primary containment isolation. Greue 7 - Process lines are closed only on the respective turbine steam supply valve not fully closed. This ensures that the valves are not open when HPCIS or RCICS action is required. GI222_1 - Line (traveling in-core probe) is isolated on high dryveli pressure or reactor lov vater level (538"). This is to assure that this line does not provide a leakage path when containment pressure or reactor water level indicatcc c possibic accident ccnditicn. The maximum closure time for the automatic isolation valves of the primary containment and reactor vessel isolation control system have been selected in consideration of the design intent to prevent core uncovering following pipe breaks cutside the primary containment and the need to contain released fission products following pipe breaks inside the primary containment. In satisfying this design intent, an additional margin has been included in specifying maximum closure times. This margin permits identification of degraded valve performance, prior to exceeding the design closure times. In order to assure that the doses that may result from a steam line break do not exceed the 10 CTR 100 guidelines, it is necessary that no fuel rod perforation resulting from the accident occur prior to closure of the main steam line isolation valves. Analyses indicate that fuel rod cladding perforations would be avoided for main steam valve closure times, includint instrument delay, as long as 10.5 seconds. These valvas are highly reliable, have low service requirements, and most are normally closed. The initiating sensors and associated trip logic are also checked to demonstrate the capability for automatic isolation. The test interval of once per operating cycle for automatie initiation results in a failure probability of 1.1 x 10-' that a line vill not isolate. More frequent testing for valve operability in accordance with Specification 1.0.MM l results in a greater assurance that the valve vill be operable when needed. The main steam line isolation valves are functionally tested per Specification 1.0.MM to establish a high degree of reliability. The primary containment is penetrated by several small ditmeter instrument lines connected to the reactor coolant system. Each instrument line contains a 0.25-inch restricting orifice inside the primary containment and an excess flow check valve outside the primary containment. BTN 3.7/4.7-48 A endment No. 130 Unit 3 --,n______-----.-

3.7/4./ masso scout aj w With doors closed and fan in operation, DOP aerosol she.11 be sprayed externally along the full linear periphery of each respective door to check the gasket seal. Any detection of DOP in the fan exhaust shall be considered an unacceptable test result and the gaskets repaired and ;est repeated. If significant painting, fire or chemical release occurs such that the HEl'A filter or charcoal adsorber could become contaminated from the fumes, chemicals or foreign material, the same tests and sample analysis shall be performed as required for operational use. The determination of significant shall be made by the operator on duty at the time of the incident. Knowledgeable staff members should be consulted prior to making this determination. Demonstration of the automatic initiation capability and operability of filter cooling is necessary to assure system performance capability. If one standby gas treatment system is inoperable, the other systems must be tested daily. This substantiates the availability of the operable systems and thus reactor

cration and rc f a;;i;. creratien can centinue for a litited period of tite.

3.7.D/4.7.D Primary Containment Isolation Valves Double isolation valves are provided on lines penetrating the primary containment and open to the free space of the containment. Closure of one of the valves in each line would be sufficient to maintain the integrity of the pressure suppression system. Automatic initiation is required to minimize the potential leakage paths from the containment in the event of a LOCA. Groue 1 - Process lines are isolated by reactor vessel low water level (378") in order to allow for removal of decay heat subsequent to a scram, yet isolate in time for proper operation of the core standby cooling systems. The valves in Group 1, except the reactor water sample line valves, are also closed when process instrumentation detects excessive main steam.line flow, high radiation, low pressure, or main steam space high temperature. The reactor water sample line valves isolate only on reactor low water level at 378" or main steam line high radiation. Greue 2 - Isolation valves are closed by reactor vessel low water level (538") or high dryvell pressure. The Group 2 isolation signal also "isolates" the reactor building and starts the standby gas treatment system. It is not desirable to actuate the Group 2 isolation signal by a transient or spurious signal. l Groue 3 - Precess lines are normally in use, and it is therefore not desirable to cause spurious isolation due to high dryvell pressure resulting from nonsafety related causes. To protect the reactor from a possible pipe break in the system, isolation is provided by high temperature in the cleanup system area or high flow through the inlet to the cleaGup system. Also, since the vessel could potentially be drained through the cleanup system, a low-level isolation is provided. BTN 3.7/4.7-49 Unit 2

s.tr*.i naams suvu6 of ~ Greues 4 and 5 - Process lines are designed to remain operable and mitigate the consequences of an accident which results in the isolation of other process lines. The signals which initiate isolation of Groups 4 and 5 process lines are therefore indicative of a condition which would render them inoperable. Greue 6 - Lines are connected to the primary containment but not directly to the reactor vessel. These valves are isolated on reactor low water level (538"), high drywell pressure, or reactor building ventilation high radiation which would indicate a possible accident and necessitate primary containment isolation. Groun 7 - Process lines are closed only on the respective turbine steam supply valve not fully closed. This assurts that the valves are not open when EPCI or RCIC action is required. Groue 8 - Line (traveling in-core probe) s isolated on high drywell pressure or reactc: low water level (538"). This in to assure that this line does not provide a leakage path when containment ytessure or reactor water level indicater a possible accident condition. The maximum closurc time for the automatic isolation valves of the primary containment and reactor vessel isolation control system have been selected in consideration of the design intent to prevent core uncovering following pipe breaks outside the primary containment and the need to contain released fission products following pipe breaks inside the primary containment. In satisfying this design intent, an additional margin has been included in specifying maximum closure times. This margin permits identification of degraded valve performance prior to exceeding the design closure times. In order to assure that the doses that may result from a steam line break do not exceed the 10 CFR 100 guidelines, it is necessary that no fuel rod perforation resulting from the accident occur prior to closure of the main steam line isolation va.1.ves. Analyses indicate that fuel rod cladding perforations would be avoided for main steam valve closure times, including instrument delay, as long as 10.5 seconds. These valves are highly reliable, have low service requirements, and most are normally closed. The initiating sensors and associated trip logic are also checked to demonstrate the capability for automatic isolation. The test intervalofonceperoperatingcycleforautomaticinitiationresultsina failure probability of 1.1 x 10- that a line will not isolate. More frequent testing for valve operability in accordance with Specification 1.0.MM l results in a greater assurance that the valve will be operable when needed. The main steam line isolation valves are functior.nlly tested per Specification 1.0.MM to establish a high degree of reliability. The primary containment is penetrated by several small diameter instrument lines connected to the reactor cLolant system. Each instrument line contair.s a 0.25-inch restricting orifice inside the primary containment and an excess flow check valve outside the primary containment. BFN 3.7/4.7-50 A endment No. 155 Unit 2

T ~

p areg f'o UNITED STATES j' ' < s.. f f,,g NUCLEAR REGULATORY COMMISSION

/ WASHINGivN, D. C. 20655 j %... '.. f TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-296 BROWNS FERRY NOCLEAR PLANT, UNIT 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 130 license No. CPR-68 1. The hu: lear Regulatory Comission (the Comission) ha:, found that: A. The application for amendment by Tennessee Valley Authority (the licensee) dated October 27, 1987, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Ch3pter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and 4 safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D. The issuance of this amendrrent will not be inimical to the coreon defense and security or to the healt'h and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requireents have been satisfied.

i 2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No. DPR-68 is hereby arrended to read as follovs: (2) Technical Specifications The Technical Specifications centained in Append ces A and B, as revised through Amendment No.130, are hereby inec.porated in the license. The licensee shall operate the facility in accordance with the Technical Specifications. 3. This license amendment is effective as of its date of issuance and shall be implemented within 60 days from the date of issuance. FOR THE FUCLEAR REGULATORY C0 MISSION t t,s.w C Suzanne ' ack, Assistant Director for Projects TVA Projects Division Office of Special Projects

Attachment:

Changes to the Technical Specifications Date of Issuance: November 22, 1988

ATTACHMENT TO LICENSE AMENDMENT NO. 130 FACILITY OPERATING LICENSE NO DPR-68 DOCKET NO. 50-296 Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages. The revised pages are identified by the captioned amendment number and centain marginal lines indicating the area of change. Overleaf pages* are provided to maintain document completeness. REMOVE INSERT 1.0-11 1.0-11* 1.0-12 1.0-12 1.0-13 3.3/4.3-11 3.3/4.3-11* 3.3/4.3-12 3.3/4.3-12 3.3/4.4-1 3.3/4.4-1 3.3/4.4-2 3.3/4.4-2 3.5/4.5-1 3.5/4.5-1 3.5/4.5-2 3.5/4.5-2 3.5/4.5-3 3.5/4.5-3* 3.5/4.5-4 3.5/4.5-4 3.5/4.5-7 3.5/4.5-7 3.5/4.5-8 3.5/4.5-8* 3.5/4.5-9 3.5/4.5-9 3.5/4.5-10 3.5/4.5-10 3.5.4.5-12 3.5/4.5-12 3.5/4.5-13 3.5/4.5-13 3.5/4.5-14 3.5/4.5-14 3.5/4.5-15 3.5/4.5-15* 3.5/4.5-34 3.5/4.5-34* 3.5/4.5-35 3.5/4.5-35 3.5/4.5-36 3.5/4.5-36 3.6/4.6-9 3.6/4.6-9* 3.6/4.6-10 3.6/4.6-10 3.6/4.6-30 3.6/4.6-30 3.6/4.6-31 3/6/4.6-31' 3.7/4.7-9 3.7/4.7-9 3.7/4.7-10 3.7/4.7-10 3.7/4.7-11 3.7/4.7-11 3.7/4.7-12 3.7/4.7-12 3.7.4.7-17 3.7/4.7-17 3.7.4.7-18 3.7/4.7-18 3.7/4.7-21 3.7/4.7-21 3.7/4.7-22 3.7/4.7-22 3.7/4.7 47 3.7/4.7-47* 3.7/4.7-48 3.7/4.7-48

1.0 DEFINITIONS (Cont'd) GG. Site Boundarv - Shall be that line beyond which the land is not evned, leased, or otherwise centro 11ed by TVA. HH. Unrestricted Area - Any area at or beyond the site boundary to which access is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials or an'; area within the site boundary used for industrial, commercial, institutional, or recreational purposes. II. Q21e Eauivalent I-131 - The DOSE EQUIVALENT I-131 shall be the concentration of I-131 (in Ci/gm) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present. The thyroid dose cenversien factor used for this calculation shall be those listed in Table III of TID-14844 "Calculation of Distance Factors for Power and Test Reactor Sites". JJ. Caseous Waste Treatment System - The charcoal adsorbtr vessels installed on the discharge of the steam jet air efttter to provide delay to a unit's offgas activity prior tT release. KK. Members of the Public - Shall include all individuals who by virtue of their occupational status have no formal association with the plant. This catego: y shall include non-employees of the licensee who are permitted to use portions of the site for recreational, occupational, or other purposes not associated with plant functions. This category shall nql include non-employees such as vending machine servicemen or postmen who, as part of their formal job function, occasionally enter restricted areas. LL. Surveillane t - Surveillance Requirements shall be met during the OPERATIONA!. CONDITIONS or other conditions specified for individual limiting conditions for operation unless otherwise stated inin individual Surveillance Requirements. Each surveillance Requirement shall be performed within the specified time interval with, (1) A maximum allevable extention not to exceed 25% of the surveillance interval, but (2) The combined time entered for any 3 consecutive surveillance intervals shall not exceed 3.25 times the specified surveillance interval Performance of a Surveillance Requirement within the specified time interval shall constitute compliance and CPERABILITY requirements for a limiting condition for operation and associated action statements unless otherwise required by these specificatiens. Surveillance requirements do not have to be performed en inoperable equipment. ETN Unit 3 1.0-11

1.0 Utrin222vns s w win w/ MM. Surveillance reauirements for ASME Section XI Pume and Valve Proaram - Surveillance requirements for Inservice Testing of ASME Code Class 1, 2, and 3 components shall be applicable as follows: 1. Inservice testing of ASME Code Class 1, 2, and 3 pumps and valves shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55a(g), except where specific written relief has been granted by the Commission pursuant to 10 CFR 50, Section 50.55a(g)(6)(i). 2. Surveillance intervals specified in Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda for the inservice testing activities required by the ASMI Boiler and Pressure Vessel Code and applicable Addenda shall be applicable as follows in these technical specifications: ASMI B:iler and Pressure Vessel Required frequencies Code and applicable Addenda for performing inservice terminology for inservice .testina activities testina activitia3 Weekly At least once per 7 days Monthly At least once per 31 days Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days Yearly or annually At least once per 366 days 3. The provisions of Specification 1.0.LL are applicable to the above required frequencies for performing inservice testing activities. 4 Performance et the above inservice testing activities shall be in addition to other specified surveillance requirements. 5. Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supercede the requirements of any technical specification. BPN 1.0-12 A end ent No. 130 Unit 3

.0 Table 1.1 SURVEILLANCE FREQUENCY NOTATION NOTATION FREQUENCY S (Shift) At least once per 12 hours. D (Daily) At least once per normal calendar 24 hour day (midnight to midnight). W (Weekly) At least once per 7 days. M (Monthly) At least once per 31 days. Q (Quarterly) At least once per 3 months or 92 daye. SA (Semi-Annually) At least once per 6 months or 184' days. Y (Yearly) At least once per year or 366 days. R (Refueling) At least once per operating cycle. S/U (Start-Up) Prior to each reactor startup. N.A. Not applicable. P (Prior) Completed prior to each release. l A endment NO. 130 it 3

0* 3.3/4.3 REACTIVITY CONTROL LIMITING C0;"JITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.3.C. Scram Insertion Times 4.3.C. Scram Insertion Times 2. The average of the scram inser-

2. At 16-veek intervals, 10%

tien times for the three fastest of the OPERABLE control OPERABLE control rods of all rod drives shall be scram-groups of four control rods in timed above 800 psig. a two-by-two array shall be no Whenever such scram time greater thans measurements are made, an evaluation shall be made % Inserted From Avs. Scram Inser-to provide reasonable Folly Withdrawn tion Times (see) assurance that proper control rod drive 5 0.398 performance is being 20 0.954 maintained. 50 2.120 90 3.800 j a. The maximum scrWA insertion time for 90% int.ertion of any OPERABLE centrol rod shall not exceed 7.00 seconds. D. Reactivity Anomalies D. Reactivity Anomalies The reactivity equivalent of During the STARTUP test the difference between the program and STARTUP following actual critical rod refueling outages, the e configuration and t.e expected critical rod configurations configuration during power vill be compared to the operation shall not exceed 1%Lk. expected configurations at If this limit is exceeded, the selected operating conditions. reactor vill be shut dovn 7hese comparisons vill be until itt cause has been Jsed as base data for determined and corrective reactivity monitoring during actions have been taken as Lubsequegt power operation appropriate. throughout the fuel cycle. At specific power operating conditions, the critical red I configuration vill be compared to the configuration expected based upon appropriately corrected past data. This comparison vill be made at least every full power month. I I Ern 3.3/4.3-11

end ent '.o.

10' t'ni t 3 ~

4 3,3/4,3 REACTIVITY CONTROL LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.3.E. If Specifications 3.3.C and 4.3.E. Surveillance requirements are 3.3.D above cannot be met, as specified in 4.3.C and an orderly shutdown shall be 4.3.D above. Initiated and the reactor shall be in the Shutdown Condition within 24 hours. F. Scram Discharme Volume (SDV) F. Scram Discharme Volu.me (SDV) 1. The scram discharge volume 1.4. The scram discharge drain and vent valves shall volume drain and vent be OPERABLE any time that valves shall be verified the reactor protection open PRIOR TO STARTUP system is required to be and monthly thereafter. OPERABLE except as The valves may be closed specified in 3.3.T.2. intermittently for testing not to exceed 1 hour in any 24-hour period during operation. 1.b. The scram discharge volume drain and vent valves shall be demonstrated OPERABLE in accordance with Specification 1.0.MM. l 2. In the event any SDV drain 2. When it is determined or vent valve beco=es that any SDV drain or inoperable, REACTOR POWER vent valve is inoperable, OPERATION may continue the redundant drain or provided the redundant vent valve shall be drain or vent valve is demonstrated OPERABLE OPERABLE. immediately and weekly th1reafter. 3. If redundant drain or ver.: 3. No additional valves become inoperable, surveillance required. the reactor shall be in HOT STANDBY CONDITION within 24 hours. BTN 3.3/4.3-12 Amendment No. 104, 130 Unit 3

',.e.

3.4/4.4 STANDBY LIOUID CONTROL SYSTEM LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS

,4

TA!rr?? LIOUID CONTPOL SYSTEM 4.4 STANDEY LIOUID CONTROL SYSTEM Arelleability Acelicability Applies to the operating status Applies to the surveillance of the Standby Liquid Control requirements of the Standby System.

Liquid Control System. Obieetive Obiective To assure the availability of a To verify the operability of the system with the capability to Standby Liquid Control System. shut down the reactor and maintain

he shutdevn eenditien without the use of control rods.

Srecificatien Seecificatten A. Normal System Availability A. Normal System Availability 1. Except as specified in The operability of the Standby 3.4.5.1, the Standby Liquid Liquid Control System shall be Control System shall be verified by the performance l 0PERABLE at all times when of the following tests: there is fuel in the reactor vessel and the reactor is not 1. Verify pump operability in a shutdown condition with in accordance with Specification 3.3.A.1 Specification 1.0.MM. satisfied, i 2. At least once during each operating cycle: a. Check that the setting of the system relief valves is 1,425 1 75 psig. b. Manually initiate the system, except exple-sive valves. Visually verify flow ty pumping boron solution through the recirculation path and back to the Standby Liquid Centro 1 Solutten Tank. After pumping boren solution, the system shall be flushed with demineralized l water. Verify minimum 3IN 3**1 fc e rd e n * ';o. 107, 125, 12 ,,n i t, ( l

s. 3.4/4.4 STANDBY LIOUID CONTROL SYSTEM LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 4.4.A Nereal System Availability (Cont'd) 4.4.A.2.b. (Cont'd) pump flow rate of 39 spa against a system head of 1275 psis by pumping dominera11 ed water from the j Standby Liquid Control Test Tank. l

c. Manually initiate one of the Standby Liquid Control System loops and pump domineralized water into the reactor vessel.

This test checks explosion of the charge associated with the tested loop, proper operation of the valves, and pump operability. Replacement charges shall be selected such that the age of charge in service shall not exceed five l years from the i i manufacturer's Assembly date. d. Both systems, includins both explosive valves, shall be tested in the course of two operating cycles. BTN 3.4/4.4-2 A end ent '.0. 107, 125 Unit 3 . -}}

v t e Letter Sequence Other
TAC:00250, (Approved, Closed) TAC:00251, (Approved, Closed) TAC:00252, (Approved, Closed)
Initiation Acceptance... Administration Questions Answers Results Approval Other: ML20196B320
MONTHYEAR1988-11-22 [Table View]

ML20196B320

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