ML18094A092
| ML18094A092 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 12/09/1988 |
| From: | TENNESSEE VALLEY AUTHORITY |
| To: | |
| Shared Package | |
| ML18033A471 | List: |
| References | |
| GL-88-01, GL-88-1, TAC-R00524, TAC-R00525, TAC-R00526, TAC-R524, TAC-R525, TAC-R526, NUDOCS 8812140159 | |
| Download: ML18094A092 (17) | |
Text
ENCLOSURE 1
BROWNS FERRY NUCLEAR PLANT REVISED TECHNICAL SPECIFICATION PAGES TECHNICAL SPECIFICATION NO. 262 UNITS 1, 2, AND 3 8812140159 SSi209 PDR ADOCK 05000259 P
MM. Survei lance e uirements for ASME Section XI Pum and Valve
~Pro ram Surveillance requirements for inservice 1estinq 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 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 ASME Boiler and Pressure Vessel Code and applicable Addenda shall be applicable as follows in these technical specifications:
ASME Boiler and Pressure Vessel Code and applicable Addenda terminology for inservice testin activities Required frequencies for performing inservice testin activities Weekly Monthly Quarterly or every 3 months Semiannually or every 6 months Every 9 months Yearly or annually At least once per 7 days At least once per 31 days At least once per 92 days At least once per 184 days At least once per 276 days 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 Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any technical specification.
6.
The inservice inspection program for piping identified in NRC Generic Letter 88-01 shall be performed in accordance with the staff positions on schedule,
- methods, personnel, and sample expansion included in this generic letter.
BFN Unit 1 1.0-12
MM.
S ve a ce e
erne ts o
S Sec o
Pum and Valve
~Pro ra Surveillance requirements for Inservice Zesting 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 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 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
, testin activities testin activit es Meekly Monthly Quarterly or every 3 months Semiannually or every 6 months Every 9 months Yearly or annually At least once per 7 days At least once per 31 days At least once per 92 days At least once per 184 days At least once per 276 days 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 Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any technical specification.
6.
The inservice inspection program for piping identified in NRC Generic Letter 88-01 shall be performed in accordance with the staff positions on schedule,
- methods, personnel, and sample expansion included in this generic letter.
BFN Unit 2 1.0-12
MM. Surveillance re uirements for ASME Section XI Pum and Valve
~Pro em gurveillenee requirements for lnservioe 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 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 ASME Boiler and Pressure Vessel Code and applicable Addenda shall be applicable as follows in these technical specifications:
ASME Boiler and Pressure Vessel Code and applicable Addenda terminology for inservice testin activit es Required frequencies for performing inservice testin activities Weekly Monthly Quarterly or every 3 months Semiannually or every 6 months Every 9 months Yearly or annually At least once per 7 days At least once per 31 days At least once per 92 days At least once per 184 days At least once per 276 days 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 Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any technical specification.
6.
The inservice inspection program for piping identified in NRC Generic Letter 88-01 shall be performed in accordance with the staff positions on schedule,
- methods, personnel, and sample expansion included in this generic letter.
BFN Unit 3
ENCLOSURE 2
DESCRIPTION AND JUSTIFICATION BROWNS FERRY NUCLEAR PLANT (BFN)
REASON FOR CHANGE The reason for this change is to amend the BFN Units 1, 2, and 3 Technical Specifications (TS) by adding item 1.0.MM.6 to the ASME Section XI Pump and Valve Program located in the Definitions Section.
This change will bring the BFN TS in compliance with the requirements specified in NRC Generic Letter 88-01, Intergranular Stress Corrosion Cracking (IGSCC).
NRC Generic Letter 88-01, dated January 25,1988 provided Licensee's with NRC's position on IGSCC for BWR Austenitic Stainless Steel Piping.
Early cases of IGSCC were observed in relatively small diameter piping up until 1982 when cracking was identified in a recirculation system at one BWR plant.
As a
- result, extensive inspection programs were conducted on BWR piping systems.
Substantial efforts in research and development have been sponsored by BWR Owners Groups for IGSCC.
The results of these efforts, along with other related work by vendors, consulting firms, and confirmatory research sponsored by NRC, have resulted in the Staff's Positions regarding the IGSCC problems.
The technical bases for these positions are detailed in NUREG-0313, rev. 2, dated January 1988 "Technical Report on Material Selection and Process Guidelines for BWR Coolant Pressure Boundary Piping".
The subject NUREG provides staff positions on materials, processes, inspection programs, technical specifications, and on primary coolant chemistry to minimize and control IGSCC.
Generic Letter 88-01 required all BWR Licensee's to provide a response to the subject letter.
By letter dated August 1, 1988, TVA provided NRC with the BFN Program responding to the IGSCC concern.
In that letter, TVA committed to amend the BFN TS prior to restart incorporating the NRC proposed TS provided in Generic Letter 88-01.
DESCRIPTION AND JUSTIFICATION FOR THE CHANGE BFN TS 235 added the Surveillance Requirements for ASME Section XI Pump and Valve Program to the Definition Section 1.0.MM.
As suggested by Generic Letter 88-01, the TS requirements regarding IGSCC are also being added to this section.
Therefore, the following TS, 1.0.MM.6, is being proposed:
1.0.MM.6 The Inservice Inspection Program for piping identified in NRC Generic Letter 88-01 shall be performed in accordance with the staff positions on schedule,
- methods, personnel, and sample expansion included in this Generic Letter.
4
JUSTIFICATION FOR THE CHANGE Through the various research conducted, NRC has stated that BWR piping weldments made of austenitic stainless steel are susceptible to IGSCC.
The three elements that, in combination, cause IGSCC are, a susceptible (sensitized)
- material, a significant tensile stress, and an aggressive environment.
The Commission has determined that, unless appropriate remedial actions are taken, BWR plants may not be in conformance with their current design and license bases, including 10CFR50 Appendix A, General Design Criteria 4, 14, and 31.
Based on this, NRC provided the industry with their positions which were issued in Generic Letter 88-01.
This applies to all BWR piping made of austenitic stainless steel that is four inches or larger in nominal diameter and contains reactor coolant at a temperature above 200 F during power operation regardless of Code Classification.
It also applies to reactor vessel attachments and appurtenances such as jet pump instrumentation penetration assemblies and vent components.
Among those positions, NRC provided a proposed TS which was to be incorporated into the Inservice Inspection (ISI) section of the TS.
Implementation of this TS endorses the staff's schedule,
- methods, personnel, and sample expansion as outlined by Generic Letter,88-01.
Implementing an IGSCC program at BFN ensures that the integrity of the subject piping is maintained.
Any cracks
- found, as a result of the IGSCC
- program, would be repaired therefore, minimizing any potential of a pipe break as a result of IGSCC.
This would not only ensure system integrity, it would also provide assurance that the health and safety of the plant workers and public would also be maintained.
Meeting the positions of the subject Generic Letter, BFN will be provi'ding appropriate remedial actions to resolve NRC concerns pertaining to IGSCC.
F,
ENCLOSURE 3
DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION BROWNS FERRY NUCLEAR PLANT (BFN)
UNITS 1, 2, AND 3 DESCRIPTION OF PROPOSED TECHNICAL SPECIFICATION AMENDMENT The proposed amendment would change the BFN Technical Specifications (TS) for units 1, 2, and 3 to add requirement 1.0.MM.6 to the Definitions Section.
This requirement would ensure that an Inservice Inspection Program for piping identified in NRC Generic Letter 88-01 be performed in accordance with developed NRC positions.
BASIS FOR PROPOSED NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION NRC has provided standards for determining whether a significant hazards consideration exists as stated in 10CFR50.92(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 of consequences of an accident previously evaluated, or 2) create the possibility of a new or different kind of accident from an 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 consequence of any accident previously evaluated.
BFN Final Safety Analysis Report (FSAR) Chapter 14 provides analysis of Design Basis Accidents (DBA) in which BFN was analyzed and licensed.
In reviewing these DBA's, the one closest to the IGSCC issue would be the Loss of Coolant Accident (LOCA) discussed in FSAR section 14.6.3.
In that analysis, it is assumed that the reactor is operating at the most severe condition at the time the recirculation pipe breaks, which would maximize the parameter of interest:
- response, fission product release or Core Standby Cooling System requirements.
In addition, the recirculation loop pipeline is assumed to be instantly severed.
This results in the most rapid coolant loss and depressurization with coolant discharged from both ends of the break.
j
' 1
~
0 The IGSCC concern of the staff results from various BWR Plants identifying that they have experienced some cracking in weldments of austenitic stainless steel piping.
The proposed TS would require an inspection program be performed in accordance with NRC guidelines to ensure that the potential of pipe weldment cracking be minimized.
This TS would apply to all BWR piping made of austenitic stainless steel that is four inches or larger in nominal diameter and contains reactor coolant at a temperature above 200 'F during power operation regardless of Code Classification.
Implementation of this TS would ensure that weldment cracking would be detected and fixed before a pipe would rupture.
As a result, the proposed TS would provide added assurance of not exceeding any assumptions or results for the LOCA analysis stated above.
In Generic Letter 88-01, the Commission stated that unless appropriate remedial actions are taken, BWR plants may not be in compliance with their current design and licensing bases, including 10 CFR 50, Appendix A, General Design Criteria (GDC) 4, 14, and 31.
NRC proposed a TS, in which BFN is hereby proposing, which will ensure implementation of a NRC approved program for IGSCC.
This program provides appropriate remedial
- actions, therefore, providing added assurance that BFN is within its design bases and appropriate 10 CFR 50 Appendix A GDC's.
2.
The proposed change does not create the possibility of a new or different kind of accident from an accident previously evaluated.
The proposed change does not change or modify the operation or design of any safety related equipment currently installed at BFN.
This proposed TS would enhance the overall plant integrity through the implementation of a program that would provide added assurance that the pressure boundary piping integrity would be maintained.
This proposed TS change is administrative in nature and does not introduce any new conditions which would create a new or different accident that has been previously analyzed.
3)
The proposed change does not involve a significant reduction in a margin of safety.
Theproposed change is administrative in nature and in fact enhances the margin of safety at BFN.
Implementation of a program in accordance with Generic Letter 88-01 requires an ISI program to monitor specific piping that may be susceptible to IGSCC.
This program would assist in detection of weldment cracking in austenitic stainless steel piping as outlined in the subject letter.
The addition of this program provides added assurance that any cracking would be detected and fixed therefore, eliminating any added potential of a pipe rupture.
The implementation of this TS enhances the overall integrity and safety of BFN.
This proposed change also supports the design and licensing bases of BFN, in addition to supporting 10CFR50 Appendix A, GDC 4, 14, and 31.
ENCLOSURE 4 BROGANS FERRY NUCLEAR PLANT UNIT 2 REVISED TECllNICAL SPECIFICATION PAGE 3.5/4.5-27 AND REVISED PAGE 4 FOR THE DESCRIPTION AND JUSTIFICATION (ENCLOSURE 2)
FOR TECHNICAL SPECIFICATION 249 SUBMITTED AUGUST 12, 1988
1 I'
3 5
1hULM (Cont'd)
There are foux RHR heat exchanger headers (Ap Bp Cp < D) vith one RHR heat exchanger from each unit on each header.
Thexe are two RHRSW pumps on each header; one normally assigned to each header (A2p B2p C2i ox'2) and one an alternate assignment (Al, Bl, Cl, ar Dl).
One RHR hest exchanger header can adequately deliver the flow supplied by both RHRSW pumps to any two af the three RHRSW heat exchnngers on the header.
One RHRSW pump cnn supply the full flaw requirement of ane RHR heat exchanger.
Two RHR exchangers can more than adequately handle the cooling requixements of one unit in any abnormal or postaccident situation.
When the decay heat level has decreased sufficiently following SlltJTDOMN, the entire shutdown cooling load can be adequately handled by one RHR Heat exchanger.
The RHR Service Water System was designed as a shared system fox three units.
The specification, as written, is conservative vhen consideration is given to particular pumps being out of service and to possible valving axxangements.
If unusual operating conditians axise such that more pumps are out of service than allowed by this specification, a special case request may be made to the HRC to allow continued operatian if the actual system cooling requirements can be assured.
Should three of the four RHRSW pumps noxmally ar alternately assigned to the RHR heat exchanger headers supplying the standby coolant supply connection become inoperable, an equal capability for long-term fluid makeup to the unit reactor and for cooling of the unit containment
,xemains operable, Because of the availability of an equal makeup and caoling capability, a 30-day repair period is )ustified.
Should the capability to pravide standby coolant supply be lost, a 10-day repair time is )ustified based on the low probability for ever needing the standby coolant supply.
Verification that the LPCI subsystem cross-tie valve is closed and powered to its operator is disconnected ensures that each LPCI subsystem remains independent and a failure of the flow path is in one subsystem will not affect'he flow path of the other LPCX subsystem.
The plant Appendix R evaluation requires that the RHRSW pumps D2 and either Cl or C2 to be operable during. Unit.2 reactor power operation'.
- These pumps are required to be operable ta ensure the one required RHRSW pump is available for a specific fire location. If ane of the two required RHRSW pumps is out of service, a hourly patrolling fire watch will be established in the appropriate fire areas/mnes as a
compensatory
- measure, t
3oS.D There is an equipment area cooler for each RHR pump and an equipment area cooler for each set (two pumps, either the A and C or B and D
pumps) of core spray pumps.
The equipment area coolers take suction near the caoling air discharge of the motor of the pump(s) served nnd discharge air near the cooling air suction of the motor of the pump(s)
- served, This ensures that cool air is supplied for cooling. the pump motors.
BFg Unit 2 3.5/4.5-27
4
S
~ - The existing unit 2 technical specifications for the RHRSW and EECW pumps permit indefinite plant operation with one RHRSW and one EECW pump inoperable when three units are operating.
The number of RHRSW pumps required to be operable is further reduced with units 1 and 3 in a cold shutdown condition or 'defueled, The plant Appendix R evaluation assumed the availability of all four EECW pumps.
In fire areas 9,
16> and 18, and fire
~ones 2-1> 2-2, 2-3, 2-4, 2-5, and 2-6, a postulated fire could result in only two EECW pumps being available that are required by the plant Appendix R evaluation. If one of these two EECW pumps was the one currently allowed by the technical specifications to be indefinitely out of service, then the required two EECW pumps for safe shutdown would not be available.
For a fire in any other areas/zones of the plant, adequate RHRSW swing/EECW pumps are available to supply necessary cooling water to the diesel generators even if one of the EECW pumps is out of service.
The plant Appendix R evaluation required that the RHRSW pumps D2 and either Cl or C2 to be operable during Unit 2 reactor power operation Th~~~ pumps <<e required to be operable to ensure the one required RHRSW pump is available for a specific fire location. If one of the two required RHRSW pumps ia out of
- service, an hourly patrolling fire watch will be established in the appropriate fire areas/zones as a compensatory
- measure, For postulated fires in any other areas of the plant (i.e., other than 2-lg 2-2y 2-3~ 2-4q 2-5> 2-6, 9, 16, 18),
one train of equipment needed to achieve and maintain hot shutdown will be free of fire damage through fire area boundary separation.
In those cases where hot shutdown is assured and alternate shutdown is not required, plant operating in'structions (e.g.,
EOIs) will be used to complete the cooldown process, The plant Appendix R
evaluation for Unit 2 operation further identified equipment which can be used to reach cold shutdown without repair.
With hot shutdown assured, adequate time is available for the operators to perform necessary actions using symptom oriented procedures (e,g.,
EOIs) to ensure that there are adequate RHRSW and EECW pumps available to achieve cold shutdown.
This vill provide the flexibilityto align equipment which may be operable but not necessarily a
preselected shutdown path.
The proposed technical specifications ensure a safe shutdown capability by providing a compensatory measure during plant operation with inoperable RHRSW and EECW pumps.
The compensatory measure of the patrolling fire watch provides assurances that the existence of unsafe or fire conditions would be minimized.
The previous discussion under MSRVs provides )ustification for patrolling fire watches.