ML18036A879
| ML18036A879 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 09/28/1992 |
| From: | TENNESSEE VALLEY AUTHORITY |
| To: | |
| Shared Package | |
| ML18036A878 | List: |
| References | |
| NUDOCS 9210070120 | |
| Download: ML18036A879 (31) | |
Text
ENCLOSURE 1
PROPOSED TECHNICAL SPECIFICATION CHANGE BROWNS PERRY NUCLEAR PLANT TVA BFN TS-330 92i0070i20 920928 PDR ADOCK 05000259 P,..
PDR.
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LIST OF EFFECTIVE PAGES UNIT 1 3.6/4.6-13 3.6/4.6.-14 3.6/4.6-33 BFN TS-330 UNIT 2 3.6/4.6-13 3.6/4.6-14 3.6/4.6-33 UNIT 3 3.6/4.6-13 3.6/4.6-14 3.6/4.6-33
J.J
3.6.F 3.6.F.3 (Cont'd) the reactor vessel water as determined by dome pressure.
The 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 t
4.6.G The structural integrity of ASHE Code Class 1, 2, and 3 equivalent components shall be maintained in accordance with Specification 4.6.G throughout the life of the plant.
- a. 'ith the structural integrity of any ASHE Code Class 1
equivalent component, which is part of the primary system, not conforming to the above requirements, restore the structural integrity of the affected component to within its limit or maintain the reactor coolant system in either a Cold Shutdown condition or less than 50'F I
above the minimum temperature required by NDT considerations, until each indication of a defect has been investigated and evaluated.
2.
Inservice inspection of ASHE Code Class 1, Class 2, and Class 3 components shall be performed in accordance with Section XI of the ASHE 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 NRC pursuant to 10 CFR 50, Section 50.55a(g)(6)(i)-
Additional inspections shall be performed on certain circumferential pipe welds to provide additional protection against pipe whip, which could damage auxiliary and control systems.
b.
With the structural integrity of any ASHE 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.
BFN Unit 1 3.6/4.6-13
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3.
For Unit 1 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:
a.
Browns Ferry Mechanical Maintenance Instruction 53, dated September 22,
- 1975, paragraph 4, defines the liquid penetrant examinations required during the first, second, third and fourth refueling outages following the fire restoration.
b.
Browns Ferry Mechanical Maintenance Instruction 46, dated July 18, 1975.
Appendix B, defines the liquid penetrant examinations required during the sixth refueling outage
'ollowing the fire restoration.
BFN Unit 1
3.6/4.6-14
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3.6/4.6 ~5 3.6.G/4.6.G (Cont'd)
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.
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 against pipe whip.
These welds were selected in respect to their distance from hangers or supports wherein a failure of the weld would permit the unsupported segments of pipe to strike the drywell wall or nearby auxiliary systems or control systems.
Selection was based on judgment from actual plant observation of hanger and support locations-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 ASHE Code.
An augmented inservice surveillance program is required to determine whether any stress corrosion'as 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.
1.
Inservice Inspection and Testing (BFNP FSAR Subsection 4.12) 2.
Inservice Inspection of Nuclear Reactor Coolant Systems,Section XI, ASHE Boiler and Pressure Vessel Code 3.-
ASME Boiler.and Pressure Vessel Code,Section III (1968 Edition) 4.
American Society for Nondestructive Testing No. SNT-TC-lA (1968 Edition) 5 ~
Mechanical Maintenance Instruction 46 (Mechanical Equipment,
- Concrete, and Structural Steel Cleaning Procedure for Residue From Plant Fire Units 1 and 2) 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 (BFNP FSAR Subsection 4.12)
(
I BFN Unit 1 3.6/4.6-33
3.6.F 3.6.F.3 (Cont'd) vessel water as determined by dome pressure.
The'otal 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 />.
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
-~..= initiate a manual reactor scram.
3.6.G 4.6.G a.
With the structural integrity of any ASME Code Class 1 equivalent component, which is part of the primary system, not conforming to the above requirements, restore the structural integrity of the affected component to within its limit or maintain the reactor coolant system in either a
COLD SHUTDOWN CONDITION or less than 50'F above the minimum temperature required by NDT consider-ations, until each indication of a defect has been inves-tigated and evaluated.
The structural integrity of ASNE Code Class 1, 2, and 3 equivalent components shall be maintained in accordance with Specification 4.6.G throughout the life of the plant.
2.
Inservice inspection of ASHE Code Class 1, Class 2, and Class 3 components shall be performed in accordance with Section XI of the ASNE 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 NRC pursuant to 10 CFR 50, Section 50.55a(g)(6)(i).
Additional inspections shall be performed on certain circumferential pipe welds to provide additional protection against pipe whip, which could damage auxiliary and control systems.
BFN Unit 2 3.6/4.6-13
3.6.G 3.6.G.l (Cont'd) b.
With the structural integrity of any ASNE 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.
BFN Unit 2 3.6/4.6-14
~t 1 =l J
- 3. 6/4. 6 MQES 3.6;G/4.6.G The requirements for the reactor coolant systems inseryice inspection program have been identified by evaluating the need for a sampling examination of areas of high stress and highest probability of failure in the systems and the need to meet as closely as possible. the requirements of Section XI, of the ASNE Boiler and Pressure
-Vessel Code.
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.
Only proven nondestructive testing techniques will be used.
Nore frequent inspections shall be performed on certain circumferential pipe welds as listed in plant procedures to provide additional protection
}
against pipe whip.
These welds were selected in respect to their distance from hangers or supports wherein a failure of the weld would permit the unsupported segments of pipe to strike the drywell wall or nearby auxiliary systems or control systems.
Selection was based on judgment from actual plant observation of hanger and support locations 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 ASNE Code.
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.
-l. -'Inservice Inspection and Testing (BFNP FSAR Subsection 4.12) 2.
Inservice Inspection of Nuclear Reactor Coolant Systems,Section XI, ASNE Boiler and Pressure Vessel Code 3.
ASNE Boiler and Pressure Vessel Code,Section III (1968 Edition) 4.
American Society for Nondestructive Testing No. SNT-TC-lA (1968 Edition) 5.
Mechanical Maintenance Instruction 46 (Nechanical Equipment,
- Concrete, and Structural Steel Cleaning Procedure for Residue From Plant Fire Units 1 and 2) 6.
Mechanical Naintenance Instructioq 53 (Evaluation of Corrosion Damage of Piping Components Which Were Exposed to Residue From Narch 22, 1975 Fire) 7.
Plant Safety Analysis (BFNP FSAR Subsection 4.12)
BFN Unit 2 3.6/4.6-33
3.6.F 3.6.F.3 (Cont'd) the reactor vessel water as determined by dome pressure.
The 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 t
t ~t 4.6.G 1.
The structural integrity of ASHE Code Class 1, 2, and 3 equivalent components shall be maintained in accordance with Specification 4.6.G throughout the life of the
. plant.
With the structural integrity of any ASNE Code Class 1
equivalent component, which is part of the primary system, not conforming to the above requirements, restore the structural integrity of the affected component to within its limit or maintain the reactor coolant system in either a Cold Shutdown condition or less than 50'F above the minimum temperature required by NDT consider-ations, until each indication of a defect has been investigated and evaluated.
l.
Inservice inspection of ASHE Code Class 1, Class 2, and Class 3 components 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 NRC pursuant to 10 CFR 50, Section 50.55a(g)(6)(i).
2.
Additional inspections shall be performed on certain circumferential pipe welds to provide additional protection against pipe whip, which could damage auxiliary and control systems.
b.
Wi,th the structural integrity of any ASNE 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.
BFN Unit 3 3.6/4.6-13
THIS PAGE INTENTIONALLYLEFT BLANK BFN Unit 3 3.6/4.6-14
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>C,
3.6/4.6 MQE5 3.6.G/4.6.G (Cont'd)
Only proven nondestructive testing techniques will be used.
Nore frequent inspections shall be performed on certain circumferential pipe welds as listed-in plant procedures to provide additional protection against pipe whip.
These welds were selected in respect to their distance from hangers or supports wherein a failure of the weld would permit the unsupported segments of pipe to strike the drywell wall or nearby auxiliary systems or control systems.
Selection was based on judgment from actual plant observation of hanger and support locations 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 ASHE Code.
Inservice Inspection and Testing '(BFNP FSAR Subsection 4.12) 2 ~
Inservice Inspection of Nuclear Reactor Coolant Systems,Section XI, ASHE
- Boiler -and Pressure Vessel Code 3.
4.
ASHE Boiler and Pressure Vessel
- Code,Section III (1968 Edition)
I American Society for Nondestructive Testing No. SNT-TC-1A (1968 Edition)
BFN Unit 3 3.6/4.6-33
J
ENCLOSURE 2 BROWNS FERRY NUCLEAR PLANT DESCRZPT10N AND JUSTZFZCATZON FOR THE PROPOSED CHANGES Summar
" of Chan es for Units 1
2 and 3
Present technical specification surveillance requirement 4.6.G.2 reads as follows for BFN Unit 1:
"Additional inspections shall be performed on certainn circumferential pipe welds as listed to provide additional protection against pipe whip, which could damage auxiliary and control systems.
Feedwater GFW-9, GFW-12, KFW-31, KFW-39, KFW-38, KFW-13 GFW-26, GFW-29, GFW-15, and GFW-32 Main steam GMS-6, KMS-24, GMS-32, KMS-104 GMS-15, and GMS-24 DSRHR-4, DSRHR-7, DSRHR-8A Core Spray Reactor Cleanup HPCI DSCS 12 I DSCS 11 I DSCS-5, and DSCS-4 DSRWC-4, DSRWC-3 DSRWC-6, DSRWC-5 THPCI
152 THPCI
153B THPCI
153 THPCI
154" Present technical specification surveillance requirement 4.6.G.2 reads as follows for BFN Unit 2:
"Additional inspections shall be performed on certain circumferential pipe welds as listed to provide additional protection against pipe whip, which could damage auxiliary and control systems.
ENCLOSURE 2
BROWNS FERRY NUCLEAR PLANT DESCRIPTION AND JUSTIFICATION FOR THE PROPOSED CHANGES Summar of Chan es for Units 1
2 and 3
(Continued)
Feedwater GFW-9, GFW-12, KFW-31, KFW-39, KFW-38, KFW-13 GFW-26, GFW-29, GFW-15, and GFW-32 Main steam GMS-6, KMS-24, GMS-32, KMS-104 GMS-15 and GMS-24 DSRHR-4, DSRHR-7, DSRHR-6 Core Spray TSC-407, TSC-423, TSCS-408, and TSC-424 Reactor Cleanup HPCI DSRWC-4, DSRWC-3 DSRWC-6, DSRWC-5 THPCI
70 THPCI
70A THPCI
71 THPCI 72" Present technical specification surveillance requirement 4.6.G.2 reads as follows for BFN Unit 3:
"Additional inspections shall be performed on certainn circumferential pipe welds as listed to provide additional protection against pipe whip, which could damage auxiliary and control systems.
Feedwater Main steam GFW-9, GFW-12, KFW-31, KFW-39, KFW-38, GMS-6, GMS-32, GMS-15, KFW-13 GFW-26, GFW-29, GFW-15, and GFW-32 KMS-24, KMS-104 and GMS-24
..vi
ENCLOSURE 2
BROWNS FERRY NUCLEAR PLANT DESCRIPTION AND JUSTIFICATION FOR THE PROPOSED CHANGES Summar of Chan es for Units 1
2 and 3
(Continued)
DSRHR-6, DSRHR-7, DSRHR-4 Core Spray TSC-407, TSC-423, TSCS-408, and TSC-424 Reactor Cleanup HPCI DSRWC 4 g DSRWC 3
DSRWC 6 g DSRWC 5 THPCI
70 THPCI
70A THPCI 71, and THPCI
72 REFERENCE 1.
Plant Safety Analysis (BFN FSAR subsection 4.12)"
2.
Proposed technical specification surveillance requirement 4.6.G.2 for BFN Units 1, 2, and 3 is as follows:
"Additional inspections shall be performed on certain circumferential pipe welds to provide additional protection against pipe whip, which could damage auxiliary and control systems."
3.
Present Bases 3.6.G/4.6.G reads in part on page 3.6/4.6-33:
"More frequent inspections shall be performed on certain circumferential pipe welds as listed in Section 4.6.G.4 to provide additional protection against pipe whip."
4.
Proposed Bases 3.6.G/4.6.G will read in part on page 3.6/4.6-33:
"More frequent inspections shall be performed on certain circumferential pipe welds as listed in plant procedures to provide additional protection against pipe whip."
mba
ENCLOSURE 2
BROGANS FERRY NUCLEAR PLANT DESCRIPTION AND JUSTIFICATION FOR THE PROPOSED CHANGES Reason for the Chan es Piping replacement in the Reactor Water Cleanup System on BFN Unit 3 due to pipe cracking willnecessitate a revision to the list of welds in Surveillance Requirement 4.6.G.2.
Instead of submitting individual technical specification amendment requests for each weld number that may change in the future, it is more appropriate to delete the listing of pipe welds from the technical specifications
~ following the guidance of Generic Letter 91-08.
Justification for the Pro osed Chan es Generic Letter (GL) 91-08 provides guidance for preparing a request for a license amendment to remove component lists from technical specifications (TS).
The proposed removal of the list of certain circumferential pipe welds in TS Surveillance Requirement. 4.6.G.2 meets the guidance in this GL.
GL 91-08 provides the following guidance for preparation of a request to remove component lists from the TS:
1.
Each TS should include an appropriate description of the scope of the components to which the TS requirements apply.
'omponents that are defined by regulatory requirements or guidance need not be clarified further.
- However, the Bases Section of the TS should reference the applicable requirements or guidance.
- 2. If the removal of a component list results in the loss of notes that modify or provide an exception to the TS requirements, the specification should be revised to incorporate that modification or exception.
The modification or exception should be stated in terms that identify any group of components by function rather than by plant identification number.
3.
Licensees should confirm that the lists of components removed from the TS are located in appropriately controlled plant
.procedures.
The list of components may be included in the next update of the FSAR.
The Bases Section of individual specifications also may reference the plant procedures or other documents that identify each component list.
The Bases Section for the containment isolation valve TS should be updated to describe the intent of opening valves under administrative control.
ENCLOSURE 2
BROWNS FERRY NUCLEAR PLANT I
DESCRIPTION AND JUSTIFICATION FOR THE PROPOSED CHANGES Justification for the Pro osed Chan es (Continued)
For this proposed amendment
- request, the issues from GL 91-08 are addressed as follows:
1.
TS Surveillance Requirement 4.6.G.2 includes a description of the pipe welds that are being deleted from the TS.
The description states that additional inspections shall be performed on certain circumferential pipe welds to provide additional protection against pipe whip, which could damage auxiliary and control systems.
TS Bases 3.6.G/4.6.G further describes "these welds by stating that they were selected in respect to their distance from hangers or supports wherein a
failure of the weld would permit the unsupported segments of pipe to strike the drywell wall or nearby auxiliary systems or.
control systems.
The Bases goes on to state that selection of the'elds was based on judgment from actual plant observation of hanger and support locations and review of drawings.
Inspection of all these welds during each 10-year inspection interval willresult in three additional examinations above. the requirements of Section XI of ASME Code.
2.
The removal of the list of pipe welds from TS Surveillance Requirement 4.6.G.2 does not result in the loss of notes that modify or provide an exception to the TS requirements.
3.
The list of pipe welds in TS Surveillance Requirement 4.6.G.2 is located in plant procedures and reference to these procedures is proposed to be added to Bases 3.6.G/4.6.G.
This proposed TS amendment request meets the guidance provided in GL 91-08 for removal of component lists.
The removal of the list of pipe welds is acceptable because it does not alter existing TS requirements or those components to which they apply.
TS Surveillance Requirement 4.6.G.2 for BFN Unit 3
presently contains a Reference 1 for the plant Safety Analysis (BFN FSAR subsection 4.12).
This reference is being deleted using the guidance of GL 91-08.
Retention of this reference in the TS could imply that a technical specification amendment is required to change Section 4.12 of the BFN FSAR.
ENCLOSURE 3
BROWNS FERRY NUCLEAR PLANT (BFN)
PROPOSED DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION Descri tion of Pro osed Technical S ecification Amendment The proposed technical specification change applies to BFN Units 1, 2 and 3.
The proposed change will delete the list of pipe welds contained in TS Surveillance Requirement 4.6.G.2.
Appropriate changes to Bases 3.6.G/4.6.G are proposed to reflect the deletion of the pipe weld list.
The pipe weld inspection requirement is included in the technical specifications to provide additional protection against pipe whip, which could damage auxiliary and control systems.
The pipe welds inspected are located on the Feedwater, Main Steam, Residual Heat Removal, Core Spray, Reactor Water Cleanup, and High Pressure Coolant Injection Systems.
Basis For Pro osed No Si nificant Hazards Consideration Determination The NRC has provided standards for determining whether a
significant hazards consideration exists as stated in 10 CFR 50.92(c).
A proposed amendment to an operating license involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not 1) involve a
significant increase in the probability or consequences of an accident previously evaluated, or 2) create the possibility of a new or different kind of accident from any accident previously evaluated, or 3) involve a significant reduction in a margin of safety.
1.
The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
The deletion of the list of pipe welds from TS Surveillance Requirement 4.6.G.2 does not remove or alter the requirement to inspect the welds nor does it change the description of the pip'e welds to be inspected.
The list of pipe welds to be inspected are contained in plant procedures which are subject to the change control provisions in the administrative controls section (Chapter
- 6) of the TS.
The procedure change control provisions of Chapter 6 of the TS provide an adequate means to control changes to this list of pipe welds without having to process a license amendment.
Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.
~I
ENCLOSURE 3
BROWNS FERRY NUCLEAR PLANT (BFN)
PROPOSED DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION
'I Basis For Pro osed No Si nificant Hazards Consideration Determination (Continued) 2.
The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
The proposed change follows the guidance provided in GL 91-08 to allow removal of component lists from the technical specifications.
The proposed change does not modify the list of pipe welds currently in the TS, nor is the frequency of
.inspection or method of inspection changed by this proposed TS change.
The list of pipe welds currently in the TS is relocated to plant procedures which are under the change control process of Chapter 6 of the TS.
Therefore, the proposed change does not create the possibility or a
new or different kind of accident from any previously evaluated.
3.
The proposed change does not involve a significant reduction in a margin of safety.
The present Surveillance Requirements in TS 4.6.G.2 have not been modified by the proposed deletion of the list of pipe welds contained in that specification.
The deletion of this list of welds from the TS will prevent the necessity to submit an amendment request each time the weld list needs to be revised.
The list of pipe welds is currently in plant procedures.
These procedures are required by Chapter 6 of the TS and changes to them are controlled by TS provisions.
Therefore, the proposed changes do not reduce any margin of safety.