ML20035A785
| ML20035A785 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 03/18/1993 |
| From: | Hebdon F Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML18036B204 | List: |
| References | |
| GL-91-08, GL-91-8, NUDOCS 9303290302 | |
| Download: ML20035A785 (21) | |
Text
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UNITED s7ATEs s
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TENNESSEE VALLEY AUTHORITY l
DOCKET NO. 50-259 BROWNS FERRY NUCLEAR PL ANT. UNIT 1 i
AMENDMENT TO FACILITY OPERATING LICENSE I
.I Amendment No.191 i
a License No. DPR-33 i
1.
The Nuclear Regulatory Commission (the Commission) has found that:
l A.
The application for amendment by Tennessee Valley Authority (the I
licensee) dated September 28, 1992, complies with the standards and
(
requirements of the Atomic Energy Act of 1954, as amended (the Act),
j and the Commission's rules and regulations set forth in 10 CFR Chapter I; d
I B.
The f acility will operate in conformity with the application, the j
provisions of the Act, and the rules and regulations of the i
Commission, C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and l
j safety of the public, and (11) that such activities will be conducted in compliance with the Commission's regulations i
i D.
The issuance of this amendment will not be' inimical to the common defense and security or to the health and safety of the public; and l
f E.
The issuance of this amendment is in accordance with 10 CFR Part 51,of the Commission's regulations and all applicable requirements have been l
satisfied.
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ADOCK 050nnpso 3
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Accordingly, the license is amended by changes to the Technical l
t Specifications as indicated in the attachment to this license amendment l
and paragraph 2.C.(2) of Facility Operating License No. DPR-33 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as i
revised through Amendment No.191, are hereby incorporated in the i
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 30 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION I
Frederick J. Hebdon, Dire or i
Project Directorate 11-4 i
Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation 4
Attachment:
Changes to the Technical l
Specifications Date of Issuance: March 18, 1993 j
1 i
i i
i A
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4 ATTACHMENT TO LICENSE AMENDMENT NO.191 FACILITY OPERATING LICENSE NO. DPR-33 l
l DOCKET NO. 50-259 l-Revise the Appendix A Technical Specifications by removing the pages i
i identified below and inserting the enclosed pages. The revised pages are j
identified by the captioned amendment number and contain marginal lines i
indicating the area of change. Overleaf
- and spillover** pages are provided to l
maintain document comoleteness.
i i
REMOVE JNSERJ 3.6/4.6-13 3.6/4.6-13 j].
3.6/4.6-14 3.6/4.6-14 j
3.6/4.6-32 3.6/4.6-32*
i 3.6/4.6-33 3.6/4.6-33 1
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3.6/4.6 PRIM &RY SYSTE4.IDUMr14RY i
a LIMIIING CONDITIONS FOR OPERATION SURVE1LIld4CE REQUIREMENTS j
i 5
3.6.F Recirculation Pump Operation f
3.6.F.3 (Cont'd) the reactor vessel water as j
determined by dome pressure. The l
total' elapsed time in natural circulation and one pump operation must be no greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
3.6.G E1rneturni Integrity 4.6.G Structurni Integrity
)
1.
The structural integrity of ASME 1.
Inservice inspection of ASME Code Class 1, 2, and 3 equivalent Code Class 1, Class 2, and components shall be maintained in Class 3 components shall be l
accordance with Specification 4.6.G performed in accordance with throughout the life of the plant.
Section XI of the ASME Boiler and Pressure Vessel Code and a.
With the structural integrity applicable Addenda as of any ASME Code Class 1 required by 10 CFR 50, equivalent component, which is Section 50.55a(g), except part of the primary system, not where, specific written relief j
conforming to the above has been granted by NRC l
requirements, restore the pursuant to 10 CFR 50, structural integrity of the Section 50.55a(g)(6)(i).
affected component to within i
its limit or maintain the 2.
Additional inspections shall reactor coolant system in be performed on certain either a Cold Shutdown circumferential pipe velds to condition or less than 50*F provide additional protection above the minimum temperaturt agail.st pipe whip, which required by NDT considerations, could damage auxiliary and
)
until each indication of a control systems.
and evaluated.
defect has beer investigated I
1 b.
With the structural integrity of any ASME Code Class 2 or 3 equivalent component not l
conforming to the above i
requirements, restore the structural integrity of the affected component to within its limit or isolate the affected component from all OPERABLE systems.
j i
i BFN 3.6/4.6-13 Amendment 191
)
x Unit 1 3
J p
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N
3 J/4.6 RRTHAPJ SXSTEM ADI,HDMX f
LIMJIING CONDJTIONS FOR OPERATION SURVEILIANCE REOUIREMENTS 4.6.G.
Structural Integrity 3.
For Unit I an augmented inservice surveillance program shall be performed to monitor potential corrosive effects of chloride residue released during the March 22, 1975 fire.
The augmented
)
inservice surveillance program is specified as follows:
l a.
Brown's Ferry Mechanical i
Maintenance Instruction 53, dated i
September 22, 1975, paragraph 4, defines the liquid penetrant examinations required during the first, second, third and fourth refueling outages following the fire
)
restoration.
]
i b.
Browns Ferry Mechanical Maintenance
)
Instruction 46, dated July 18, 1975.
{
Appendix B,. defines the
,~
liquid penetrant examinations required during the sixth refueling outage following the fire restoration.
I J
BFN 3.6/4.6-14 Amendment 191 Unit 1 4
i
.. ~ -
... -.. - - - - - - - - - - - ~ ~ ~ -
~ - - - - - -
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I I
3.6/4.6 BASES i
3.6.E/4.6.E (Cont'd)
I 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 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) l diffuser measurements vill be taken to define the location within the vessel of failed jet pump nozzle (or riser) and the unit shut down for repairs. If the potential blevdown flow area is increased, the system resistance to the recirculation pump is also reduced; hence, the affected 7
drive pump vill "run out" to a substantially higher flow rate (approximately 115 percent to 120 percent for a single nozzle failure).
l If the two loops are balanced in flow at the same pump speed, the resistance characteristics cannot have changed. Any imbalance between I
4 drive 1nop finv raten unuld ha indicated by the plant process instrumentation.
In addition, the affected jet pump would provide a leakage path past the core thus reducing the core flow rate. The reverse flow through the inactive jet pump would still be indicated by a positive
]
differential pressure but the net effect vould be a slight decrease i
i (3 percent to 6 percent) in the total core flow measured. This decrease, together with the loop flow increase, vould result in a lack of correlation between measured and derived core flow ra,te.
Finally, the affected jet pump diffuser differential pressure signal vould be reduced I
because the backflow would be less than the normal forward flow.
A nozzle-riser system failure could also generate the coincident failure of a jet pump diffuser bady; however, the converse is not true. The lack of any substantial stress in the jet pump diffuser body makes failure j
impossible without an initial nozzle-riser system failure.
3.6.F/4.6.F Recirculation pt,p Operation j
l M**ady-etete eperetien vitheut ferced recirculation vill not be permitted i
for mere then 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. And the start of a recirculation pump from the natural circulation condition will not be permitted unless the temperature difference between the loo'p to be started and the core 1
coolant temperature is less than 75"F.
This reduces the positive reactivity insertion to an acceptably lov value.
i Requiring the discharge valve of the lover speed loop to remain closed until the speed of the faster pump is belov 50% of its rated speed provides assurance when going from one-to-two pump operation that excessive vibration of the jet pump risers vill not occur.
l 3.6.G/4.6.C Structural Interrity The requirements for the reactor coolant systems inservice inspection program have been identified by evaluating the need for a sampling examination of areas of high stress and highest probability of failure in i
the system and the need to meet as closely as possible the requirements l
of Section XI, of the ASME Boiler and Precsure Vessel Code.
- 6'**
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3.6/4.6 BAEES 3.6.G/4.6.C (Cont'd) i The program reflects the built-in limitations of access to the reactor l
a 5
coolant systems.
j It is intended that the required examinations and inspection be completed
~
l during each 10-year interval. The periodic examinations are to be done during refueling outages or other extended plant shutdown periods.
j Only proven nondestructive testing techniques will be used.
l 4
i More frequent inspections shall be performed on certain circumferential pipe welds as listed in plant procedures to provide additional protection l
against pipe whip. These welds were selected in respect to their j
distance from hangers or supports wherein a failure of the weld would I
permit the unsupported segments of pipe to strike the drywell wall or l
nearby auxiliary systems or control systems. Ec1c. tion was based on 4
i judgment from actual plant observation of harger e: r support locations and review of drawings.
Inspection of all thet.e welds during each 10-year inspection interval will result in three additional examinations above the requirements of Section XI of ASME Code.
4 An augmented inservice surveillance program is required to determine whether any stress corrosion has occurred in any stainless steel piping,
[
]
stainless components, and highly-stressed alloy steel such as hanger springs, as a result of environmental conditions associated with the March 22, 1975 fire.
REEERENCIS 1.
Inservice Inspection and Testing (BFNP FSAR Subsection 4.12) i 4
2.
Inservice Inspection of Nuclear Reactor Coolant Systems,Section XI, ASME Boiler and Pressure Vessel Code e
3.
ASME Boiler and Pressure Vessel Code, Se,ction III (1968 Edition)
~
4 American Society for Nondestructive Testing No. SNT-TC-1A l
(1968 Edition) a 5.
Mechanical Maintenance Instruction 46 (Mechanical Equipment, i
Concrete, and Structural Steel Cleaning Procedure for Residue From l
1 Plant Fire - Units 1 and 2)
I i
l 6.
Mechanical Maintenance Instruction 53 (Evaluation of Corrosion Damage of Piping Components Which Were Exposed to Residue From March 22, 1975 Fire) 7.
Plant Safety Analysis (BENP FSAR Subsection 4.12) i l
i I
BFN 3.6/4.6-33 Amendment 191 j
Unit 1 5
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- e UNITED STATES y
- v. <[ gg NUCLEAR REGULATORY COMMISSION
- , p,y #.L WASHINGTON, D. C. 20555
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w TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-260 BROWNS FERRY NUCLEAR PLANT. UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE r
i Amendment No. 206 License K,.
DPR-52 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Tennessee Valley Authority (the licensee) dated September 28, 1992, 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 Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Com.iss t un; C.
There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in ccmpliance with the Commission's regulations; D.
The issuance of this amendment will not be' inimical to the common defense and security or to the health and safety of the public; and i
E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been i
satisfied.
1
?
I i
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P e.
' 2.
Accordingly, the license is amended by changes to the Technical i
Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No. DPR-52 is hereby amended to read as follows:
4 (2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.206, are hereby incorporated 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 30 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION MS~
Frederick J. Hebdon, Director Project Directorate 11-4 Division of Reactor Projects - I/II I
Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
March 18. 1993 I
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2 ATTACHMENT TO LICENSE AMENDMENT NO. 206 FACILITY OPERATING LICENSE NO. DPR-52 DOCKET NO. 50-260 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 contain marginal lines inaicating the area of change. Overleaf
- and spillover** pages are provided to maintain document completeness.
REMOVE INSERT s
3.6/4.6-13 3.6/4.6-13 j
l 3.6/4.6-14 3.6/4.6-14 3.6/4.6-32 3.6/4.6-32*
3.6/4.6-33 3.6/4.6-33 i
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i 32b/4.6 RRIMARY SYSTEM SQMND6AX i
LIMITING CONDITIONS FOR OPERATION SURVEILIANCE REOUIREMENTS 3.6.F Recirculation Pump Operation
{
3.6.F.3 (Cont'd) vessel water as determined by dome pressure. The total elapsed time in natural j
circulation and one pump operation must be no greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
i 4.
The reactor shall not be operated with both recirculation pumps out-of-service while the reactor is in the RUN mode. Following a trip of both recirculation pumps while in the RUN mode, immediately l
initiate a manual reactor scram.
I 3.6.G Structural Intecrity 4.6.0 Structural Intecrity t
1.
The structural integrity of 1.
Inservice inspection of ASME ASME Code Class 1, 2, and Code Class 1, Class 2, and 3 equivalent components shall.
Class 3 components shall be l
be maintained in meenrdance performed in accordance with with Specification 4.6.G Section XI of the ASME Boiler throughout the life of the and Pressure Vessel Code and plant applicable Addenda as required by 10 CFR 50, Section 50.55a(g),
a.
With the structural except where specific written integrity af any ARMF relief has been granted by NRC Code Clace 1 equivalent pursuant to 10 CFR 50, Section component, which is part 50.55a(g)(6)(i).
of the primary system,
~
not conforming to the j
above requirements, restore 2.
Additional inspections the structural integrity of shall be performed on the affected component to certain circumferential within its limit or maintain pipe welds to provide the reactor coolant system in additional protection either a COLD SHUTDOWN against pipe whip, CONDITION or less than 50*F which could damage above the minimum temperature auxiliary and control required by NDT consider-systems.
ations, until each indication j
of a defect has been inves-tige'.ed and evaluated.
l l
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BFN 3.6/4.6-13 Amendment 206 Unit 2
i 13/4.6 PRIMARY SYSTEM EDUhTRJ lit!ITING CONDITIONS FOR OPERATION S E Ell,I/SCE REOUIREMENTS j
i 3.6.G Structurni Integrity j
3.6.G.1 (Cont'd) t b.
With the structural integrity of any ASME Code Class 2 or 3 equivalent component not conforming to the above l
requirements, restore the i
structural integrity of tne affected component to within its limit or isolate the i
affected component from all OPERABLE systems.
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-Amendment 206 i
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3.6/4.6 BASES I
i j
3.6.E/4.6.E (Cont'd) i
]
If they do differ by 10 percent or more, the core flow rate measured by i
j the jet pump diffuser differential pressure system must be checked i
j against the core flow rate derived frem the measured values of loop flov i
to core flow correlation. If the difference between measured and derived j
core flow rate is 10 percent or more (with the derived value higher)
(
j diffuser measurements vill be taken to define the location within the j
vessel of failed jet pump nozzle (or riser) and the unit shut down for repairs.
If the potential blowdown flow area is increased, the system 3
resistance to the recirculation pump is also reduced; hence, the affected l
drive pump will "run out" to a substantially higher flow rate
]
(approximately 115 percent to 120 percent for a single nozzle failure).
j If the two loops are balanced in flow at the same pump speed, the i
resistance characteristics cannot have changed. Any imbalance between j
- iri : Ice; f1;; retca weuld bc
- diu Led by the plant process u
i instrumentation. In addition, the affected jet pump would provide a I
j leakage path past the core thus reducing the core flow rate. The reverse
{
flow through the inactive jet pump would still be indicated by a positive j
differential pressure but the net effect would be a slight decrease
{
(3 percent to 6 percent) in the total core flow measured. This decrease, j
together with the loop flow Jncrease, would result in a lack of correlation between measured and derived core flow rate. Finally, the l
affected jet pump diffuser differential pressure signal vould be reduced i
j because the backflow would be less than the normal forward flow.
f i
A nozzle-riser system failure could also generate the coincident failure l
1 cf a jet pump diffue,er hody; however, the converse is not true. The lack 3
of any substantial stress in the jet pump diffuser body makes failure i
f impossible without an initial nozzle-riser system failure.
J t
l 3.6.F/4.6.F Recirculation Pump Operation l
1
{
Opu6 tion withuut ivreed recirculation is permitted for up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is not in the RUN mode. And the start of a recirculation pump from the natural circulation condition vill not be permitted unless the temperature difference between the loop to be f
j started and the core coolant temperature is less than 75'F.
This reduces-the positive reactivity insertion to an acceptably lov value.
l s
l Requiring at least one recirculation pump to be operable while in the RUN i
mode (i.e., requiring a manual scram if both recirculation pumps are i
i tripped) provides protection against the potential occurrence of core l
thermal-hydraulic instabilities at low flow conditions.
I t
Requiring the discharge valve of the lower speed loop to remain closed
{
until the speed of the faster pump is belov 50% of its rated speed 4
provides assurance when going from one-to-tvo pump operation that
)
excessive vibration of the jet pump risers vill not occur.
I l
1 i
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t BFN 3.6/4.6-32 AMUlDMER fi0. I 9 8 i
Unit 2 i
1 1
I i
1 l
3 1
3.6/4.6 BASES l
1 3.6.G/4.6.G Structural Integrity
~
0 The requirements for the reactor coolant systems inservice inspection program have been identified by evaluating the need for a sampling i
examination of areas of high stress and highest probability of failure in l
the system and the need to meet as closely as possible the requirements
~
]
of Section XI, of the ASME Boiler and Pressure Vessel Code.
1 The program reflects the built-in limitations of access to the reactor coolant systems.
It is intended that the required examinations and inspection be completed during each 10-year interval. The periodic examinations are to be done during refueling outages or other extended plant shutdown periods.
t 4
I Only proven nondestructive testing techniques will be used.
More frequent inspections shall be performed on certain circumferential pipe welds as listed in plant procedures to provide additional protection l against pipe whip. These welds were selected in respect to their J
distance from hangers or supports wherein a failure of the weld would i
permi;. the unsupported segments of pipe to strike the drywell vall or j
nearby auxiliary systems or control systems. Selection was based on judgment f rom actual plant observation of hanger and support locations 1
and review of drawings.
Inspection of all these. welds during each 10-year inspection interval will result in three additional examinations above the requirements of Section XI of ASME Code.
1 An augmented inservice surveillance program is required to determine l
i whether any stress corrosion has occurred in any stainless steel piping, l
j stainless components, and highly-stressed alloy steel such as hanger
[
springs, as a result of environmental conditions associated with the
]
March 22,1975 fire.
9 1
REFERENCES l
1.
Inservice Inspection and Testing _(BFNP FSAR Subsection 4.12) 2.
Inservice Inspection of Nuclear Reactor Coolant Systems,Section XI, ASME Boiler and Pressure Vessel Code j
3.
ASME Boiler and Pressure Vessel Code,Section III (1968 Edition) i 4.
American Society for Nondestructive Testing No. SNT-TC-1A
[
4 (1968 Edition) l s
i 5.
Mechanical Maintenance Instruction 46 (Mechanical Equipment, Concrete, and Structural Steel Cleaning Procedure for Residue From Plant Fire - Units 1 and 2) d 6.
Mechanical Maintenance Instruction 53 (Evaluation of Corrosion Damage of Piping Components Which Were Exposed to Residue From March 22, i
1975 Fire) i 2
7.
Plant Saf ety Analysis (BFNP FSAR Subsection 4.12) i
]
4 i
BFN 3.6/4.6-33 Amendment 206 i
t
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Unit 2 e
j.. me r-4
,o, Ult 3iTED STATES l
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- 3 W 'i NUCLEAR REGULATORY COMMISSION k.
E W ASHINGTON. D. C. 20555 i
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i TENNESSEE VALLEY AUTHORIT1 DOCKET NO. 50-296 BROWNS FERRY NUCLEAR PLANT. UNIT 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 163 i
License No. DPR-68 i
1.
The Nuclear Regulatory Commission (the Commission) has found that:
l A.
The application for amendment by Tennessee Valley Authority (the licensee) dated September 28, 1992, complies with the standards and i
requirements of the Atomic Energy Act of 1954, as amended (the Act),
and the Commission's rules and regulations set forth in 10 CFR Chapter I; j
B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by l
this amendment can be conducted without endangering the health and t
safety of the public, and (ii) that such activities will.be conducted in compliance with the Commission's regulations, D.
The issuance of this amendment will not be inimical to the common r
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 Ccomission's regulations and all applicable requirements have been satisfied.
t e
r l
i
. 2.
Accordingly, the license is amended by changes to the Technical t
Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of facility Operating License No. DPR-68 is hereby amended to read as follows:
(2) Technical Specifications l
The Technical Specifications contained in Appendices A and B, as revised through Amendment No.163, are hereby incorporated 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 30 days from the date of issuance.
t FOR THE NUCLEAR REGULATORY COMMISSION M
Frederick J. Hebdon, Direct r Project Directorate 11-4 Division of Reactor Projects - 1/II Office of Nuclear Reactor Regulation i
Attachment:
Changes to the Technical Specifications nate of Iccuanca-March 19, 1993 l
i 9
l k
I v
1
3 l
i ATTACHMENT TO LICENSE AMENDMENT NO.163 r
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 contain marginal lines 1
indicating the area of change. Overleaf
- and spillover** pages are provided to maintain document completeness.
i REMOVE INSERT 3.6/4.6-13 3.6/4.6-13 3.6/4.6-14 3.6/4.6-14 3.6/4.6-32 3.6/4.6-32*
4 3.6/4.6-33 3.6/4.6-33 t
i A
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1 l
.3.o/=.0 l'RIMARY SXSIE1392hAhl i
LIMITING CONDITIONS FOR OPERATION SURVEILIANCE REOUIREMENTS 3.6.F Recirculation Pump Operation 3.6.F.3 (Cont'd) i f
the reactor vessel water as determined by dome pressure. The total elapsed time in natural circulation and one pump operation must be no greater l
than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
i 3.6.G Structural Intrgrity 4.6.G Structurni Intecrity f
1.
The structural integrity of ASME 1.
Inservic'e inspection of ASME cooe Class 1, i, ano 3 equivalent Code Class 1, Class 2, and j
components shall be maintained Class 3 components shall be i
in accordance with Specification performed in accordance with 4.6.G throughout the life of the Section XI of the ASME Boiler plant.
and Pressure Vessel Code and applicable Addenda as required a.
With the structural integrity by 10 CFR 50 Section 50.55a(g),
of any ASME Code Class 1 except where specific written
~
equivalent component, which relief has been_ granted by NRC is part of the primary system, pursuant to 10 CFR 50, Section not conforming to the above 50.55a(g)(6)(1).
i requirements, restore the j
structural integrity of the 2.
Additional inspections shall be i
affected component to within performed on certain i
~
its limit or maintain the circumferential pipe welds reactor coolant system in either to provide additional a Cold Shutdown conditio::
protection against pipe whip, er less than 50*F above which could damage auxiliary j
i the minimum temperature and control systems.
required by NDT consider-ations, until each indication l
~"
of a defect has been i
investigated and evaluated.
b.
With the structural integrity i
of any ASME Code Class 2 or 3 equivalent component not conforming to the above requirements, restore the structural integrity of the affected component to within its limit or isolate the affected component from all OPERABLE systems.
}
i
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1 1
BFN 3.6/4.6-13 Amendment 163 I
Unit 3 i
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BFN 3.6/4.6-14 Amendment 163 l
Unit 3 t
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3.6/4.6 BASES 1
3.6.E/4.6.E (Cont'd) 3 t
4 l
area is increased, the system resistance to the recirculaticn pump is also j
reduced; hence, the affected drive pump vill "run out" to a substantially l
higher flow rate (approximately 115 percent to 120 percent for a single nozzle 1
3 failure). If the two loops are balanced in flow at the same pump speed, the l
resistance characteristics cannot have changed. Any imbalance between drive j
loop flow rates would be indicated by the plant process instrumentation.
In j
addition, the affected jet pump would provide a leakage path past the core t
j thus reducing the core flow rate. The reverse flow through the inactive jet pump would still be indicated by a positive differential pressure but the net i
effect would be a slight decrease (3 percent to 6 percent) in the total core j
flow measured. This decrease, together with the loop flov increase, would l
result in a lack of correlation between measured and derived core flow rate.
l Finally, the affected jet pump diffuser differential pressure signal vould be l
reduced becevee the bichfler veuld be 1e:: than the normal forward flow.
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A nozzle-riser system failure could also generate the coincident failure of a j
jet pump diffuser body; however, the converse is not true. The lack of any l
substantial stress in the jet pump diffuser body makes failure impossible j
vithout an initial nozzle-riser system failure.
3 i
j 3.6.F/4.6.F Recirculation Pump Operation i
l l
Steady-state operation without forced recirculation vill not be permitted for 4
more than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. And the start of a recirculation pump from the natural j
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circulation condition vill not be permitted unless the temperature difference j
t between the 1:0; :: bc started and the core coolant temperature is less than j
i 75*F.
This reduces the positive reactivity insertien to an acceptably lov value.
1 Requiring the discharge valve of the lover speed loop to remain closed until the speed of the faster pump is belov 50 percent of its rated speed provides tecurene: Jhen going from enc-t;-tv; pump cperation that excessive vibration
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cf the jet pucp tisers will not occur.
J 3.6.G/4.6.G Structural Interrity l
1 1
The requirements for the reactor coolant systems inservice inspection program 1
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have been identified by evaluating the need for a sampling examinatien of
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j areas of high stress and highest probability of failure in the system and the i
need to meet as closely as possible the requirements of Section XI, of the j
ASME Boiler and Pressure Vessel Code.
I i
The program reflects the built-in limitations of access to the reactor coolant i
systems.
1 i
1 l
It is intended that the required examinations and inspection be completed i
i during each 10-year interval. The periodic examinations are to be done during I
refueling outages or other extended plant shutdown periods.
I I
l BFN 3.6/4.6-32 AMENDMEKT NO.15 2 d
Unit 3 i
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i b.o/4.6 7dSLS 4
1 l,
3.6.C/4.6.G (Cont'd) t I
Only proven nondestructive testing techniques will be used.
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l
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More frequent inspections shall be performed on certain circumferential pipe welds as listed in plant procedures to provide additional protection against l
j l
pipe whip. These welds were selected in respect to their distance from j
j hangers or supports wherein a failure of the veld would permit the unsupported segments of pipe to strike the dryvell wall or nearby auxiliary systems or l
l control systems. Selection was based on judgment from actual plant
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j observation of hanger and support locations and review of drawings.
}
J Inspection of all these velds during each 10-year inspection interval will
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j result in three additional examinations above the requirements of Section XI of ASME Code.
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i References 1.
Inservice Inspection and Testing (BFNP FSAR Subsection 4.12)
_i 1
j 2.
Inservice Inspection of Nuclear Reactor Coolant Systems,Section XI, ASME l
Boiler and Pressure Vessel Code 3.
ASME Boiler and Pressure Vessel Code, Section 111 (1968 Edition)
I i
j 4.
American Society for Nondestructive Testing No. SNT-TC-1A (1968 Edition) l i
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4 Ii 1
3 8
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4 3
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lj ETN 3.6/4.6-33 Amendment 163 Unit 3 j
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