ML18039A887
| ML18039A887 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 09/30/1999 |
| From: | Abney T TENNESSEE VALLEY AUTHORITY |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| TVA-BFN-TS-381, NUDOCS 9910180109 | |
| Download: ML18039A887 (174) | |
Text
ENCLOSURE 1
TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)
UNITS 1, 2 AND 3 PROPOSED TECHNICAL SPECIFICATION (TS)
CHANGE TS-381 DESCRIPTION AND EVALUATION OF THE PROPOSED CHANGE INDEX I.
DESCRIPTION AND EVALUATION OF THE PROPOSED CHANGE E1-1 II.
SAFETY ANALYSIS E1-18 III.
NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION E1-18 IV.
ENVIRONMENTAL IMPACT CONSIDERATION E1-20
DESCRIPTION OF THE PROPOSED CHANGES AND REASON FOR THE PROPOSED CHANGES These administrative changes delete license conditions that have become outdated, are no longer applicable or are redundant and consolidate remaining license 'conditions into one location in each units Technical Specifications (TS).
These changes are proposed to ensure the license conditions in the Operating License (OL) reflect the current status of the plant and to simplify the use of these documents.
The proposed administrative changes to the Units 1, 2 and 3
OL consist of the following:
~
Deletion of an expired general design criterion exemption,
~
Deletion of redundant paragraphs on the physical security plan and the guard training and qualification
- plan,
~
Deletion of an authorization to temporarily store low level radioactive material on site which has expired,
~
Deletion of a post-Three Mile Island related commitment which has been completed,
~
Deletion of an authorization to plug bypass flow holes in the lower core support plate,
~
Deletion of the authorization to drill bypass flow holes in Type 2 and Type 3 fuel assemblies,
~
Deletion of the authorization to modify Units 1, 2 and 3
Emergency Core Cooling Systems for performance improvements which have been completed,
~
Deletion of requirements regarding the Mark I containment owners group short term and long term program related to dynamic loads associated with a postulated loss-of-coolant accident which have been completed, and
~
Consolidate license conditions which currently exist in two locations in each units TS.
Pro osed Chan e
Unit 2 OL page 4, paragraph 2.C.(4) currently reads as follows:
The licensee is hereby granted an exemption from the requirements of General Design Criterion 4 with respect to high energy pipes outside the containment in accordance with the conditions set forth in the Technical Specifications, section 3.6.G.2 which requires completion of those items listed in "Concluding Report on the Effects of Postulated Pipe Failure Outside of Containment for the Browns Ferry Nuclear Plant Units 2 and 3" and related to Unit 2 prior to startup of Unit 2 following the first refueling outage.
For Unit 2, delete OL paragraph 2.C.(4).
Reason for the Pro osed Chan e
The temporary exemption for General Design Criterion (GDC) 4 was granted by the NRC in amendment number 1 to the Unit 2 Operating License by letter from the NRC to TVA dated August 2, 1974.
GDC 4 is the General Design Criteria for Nuclear Power Plants from 10 CFR 50 Appendix A entitled "Environmental and Dynamic Effects Design Bases".
This exemption was granted as a result of TVA's report entitled "Concluding Report on the Effects of Postulated Pipe Failure Outside of Containment for the Browns Ferry Nuclear Plant Units 2 and 3" dated March 1, 1974.
An exemption to the requirements of GDC 4 is no longer needed since Browns Ferry Unit 2 now meets these requirements.
TVA Report Civil Engineering Branch (CEB) 88-06-C entitled "Pipe Rupture Evaluation Program for Inside and Outside Primary Containment for the Browns Ferry Nuclear Plant Unit 2" documents that all the required actions in "Concluding Report on the Effects of Postulated Pipe Failures Outside Containment for the Browns Ferry Nuclear Plant Units 2 and 3" have been completed.
The exemption for GDC 4 concerning high energy pipes outside containment was applicable during the period of time prior to the startup of Unit 2 following the first refueling outage.
Unit 2 is currently in its eleventh operating cycle following its tenth refueling outage.
This exemption has expired, is no longer needed, and therefore can be removed from the Unit 2 license conditions.
E1-2
0
Pro osed Chan e
Unit 1
OL page 4, paragraph 2.C.(10), Unit 2 OL page S,
paragraph 2.C.(10),
and Unit 3 OL page 4, paragraph 2.C.(5) currently read as follows:
The licensee shall follow all provisions of the NRC approved Guard Training
& Qualification Plan, including amendments and changes made pursuant to 10 CFR 50.54(p).
The approved Guard Training Qualification Plan is identified as "Browns Ferry Nuclear Power Station Guard Training
& Qualification Plan," dated August 17,
- 1979, as revised by pages dated January 24,
- 1980, May 21,
- 1980, October 1,
- 1980, and March 9, 1981 and as may subsequently be revised in accordance with 10 CFR 50.54(p).
The Guard Training
& Qualification Plan shall be followed, in accordance with 10 CFR 73.5S(b),
60 days after the date of this amendment.
Unit 1
OL page 4, paragraph 2.C (8), Unit 2 OL page 5,
paragraph 2.C.
(8),
and Unit 3 OL page 4, paragraph 2.C.
(4) currently read as follows:
The licensee shall maintain in effect and fully implement all provisions of the Commission approved physical security plan including amendments made pursuant to the authority of 10 CFR 50.54(p)
The approved plan, which contains information protected under 10 CFR 73.21, is entitled "Browns Ferry Nuclear Plant Physical Security Plan",
dated May 15, 1982 (TVA letter dated June 11, 1982) and revisions submitted by TVA letters dated August 31, 1982 and October 19, 1982.
For Units 1 and 2 delete OL paragraphs 2.C. (8) and (10).
For Unit 3 delete OL paragraphs 2.C.(4) and (5).
Reason for the Pro osed Chan e
The implementation requirements of the physical security
- plan, the guard training and qualification plan and the safeguards contingency plans are currently discussed in three separate license conditions in Units 1, 2 and 3 TS.
The most recently added license condition encompasses the requirements of the two earlier license conditions.
Units 1 and 2
OL paragraphs 2.C.(10) and Unit 3 OL paragraph 2.C.(5) were added by amendments 72, 69 and 44 to Operating License numbers DPR-33, DPR-52 and DPR-68 by letter from the NRC to TVA dated June 15, 1981.
These amendments modified each units OL to include a requirement to maintain a Guard Training and Qualification Plan to be followed in accordance with 10 CFR 73.55(b)(4).
0
These amendments were made in response to letters from TVA dated August 17,
- 1979, February 20,
- 1980, June 2,
- 1980, October 24,
- 1980, and April 7, 1981.
TVA requested further license amendments in a letter dated April 3,
- 1984, regarding changes to the physical security plan dated May 15, 1982.
Additional changes to the license amendment request were submitted to the NRC August 31, 1982 and October 19,1982.
License amendment numbers
- 115, 109 and 83 to Operating Licenses DPR-33, DPR-52 and DPR-68 were subsequently granted in a letter from the NRC to TVA dated October 29, 1984.
These amendments were requested due to upgrades in the Browns Ferry Physical Security Plan.
License Conditions added as a result of this amendment were 2.C.(8) for Units 1
and 2, and 2.C.(4) for Unit 3.
The physical security plan and the guard training and qualification plan are also described in Unit 1 and Unit 2 OL paragraphs 2.C.(11) and Unit 3 OL paragraph 2.C.(6).
This paragraph was added to each units OL as a result of additional enhancements to the Physical Security Plan by license amendments
1988.
TVA requested these amendments by letters dated September 15,
- 1987, November 23, 1987 and May 24, 1988.
The additional license condition supersedes the requirements of the previous two security related license conditions by referencing more recent revisions of the Physical Security Plan and the Security Personnel Training and Qualification Plan.
It also included the Browns Ferry Safeguards Contingency Plan into the license conditions.
Additionally, it required the licensee to satisfy the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 and accompanying 10 CFR 73.70 record reporting requirements.
Units 1 and 2
OL page 5, paragraph 2.C.(11) and the Unit 3 OL page 4, paragraph 2.C.(6) currently read as follows:
The licensee shall fully implement and maintain in effect all provisions of the Commission-approved physical security, guard training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p)
The
- plans, which contain Safeguards Information protected under 10 CFR 73.21, are entitled:
"Browns Ferry Physical Security Plan", with revisions submitted through May 24, 1988; "Browns Ferry Security
Personnel Training and Qualification Plan", with revisions submitted through April 16, 1987; and "Browns Ferry Safeguards Contingency Plan", with revisions submitted through June 27, 1986.
Changes made in accordance with 10 CFR 73.55 shall be implemented in accordance with the schedule set forth therein.
It is proposed that paragraphs 2.C.(8) and (10) for Units 1
and 2 be deleted.
It is proposed that paragraphs 2.C.(4) and (5) for Unit 3 be deleted.
It is also proposed that Units 1 and 2
OL page 5, paragraph 2.C.(11) and the Unit 3 OL page 4, paragraph 2.C.(6) remain unchanged.
Units 1 and 2
OL page 5, paragraph 2.C.(11) and the Unit 3 OL page 4,
paragraph 2.C.(6) supersede the requirements of the earlier license conditions.
TVA is requesting this change to eliminate the redundant superseded license conditions.
Pro osed Chan e
Unit 1
OL page 5, paragraph 2.C.(12), Unit 2 OL page 6,
paragraph 2.C.(12) and Unit 3 OL page 5, paragraph (8) currently read as follows:
The licensee is authorized to temporarily store low-level radioactive waste in an existing covered pavilion that is situated outside the security fence, as presently located, but inside the site exclusion area.
The total amount of low-level waste to be stored shall not exceed 1320 curies of total activity.
This authorization expires two years from the effective date of this amendment and is subject to all the conditions and restrictions in TVA's application dated January 21, 1980.
Delete Unit 1 OL page 5, paragraph 2.C.(12), Unit 2 OL page 6, paragraph 2.C.(12) and Unit 3 OL page 5, paragraph 2.C. (8).
Reason for the Pro osed Chan e
Browns Ferry requested this authorization to temporarily store radioactive waste on site by letter from TVA to the NRC dated January 21, 1980.
The radioactive waste was to be stored in a warehouse initially designed for storage of electrical cable.
This warehouse was referred to as the "cable warehouse".
The request was supplemented by a letter dated February 25,
- 1980, which included estimates of the incremental occupational man-rem exposure expected to be incurred as a result of the temporary onsite storage of low-level waste at Browns Ferry.
By letter from TVA to the NRC dated March 13,
- 1980, TVA amended the request to specify that all of the waste stored in the cable warehouse would be removed no later than two years from the official date of NRC approval of the license amendment.
At that time, the
need for temporary storage of radioactive waste was due to a severe cutback in TVA's allocation of radioactive waste
~
shipments to the Barnwell burial site.
It was also believed that there was a very real possibility of a complete, closure of all commercial disposal facilities.
TVA's intent was to use the cable warehouse as an interim measure which was part of a plan to provide for long term onsite storage of radioactive waste.
The long term plan included constructing permanent facilities for the purpose of radioactive waste storage.
The Commission issued amendments 60, 55 and 32 to Operating Licenses DPR-33, DPR-52 and DPR-68 on March 17, 1980.
The authorization to temporarily store low-level radioactive waste expired two years from the date of that license amendment.
The cable warehouse has not been used for storage of radioactive material since the expiration of the license amendment.
On November 10, 1981 the NRC issued Generic Letter (GL) 81-38 which allowed licensees to determine the acceptability of onsite low level radioactive waste storage 'through the 10 CFR 50.59 process thereby eliminating the need for any further license amendments for these types of situations.
The remaining license condition has expired and is no longer needed.
Therefore, this license condition may be deleted.
Pro osed Chan e
Unit 2 OL page 6, paragraph 2.C.(13) currently reads:
Commission Order dated March 25, 1983 is modified as follows: in Attachment 1, for item II.F.1.1.
and II.F.1.2 change "12/31/84" to "Prior to startup in Cycle 6."
Unit 3 OL page 6, paragraph 2.D.(4) currently reads:
Commission order dated March 25, 1983 is modified as follows: In Attachment 1, for item II.F.1.1 and II.F.1.2 change "12/31/84" to "Prior to Unit 2 startup in Cycle 6."
Delete Unit 2 OL paragraph 2.C.(13) and Unit 3 OL paragraph 2.D. (4).
Reason for the Pro osed Chan e
On March 17,
- 1982, and May 5,
- 1982, the NRC issued GL 82-06 and 82-10 which requested that licensees reconfirm or establish a firm schedule for completing certain TMI-2 action items, the requirements for which are specified in NUREG-0737.
Schedules for completion of these items were transmitted to NRC by letter from TVA dated January 14, 1983.
NRC correspondence regarding these schedules may be found in letters dated February 1,
1983 and March 25, 1983.
In the latter correspondence, the NRC indicated to TVA that the implementation schedule proposed by TVA was acceptable.
A request for a modification to the agreed upon implementation dates was requested by TVA in a letter dated November 28, 1984.
NRC response dated December 7,
1984 indicated that this request should proceed in accordance with the formal license amendment process.
Accordingly, TVA requested a license amendment by letter dated December 13, 1984.
Amendment numbers 110 and 85 to Operating Licenses DPR-52 and DPR-68, were subsequently granted as indicated in a letter from the NRC to TVA dated February 12, 1985.
This amendment modified commission orders dated March 25,
- 1983, to extend the date for installation of noble gas and iodine effluent monitors with local readout capability from December 31,
- 1984, to prior to Unit 2 startup in cycle 6.
By letter dated January 4,
- 1991, TVA notified NRC that the items II.F.1.1 and II.F.1.2 had been implemented at Browns Ferry.
NRC Inspection Report (IR) numbers 50-259/91-06, 50-260/91-06 and 50-296/91-06 were transmitted to TVA by letter dated April 12, 1991.
In this IR, the following status of the applicable TMI action items was given:
"The inspector reviewed the following TMI items:
(OPEN)
- 259, 260, 296/II.F.1.2.A, Accident Monitoring Noble Gas Monitor.
(OPEN)
- 259, 260, 296/II.F.1.2.B, Accident Monitoring, Iodine/Particulate Sampling.
These two items were reviewed and documented in IRs 90-29 and 91-01.
All field modification work was completed.
TVA was in the process of completing the applicable operating procedures, the surveillance instruction and the maintenance procedures.
The last outstanding activity for these two items will be the successful completion of the SI which was scheduled shortly after the period of this inspection.
NRC Inspection Report numbers 50-259/91-16, 50-260/91-16 and 50-296/91-16 included the results of further reviews of the status of these actions.
In this report transmitted by letter to TVA dated June 5,
- 1991, the TMI action regarding item II.F.1.2.A was closed.
The report stated that the SI's were completed satisfactorily on May 17, 1991.
Item II.F.1.2.B was evaluated as resolved for restart but not closed.
An unresolved item was opened (URI 91-01-01) concerning the adequacy of calculations for sampling line cases as detailed in the inspection report.
E1-7
On May 23, 1991 permission to restart Browns Ferry Unit 2 was granted.
This approval was documented in a letter to
~
TVA from the NRC dated June 18, 1991.
URI 91-01-01 was closed as documented in NRC Inspection Report number 50-259/92-10, 50-260/92-10 and 50-296/92-10 as documented by letter from the NRC to TVA dated April 28, 1992.
These license conditions were required to be satisfied prior to the restart of Unit 2 which occurred in May of 1991.
As documented
- above, the requirements of the license conditions have been satisfied.
Therefore, the license conditions may be deleted.
Pro osed Chan e
Unit 1
OL page 3, paragraph 2.C.(5) and Unit 2 OL page 4,
paragraph 2.C.(5) currently read as follows:
The facility may be modified by plugging the bypass flow holes in the lower core support plate as described in Browns Ferry Nuclear Plant Units 1 and 2 Safety Analysis Report for Plant Modifications to Eliminate Significant In-Core Vibrations (NEDC-21091),
October 1975.
The reactor shall not be operated with the plugs installed in the lower core support plate bypass flow holes without further authorization by the NRC.
Delete Unit 1 OL paragraph 2.C.(5) and Unit 2 OL paragraph 2.C. (5).
Reason for the Pro osed Chan e
By letter dated November 5,
- 1975, TVA submitted a request to the NRC for authorization to plug bypass flow holes in the lower core support plate on Browns Ferry Units 1 and 2.
The Commission granted authorization to proceed with installation of the plugs by issuing license amendments numbers 17 and 14 to Browns Ferry Operating Licenses DPR-33 and DPR-52 for Units 1 and 2 by letter dated November 14, 1975.
This amendment did not authorize operation under this condition.
Units 1 and 2 were modified by plugging bypass flow holes in the lower core support plate in accordance with this authorization for the purpose of eliminating significant in core vibrations as described in Browns Ferry Nuclear Plant Units 1 and 2 Safety Analysis Report for Plant Modifications to Eliminate Significant In-Core Vibrations (NEDC-21091).
NRC approval was subsequently obtained for Units 1 and 2 to operate with these plugs installed by letter from the NRC to TVA dated August 20, 1976.
This approval was granted by issuance of amendment numbers 27 and 24 to facility Operating Licenses DPR-33 and DPR-52 for Browns Ferry Units
1 and 2.
Browns Ferry Units 1 and 2 remain in the condition approved by these amendments with the bypass flow holes in
~
the lower core support plate plugged.
Since the original license condition authorized the holes to be plugged but specifically prohibited operation under this modified condition without further authorization by the
- NRC, and the authorization to operate with the holes plugged was subsequently
- granted, this license condition may be deleted.
Pro osed Chan e
Unit 1
OL page 4, paragraph 2.C.(6) and Unit 2 OL page 4,
paragraph 2.C.(6) currently read as follows:
The facility may be modified by drilling bypass flow holes in Type 2 and Type 3 fuel assemblies as described in NED0-21091, "Browns Ferry Nuclear Plant, Units 1
& 2 Safety Analysis Report for Plant Modifications to Eliminate Significant In-Core Vibrations;" and NEDE-
- 21156, "Supplemental Information for Plant Modification to Eliminate Significant In-Core Vibrations," dated January 1976.
Delete Unit 1 OL paragraph 2.C.(6) and Unit 2 OL paragraph 2.C.
(6).
Reason for the Pro osed Chan e
By letter from the NRC to TVA dated March 3,
- 1976, amendment numbers 20 and 17 were granted for Operating Licenses DPR-33 and DPR-52 for Browns Ferry Units 1 and 2.
In this amendment, authorization was granted to modify Type 2 and Type 3 fuel assemblies in BFN Units 1 and 2 by drilling bypass flow holes in these assemblies.
However, this amendment did not authorize operation under this condition.
Approval of this modification was requested for the purpose of eliminating significant in core vibrations as described in Browns Ferry Nuclear Plants Units 1 and 2 Safety Analysis Report for Plant Modifications to Eliminate Significant In-Core Vibrations (NEDC-21091).
Accordingly bypass flow holes were drilled in these fuel assemblies and authorization to restart under these conditions was obtained.
By letter from the NRC to TVA dated August 20,
- 1976, approval to operate in this condition was granted by issuance of amendment numbers 27 and 24 to facility Operating Licenses DPR-33 and DPR-52 for Browns Ferry Units 1 and 2.
These fuel assemblies have been discharged from both units 1
and 2 reactor vessels.
These type of assemblies are of an outdated design and as such will not be used in the future.
Therefore, this license condition may be deleted.
E1-9
Pro osed Chan e
~
Unit 1
OL page 4, paragraph 2.C.(7) and Unit 2 OL page 4,
paragraph 2.C.(7) currently read as follows:
The facility may be modified as described in "Browns Ferry Nuclear Plant Units 1 and 2 Emergency Core Cooling Systems Low Pressure Coolant Injection Modifications for Performance Improvement (December 1975)" submitted by application dated December 1,
1975 and supplements dated February 12,
- 1976, March 24,
- 1976, March 30,
- 1976, May 21,
- 1976, June 11,
- 1976, and July 21, 1976.
Delete Unit 1 OL paragraph 2.C.(7) and Unit 2 OL paragraph 2.C. (7).
Reason f'r the Pro osed Chan e
Modifications to the Browns Ferry Units 1 and 2 Low Pressure Coolant Injection (LPCI) system as referenced in "Browns Ferry Nuclear Plant Units 1 and 2 Emergency Core Cooling Systems Low Pressure Coolant Injection Modifications for Performance Improvement (December 1975)" submitted by application dated December 1,
1975 and supplements dated February 12,
- 1976, March 24,
- 1976, March 30,
- 1976, May 21,
- 1976, June 11,
- 1976, and July 21,
- 1976, have been completed.
Subsequent authorization was obtained from the NRC to operate with these modifications by letter from the NRC to TVA dated August 20, 1976.
This letter issued facility Operating License amendment numbers 27 and 24 to licenses DPR-33 and DPR-52 for Browns Ferry Units 1 and 2.
Accordingly, these license conditions are no longer necessary and may be deleted.
Pro osed Chan e
Unit 1
OL page 4, paragraph 2.C.(9) and Unit 2 OL page 5,
paragraph 2.C.(9) currently read as follows:
The facility may be modified as described in "Browns Ferry Nuclear Plant Units 1 and 2 Emergency Core Cooling Systems Low Pressure Coolant Injection Modifications For Performance Improvement (October 1977)" submitted by letter dated December 28, 1977 and supplemented by letter dated December 13, 1978.
Delete Unit 1 OL paragraph 2.C.(9) and Unit 2 OL paragraph 2.C. (9).
Unit 3 OL page 6, paragraph 2.D.(3) currently reads as follows:
The facility may be modified as described in "Browns Ferry Nuclear Plant Unit 3 Emergency Core Cooling Systems Low Pressure Coolant Injection Modifications for Performance Improvement (October 1977)" and as described in TVA's letter of December 28, 1977 transmitting the aforementioned report and in TVA's supplemental letter of December 13, 1978.
Delete Unit 3 OL paragraph 2.D.(3).
Reason for the Pro osed Chan e
TVA committed to perform these modifications as documented in letters to the NRC dated December 28, 1977 and December 13, 1978.
Authorization to perform these modifications was granted by amendments 51, 45 and 23 to facility Operating Licenses DPR-33, DPR-52 and DPR-68 as transmitted by letter from the NRC to TVA dated May 11, 1979.
In this correspondence, TVA also committed to submit Technical Specification changes associated with the modifications in the reload amendment requests for each unit.
The modifications were to be completed during the second refueling outage for Unit 3 and the third refueling outages for Units 1 and 2.
Accordingly, amendment numbers 59 and 54 to Operating Licenses DPR-33 and DPR-52
, dated February 25,
- 1980, contained changes in the Technical Specifications for Units 1 and 2 to reflect the completed modifications to the Low Pressure Coolant Injection System.
This amendment was requested by letters from TVA to the NRC dated October 4,
- 1979, January 15, 1980 and January 29, 1980.
Amendment 28 to Operating License DPR-68 was transmitted by letter from the NRC to TVA dated November 30, 1979, in which the completed LPCI modifications were incorporated into the Technical Specifications for Unit 3.
This amendment was requested by letters from TVA to the NRC dated August 6,
- 1979, September 26,
- 1979, October 10, 1979 and October 25, 1979.
Also, the license amendments required that the Browns Ferry FSAR be revised to document the change in plant design.
The FSAR has been revised to read as follows:
"The LPCI has been modified for Units 1, 2,
and 3.
This modification has been described in reports entitled "Browns Ferry Nuclear Plant Units 1 and 2 Emergency Core Cooling Systems Low Pressure Coolant Injection Modifications for Performance Improvement" (October, 1977) for Units 1 and 2 and for Unit 3.
The reports were submitted to the Nuclear Regulatory Commission."
All required LPCI modifications have been completed as described in the Technical Specifications and the FSAR.
~
Therefore these license conditions may be deleted.
Pro osed Chan e
Unit 3 OL pages 5 and 6, paragraph 2.D.(1) and 2.D.(2) currently read as follows:
The licensee is required to assure that:
(a)The plant unique analysis for torus support structures and attached piping for the facility meets the approved Mark I Owners Group short-term acceptance criteria when subjected to dynamic loads associated with a postulated loss-of-coolant accident.
Should the licensee determine that the results of the plant unique analysis are not in conformance with the approved Mark I Owners Group short-term acceptance
- criteria, a
specific action plan will be developed by the licensee for the facility and presented to the Commission.
This action plan will include as a minimum the following information:
(1)The value of the load factor for which the criteria are satisfied.
(2)A description of proposed plant modifications or other action which will result in reduced loads or increased capacities that would satisfy the criteria.
(3)If a plant hardware modification is made, the acceptance criteria will be described on a plant unique basis.
(b)Upon completion of the Mark I Owners Group long-term program related to dynamic loads associated with a postulated loss-of-coolant accident, areas of design found not meeting the original design safety margins approved for the construction permit will be modified on a timely schedule to restore the original design safety margins.
(2)
The licensee is required, upon completion of the Mark I Owners Group containment long-term program related to relief valve operation, to make such modifications on a timely basis as may be necessary to restore the original design safety margins approved for the construction permit and used for the design of the torus structures when subjected to relief valve operation.
Delete Unit 3 OL paragraph 2.D. (1) and 2.D. (2).
Reason for the Pro osed Chan e
~
By'letter dated August 2,
- 1976, the NRC amended the Operating License for Browns Ferry Unit 3, License DPR-68, to reflect the requirements detailed above regarding the torus modifications planned as a result of the Mark I Containment owners group studies.
TVA provided a schedule for completion of these modifications for BFN units 1,
2 and 3 to the NRC by letter dated May 19, 1980.
As described in this letter, implementation of these modifications was scheduled for late 1983 on Unit 3.
On January 3,
- 1984, TVA submitted a letter containing the Plant Unique Analysis Report (PUAR) for the Browns Ferry Nuclear Plant to the NRC detailing the status of the modifications.
Torus modifications have been completed as described in the Mark I containment owners group short term and long term programs and as detailed in the PUAR submitted to the NRC on January 3,
1984.
As stated in the PUAR, the information contained therein complies with the acceptance criteria in NUREG 0661 and the associated Browns Ferry orders dated January 19, 1982.
This issue is also described in the Safety Evaluation Report for the Browns Ferry Unit 2 Restart (NUREG-1232, Volume 3
Supplement 2, Section 2.2.4.3, dated January 1991).
Section 2.2.4.3 of this report states that all issues relating to this design issue (Torus Modifications) were resolved under TVA's Browns Ferry Nuclear Plant Torus Integrity Long-Term Program.
Required modifications have been completed and found satisfactory, therefore the license conditions requiring these modifications may be deleted.
E1-13
~
Prb osed Chan e
Appendix B in Units 1, 2 and 3 Technical Specifications currently contain license conditions in addition to the conditions in the facility operating license section.
Unit 1 Appendix B currently contains the following additional license conditions.
Amend.
Number Additional Conditions Im lementation Date 234 The licensee is authorized to relocate certain requirements included in Appendix A and the former Appendix B to licensee-controlled documents.
Implementation of this amendment shall include the relocation of these requirements to the appropriate documents, as described in the licensee's application dated September 6,
- 1996, as supplemented December 11, 1996, April 11, May 1, August 14, October 15, November 5
and 14, December 3,
4, ll, 22, 23, 29, and 30,
- 1997, January 23, March 12 and 13, April 16, 20, and 28, May 7, 14, 19, and 27, and June 2, 5, 10 and 19,
- 1998, evaluated in the NRC staff's Safety Evaluation enclosed with this amendment.
This amendment is effective immediately and shall be implemented within 90 days of the date of this amendment.
234 The licensee shall review the Technical Specification (TS) changes made by License Amendment No.
234 and any subsequent TS
- changes, verify that the required analyses and modifications needed to support the changes are complete, and submit them for NRC review and approval prior to entering the mode for which the TS applies.
This amendment is effective immediately and shall be implemented prior to entering the mode for which the TS applies.
It is proposed that these two license conditions be relocated to the Unit 1 Operating License sections 2.C.(3) and 2.C.(4).
E1-14
Unit 2 Appendix 8 currently contains the following additional license conditions.
Amend.
NumEer Additional Conditions Im lementation Date 253 The licensee is authorized to relocate certain requirements included in Appendix A and the former Appendix B to licensee-controlled documents.
Implementation of this amendment shall include the relocation of these requirements to the appropriate documents, as described in the licensee's application dated September 6,
- 1996, as supplemented May 1, August 14, November 5
and 14, December 3,
4, 11, 22, 23, 29, and 30,
- 1997, January 23, March 12, April 16, 20 and 28, May 7, 14, 19, and 27, and June 2,
5, 10 and 19,
- 1998, evaluated in the NRC staff's Safety Evaluation enclosed with this amendment.
This amendment is effective immediately and shall be implemented within 90 days of the date of this amendment.
254 254 TVA will perform an analysis of the design basis loss-of-coolant accident to confirm compliance with General Design Criterion (GDC)-19 and offsite limits considering main steam isolation valve leakage and emergency core cooling system leakage.
The results of this analysis will be submitted to the NRC for its review and approval by March 31, 1999.
Following NRC approval, any required modifications will be implemented during the refueling outages scheduled for Spring 2000 for Unit 3 and Spring 2001 for Unit 2.
TVA will maintain the ability to monitor radiological conditions during emergencies and administer potassium-iodide to control room operators to maintain doses within GDC-19 guidelines.
This ability will be maintained until the required modifications, if any, are complete.
Classroom and simulator training on all power uprate related changes that affect operator performance will be conducted prior to operating at uprated conditions.
Simulator changes that are consistent with power uprate conditions will be made and simulator fidelity will be validated in accordance with ANSI/ANS 3.5-1985.
Training and the plant simulator will be
- modified, as necessary, to incorporate changes identified during startup testing.
This amendment is effective immediately.
This amendment is effective immediately.
It is proposed that these three license conditions be relocated to the Unit 2 Operating License sections 2.C.(3),
2.C.(4) and 2.C.(5).
~
Unit 3 Appendix B currently contains the following additional license conditions.
Amend.
NumEez Additional Conditions Im lementation Date 212 The licensee is authorized to relocate certain requirements included in Appendix A and the former Appendix B to licensee-controlled documents.
Implementation of this amendment shall include the relocation of these requirements to the appropriate documents, as described in the licensee's application dated September 6,
- 1996, as supplemented May 1, August 14, November 5
and 14, December 3,
4, 11, 22, 23, 29, and 30,
- 1997, January 23, March 12, April 16, 20, and 28, May 7,.14, 19, and 27, and June 2, 5, 10 and 19,
- 1998, evaluated in the NRC staff's Safety Evaluation enclosed with this amendment.
This amendment is effective immediately and shall be implemented within 90 days of the date of this amendment.
214 214 TVA will perform an analysis of the design basis loss-of-coolant accident to confirm compliance with General Design Criterion (GDC)-19 and offsite limits considering main steam isolation valve leakage and emergency core cooling system leakage.
The results of this analysis will be submitted to the NRC for its review and approval by March 31, 1999.
Following NRC approval, any required modifications will be implemented during the refueling outages scheduled for Spring 2000 for Unit 3 and Spring 2001 for Unit 2.
TVA will maintain the ability to monitor radiological conditions during emergencies and administer potassium-iodide to control room operators to maintain doses within GDC-19 guidelines.
This ability will be maintained until the required modifications, if any, are complete.
Classroom and simulator training on all power uprate related changes that affect operator performance will be conducted prior to operating at uprated conditions.
Simulator changes that are consistent with power uprate conditions will be made and simulator fidelity will be validated in accordance with ANSI/ANS 3.5-1985.
Training and the plant simulator will be
- modified, as necessary, to incorporate changes identified during startup testing.
This amendment is effective immediately.
This amendment is effective immediately.
It is proposed that these three license conditions be relocated to the Unit 3 Operating License sections 2.C.(3),
2.C.(4) and 2.C.(5).
Reason for the Pro osed Chan e
Currently, license conditions exist in two locations in Browns Ferry Units 1, 2 and 3 Technical Specifications.
License Conditions are specified in the Operating License as well as in Appendix B. It is proposed that the license conditions that are currently located in Appendix B'n each units Technical Specifications be moved into the license conditions sections in order to simplify the use of these documents.
El-17
II.
SAFETY ANALYSIS The proposed revisions are administrative in nature and do not reflect a change in the operation of BFN.
This request is submitted for the purpose of updating the BFN Technical Specifications license conditions to ensure they reflect the current status of the plant.
The revisions are being made to delete requirements that have been completed, are no longer needed, are redundant and to consolidate license conditions which currently exist in two locations in each units Technical Specifications.
Accordingly, there are no safety related concerns with the proposed revisions.
III. NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION TVA has concluded that operation of Browns Ferry Nuclear Plant (BFN) Units 1, 2 and 3 in accordance with the proposed change to the Operating Licenses does not involve a significant hazards consideration.
TVA's conclusion is based on its evaluation, in accordance with 10 CFR 50.91(a)(1),
of the three standards set forth in 10 CFR
- 50. 92 (c).
The proposed change consists of administrative revisions to the Operating License (OL) for BFN Units 1, 2 and 3 that delete license conditions that have become outdated, are no longer applicable, are redundant and to consolidate license conditions which currently exist in two locations in each units Technical Specifications.
These changes are proposed to ensure the license conditions in the FOL reflect the current status of the plant.
A.
The ro osed amendment does not involve a si nificant increase in the robabilit or conse ences of an accident reviousl evaluated.
The changes requested by this submittal are administrative in nature and do not change the way BFN operates.
The proposed changes are intended to:
~ delete redundant paragraphs,
~ delete requirements and authorizations for modificat'ions that have been completed,
~ delete an authorization to temporarily store radioactive material on site,
~ delete an exemption from a General Design Criterion which has expired, and
~ consolidate license conditions which currently exist in two locations in each units Technical Specifications
The change does not affect any design bases accident or the ability of any safe shutdown equipment to perform its design function.
There are no physical modifications that are required to implement this license condition update.
There is no impact on plant equipment or changes to operating procedures.
Therefore, the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.
The ro osed amendment does not create the ossibilit of a new or different kind of accident from an accident reviousl evaluated.
The changes described above are administrative in nature and do not change the way BFN operates.
There are no physical modifications authorized by the proposed changes and there are no procedure or process changes that are requested.
Changes requested are intended to ensure the license conditions reflect the current status of the plant.
There is no impact on any accident analysis created by this change.
Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.
The ro osed amendment does not involve a si nificant reduction in a mar in of safet The changes described above are administrative in nature and do not change the way BFN operates.
There are no procedural or physical changes required by this amendment.
The license conditions are being updated partially as a result of NRC information Notice 97-43 which highlighted the importance of periodically verifying compliance with the Operating License.
These changes are intended to delete license conditions which are no longer needed or are redundant in order to ensure the license conditions accurately reflect the current status of the licensed facility.
The change does not affect any design bases accident or the ability of any safe shutdown equipment to perform its design function, therefore no margins of safety have been affected by any of these changes.
Accordingly, the proposed amendment does not involve a significant reduction in a margin of safety.
IV.
ENVIRONMENTAL IMPACT CONSIDERATION Thh ro osed chan e does not invo p
p g
lve a significant hazards consideration, a significant change in the types of or significant increase in the amounts of any effluents that may be released off site, or a significant increase in individual or cumulative occupational radiation exposure.
Therefore, the proposed change meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Therefore, pursuant to 10 CFR 51.22(b),
an environmental assessment of the proposed changes is not required.
E1-20
~
v.
2.
RE FERENCE S General Desi n Criterion Exem tion Concluding Report on the Effects of Postulated Pipe Failure Outside Containment for the Browns Ferry Nuclear Plant Units 2 and 3, dated March 1, 1974.
Amendment 1 to Operating License DPR-52, dated August 2, 1974.
3.
TVA Report CEB 88-06-C entitled "Pipe Rupture Evaluation Program for Inside and Outside Primary Containment for the Browns Ferry Nuclear Plant Unit 2", originally issued November 14, 1988 and revised August 1,
- 1990, September 28,
- 1995, November 1,
1995 and February 19, 1998.
Securit Plan Application for license amendment from TVA dated August 17, 1979.
2.
4.
5.
7.
8.
9.
10.
Supplemental application for license amendment from TVA to NRC dated February 20, 1980.
Supplemental application for license amendment from TVA to NRC dated June 2,
1980.
Supplemental application for license amendment from TVA to NRC dated October 24, 1980.
Supplemental application for license amendment from TVA to NRC dated April 7, 1981.
Amendments 72, 69 and 44 to Operating License numbers DPR-33, DPR-52 and DPR-68, letter from the NRC to TVA dated June 15, 1981.'pplication for license amendment from TVA to NRC dated April 3, 1984.
Supplemental application for license amendment from TVA to NRC dated August 31, 1982.
Supplemental application for license amendment from TVA dated October 19, 1982.
'mendments
Application for license amendment from TVA dated September 15, 1987.
Supplemental application for license amendment from TVA
dated November 23, 1987.
13:
14.
Supplemental application for license amendment from TVA dated May 24, 1988.
Amendments
1988.
Tem ora authorization to store radioactive waste onsite 2.
3.
Letter from TVA to NRC requesting authorization to store radioactive material onsite, dated January 21, 1980.
Letter from TVA to NRC to submit estimates of incremental occupational man-rem exposure expected to be incurred as a
result of the temporary storage of low-level radioactive material at Browns Ferry, dated February 25, 1980.
Letter from TVA to NRC revising the request for storage of radioactive waste onsite to specify a time period of 2 years for this authorization, dated March 13, 1980.
4.
Amendments 60, 55 and 32 to license numbers DPR-33, DPR-52 and DPR-68, dated March 17, 1980.
5.
NRC Generic Letter 81-38, Storage of low-level radioactive wastes at Power Reactor Sites, dated November 10, 1981.
TMI/NUREG-0737 Items Letter from TVA to the NRC regarding implementation schedules for NUREG-0737 requirements, dated January 14, 1983.
2.
3.
Letter from NRC to TVA regarding schedules on NUREG-0737
- Items, dated February 1,
1983.
Letter from NRC to TVA regarding the acceptability of the scheduled implementation dates for NUREG-0737 Items, dated March 25, 1983.
Letter from TVA to NRC requesting modification to the required implementation dates for NUREG-0737 Item II.F.1, dated November 28, 1984.
5.
Letter from NRC to TVA regarding the need to request a
license amendment for changes to NUREG-0737 implementation
- dates, dated December 7,
1984.
6.
~
'etter from TVA to NRC requesting license amendment to extend the deadlines for implementation of NUREG-0737 Items II.F.1.1 and II.F.1.2, dated December 13, 1984.
License amendment numbers 110 and 85 to licenses DPR-52 and DPR-68, transmitted by letter from NRC to TVA dated February E1-22
- 4
0
~
9.
10.
12.
12, 1985.
Letter from TVA to NRC, Notification of NUREG-0737, Items II.F.1.1 Noble Gas Effluent and II.F.1.2 Iodine/Particulate Monitors Implementation, dated January 4,
1991.
NRC Inspection Report numbers 50-259/91-06, 50-260/91-06 and 50-296/91-06, dated April 12, 1991.
NRC Inspection Report numbers 50-259/91-16, 50-260/91-16 and 50-296/91-16, dated June 5,
1991.
Letter from NRC to TVA regarding the restart of Browns Ferry Unit 2, dated June 18, 1991.
NRC Inspection Report numbers 50-259/92-10, 50-260/92-10 and 50-296/92-10, dated April 28, 1992.
Plu in lower core su ort late b ass flow holes Letter from TVA to NRC, Request to the NRC for authorization to plug bypass flow holes in the lower core support plate on Browns Ferry Units 1 and 2, dated November 5,
1975.
2.
3.
4.
License amendments numbers 17 and 14 to Browns Ferry Operating Licenses DPR-33 and DPR-52 for Units 1 and 2, NRC letter dated November 14, 1975.
General Electric Company Report NEDC 21091, "Browns Ferry Nuclear Plants Units 1 and 2 Safety Analysis Report for Plant Modifications to Eliminate Significant In-Core Vibrations", dated October
- 197S, TVA submittal by letter to NRC dated November 5,
1975.
License amendment numbers 27 and 24 to facility Operating Licenses DPR-33 and DPR-52 for Browns Ferry Units 1 and 2, letter from NRC dated August 20, 1976.
Drillin b ass flow holes in t e 2 and 3 fuel assemblies 2.
3.
Letters from TVA to NRC requesting authorization to drill bypass flow holes in type 2 and 3 fuel assemblies lower tie
- plates, dated November 5,
- 1975, November 28, 1975 and February S,
1976.
License amendments 20 and 17 to Operating Licenses DPR-33 and DPR 52, letter from NRC to TVA dated March 3, 1976.
General Electric Company Report NEDC 21091, "Browns Ferry Nuclear Plants Units 1 and 2 Safety Analysis Report for Plant Modifications to Eliminate Significant In-Core Vibrations", dated October
- 1975, TVA submittal by letter to NRC dated November 5,
1975.
E1-23
License amendment numbers 27 and 24 to facility Operating Licenses DPR-33 and DPR-52 for Browns Ferry Units 1 and 2, letter from NRC dated August 20, 1976.
ECCS Modifications for Performance Im rovement "Browns Ferry Nuclear Plant Units 1 and 2 Emergency Core Cooling Systems Low Pressure Coolant Injection Modifications for Performance Improvement (December 1975) "
Letter from TVA to NRC, 10 CFR Part 50 Appendix K
Calculations and proposed revisions to Technical Specifications for Browns Ferry Nuclear Plants Units 1 and 2,
Tennessee Valley Authority, Docket Nos.
50-259, 50-260, dated December 1,
1975.
Supplement to request regarding "Browns Ferry Nuclear Plant Units 1 and 2 Emergency Core Cooling Systems Low Pressure Coolant Injection Modifications for Performance Improvement (December 1975)" from TVA to NRC dated February 12, 1976.
Supplement to figure 23 from "Browns Ferry Nuclear Plant Units 1 and 2 Emergency Core Cooling Systems Low Pressure Coolant Injection Modifications for Performance Improvement (December 1975)", letter from TVA to NRC, dated March 24, 1976.
Supplement to request regarding "Browns Ferry Nuclear Plant Units 1 and 2 Emergency Core Cooling Systems Low Pressure Coolant Injection Modifications for Performance Improvement (December 1975)" from TVA to NRC dated March 30, 1976.
"Browns Ferry Nuclear Plant Units 1-3, RHR Pump Protection Against Operation in Excess of Design Runout", letter from TVA to NRC dated May 21, 1976.
Additional information regarding "Browns Ferry Nuclear Plant Units 1-3, RHR Pump Protection Against Operation in Excess of Design Runout", letter from TVA to NRC dated June 11, 1976.
Additional information regarding the installation of orifices in the RHR pump discharge lines at the Browns Ferry Nuclear Plant units 1-3, letter from TVA to NRC dated July 21, 1976.
License amendment numbers 27 and 24 to facility Operating Licenses DPR-33 and DPR-52 for Browns Ferry Units 1 and 2, letter from NRC dated August 20, 1976.
TVA letter to the NRC regarding modifications to the LPCI
- system, dated December 28, 1977.
TVA letter to the NRC regarding modifications to the LPCI
- system, dated December 13, 1978.
12.
License amendments 51, 45 and 23 to facility Operating Licenses DPR-33, DPR-52 and DPR-68, authorization to perform LPCI modifications, letter from the NRC to TVA dated May 11, 1979.
13.
14.
License amendment numbers 59 and 54 to Operating Licenses DPR-33 and DPR-52, letter from NRC dated February 25, 1980.
Letter from TVA to the NRC, request for license amendment, dated October 4,
1979.
15.
16.
17.
Letter from TVA to the NRC, request for license amendment, dated January 15, 1980.
Letter from TVA to the NRC, request for license amendment, dated January 29, 1980.
License amendment 28 to Operating License DPR-68, letter from the NRC to TVA dated November 30, 1979.
18.
19.
Letter from TVA to the NRC, request for license amendment, dated August 6,
1979.
Letter from TVA to the NRC, request for license amendment, dated September 26, 1979.
Letter from TVA to the NRC, request for license amendment, dated October 10, 1979.
21.
22.
Letter from TVA to the NRC, request for license amendment, dated October 25, 1979.
Browns Ferry FSAR section 6.4.4 entitled "Low Pressure Coolant Injection System".
Unit 3 Torus Modifications 2.
License amendment number 1 to facility Operating License DPR-68, letter from NRC to TVA dated August 2, 1976.
Letter from TVA to NRC detailing the schedule for implementation of torus modifications, dated May 19, 1980.
3.
4.
Letter from TVA to NRC transmitting the Torus Integrity Long-Term Program Plant Unique Analysis Report (PUAR) for the Browns Ferry Nuclear Plant.
NUREG-1232, Volume 3, Supplement 2,
Browns Ferry Unit 2
- Restart, dated January 1991.
Plant Unique Analysis Report (PUAR) for the Browns Ferry Nuclear Plant, January 3,
1984.
P
ENCLOSURE 2 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)
UNITS 1, 2 AND 3 PROPOSED TECHNICAL SPECIFICATION (TS)
CHANGE TS-381 MARKED PAGE S I.
AFFECTED PAGE LIST Technical Specifications Unit 1
3 4
5 6
Appendix B
(page 1)
II.
MARKED PAGES Unit: 2 Appendix B
(pages 1,
2)
Unit 3 3
5 6
Appendix B
(pages 1,
2)
See attached.
E2-1
P
TENNESSEE VALLEYAUTHORITY BROWNS FERRY NUCLEAR PLANT UNIT 1 DOCKET NO. 50-259 FACILITYOPERATING LICENSE License No. DPR-33 1.
The Atomic Energy Commission (the Commission) having found that:
A.
The application for license filed by the Tennessee Valley Authority (the licensee) 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 and all required notifications to other agencies or bodies have been duly made; B.
Construction of the Browns Ferry Nuclear Plant, Unit 1 (the facility) has been substantially completed in conformity with Construction Permit No. CPPR-29 and the application, as amended, the provisions of the Act and the rules and regulations of the Commission; C.
The facilitywilloperate in conformity with the application, as amended, the provisions of the Act, and the rules and regulations of the Commission; D.
There is reasonable assurance:
(i) that the activities authorized by this amended operating license can be conducted without endangering the health and safety of the public, and (ii) that such activities willbe conducted in compliance with the rules and regulations of the Commission; E.
The licensee is technically and financially qualified to engage in the activities authorized by this amended operating license in accordance with the rules and regulations of the Commission; F.
The licensee has satisfied the applicable provisions of 10 CFR Part 140, "Financial Protection Requirements and Indemnity Agreements," of the Commission's regulations; G.
The issuance of this amended operating license will not be inimical to the common defense and security or to the health and safety of the public; BFN-UNIT 1 Amendment No. 2 December 20, 1973
H.
Afterweighing the environmental, economic, technical and other benefits of the facility against environmental costs and considering available alternatives, the issuance of
, Facility Operating License No. DPR-33, as amended, is in accordance with 10 CFR Part 50, Appendix D, of the Commission's regulations; and I.
The receipt, possession, and use of source, byproduct and special nuclear material as authorized'by this amended license will be in accordance with the Commission's regulations in 10 CFR Parts 30, 40, and 70, including 10 CFR Sections 30.33, 40.32, 70.23, and 70.31.
2.
The Atomic Safety and Licensing Board having dismissed the proceeding relating to the licensing action in a "Memorandum and Order," dated November 27, 1973, Facility Operating License No. DPR-33 issued to the Tennessee Valley Authority on June 26, 1973 is hereby amended in its entirety to read as follows:
A.
This amended license applies to the Browns Ferry Nuclear Plant, Unit 1, a boiling water nuclear reactor and associated equipment (the facility), owned by the Tennessee Valley Authority. The facility is located in Limestone County, Alabama, and is described in the "Final Safety Analysis Report" (Amendment 9) as supplemented and amended (Amendments 10 through 52), the licensee's Draft Environmental Statement and supplement thereto dated July 1971, and November 8, 1971, respectively, and the licensee's Final Environmental Statement dated September 1, 1972.
B.
Subject to the conditions and requirements incorporated herein, the Commission hereby licenses Tennessee Valley Authority:
(1)
Pursuant to Section 104b of the Act and 10 CFR Part 50, "Licensing of Production and Utilization Facilities," to possess, use, and operate the facility at the designated location in Limestone County, Alabama, in accordance with the procedures and limitations set forth in this amended license; (2)
Pursuant to the Act and 10 CFR Parts 40 and 70, to receive, possess, and use at any time source and special nuclear material as reactor fuel in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report as supplemented and amended; (3)
Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use at any time any byproduct, source, and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; BFN-UNIT 1 Amendment No. 192 April01, 1993'
~,
(4)
(5)
Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use in amounts as required any byproduct, source, or special nuclear material without restriction to chemical or physical form for sample analysis or equipment and instrument calibration or associated with radioactive apparatus or components; Pursuant to the Act and 10 CFR Parts 30 and 70, to possess but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.
C.
This amended license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, Section 40 41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 3293 megawatts thermal.
Technical S ecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 236, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
For Surveillance Requirements (SRs) that are new in Amendment 234 to Final Operating License DPR-33, the first performance is due at the end of the first surveillance interval that begins at implementation of Amendment.
For SRs that existed prior to Amendment 234, including SRs with modified acceptance criteria and SRs whose frequency of performance is being extended, the first performance is due at the end of the first surveillance interval that begins on the date the surveillance was last performed prior to implementation of Amendment.
(3)
Geleted-.Amendment 234: The licensee is authorized to relocate certain requirements included in Appendix A and the former Appendix B to licensee-controlled documents.
Implementation of this amendment shall include the relocation of these requirements to the appropriate documents, as described in the licensee's application dated September 6, 1996, as supplemented December 11, 1996, April 11, May 1, August 14, October 15, November 5 and 14, December 3, 4, 11, 22, 23, 29, and 30, 1997, January 23, March 12 and 13, April 16, 20, and 28, May 7, 14, 19, and 27, and June 2, 5, 10 and 19, 1998, evaluated in the NRC staff's Safety Evaluation enclosed with this amendment. This amendment is BFN-UNIT 1 Amendment Nos. 235 and 236 November 30, 1998/December 23, 1998
effective immediately and shall be implemented within 90 days of the date of this amendment.
(4)
Q@eted-.Amendment 234:-The licensee shall review the Technical Specification (TS) changes made by License Amendment No. 234 and any subsequent TS changes, verify that the required analyses and modifications needed to support the changes are complete, and submit them for NRC review and approval prior to entering the mode for which the TS applies. This amendment is effective immediately and shall be implemented prior to entering the mode for which the TS applies.
(5)
The-fasility-may-be-medified-by-pluggIRg4he-bypass-fI Wi 44'Wy ARalysis-Reper44er-RaRt-MediAeatleRs-te-EIImIRate-SigRNeaMR-Qere Oil'~
I
~N l4 p
h4 vugh-the-plugs-IRstalled-IR4he-Iewer-seFe-suppeft-pla Gu fuFther-autherhatieR-by-the-NRQ-.Deleted Amendment Nos. 235 and 236 November 30, 1998/December 23, 1998
The-fasili
'liiag-bypass-Aew-heles-ie Type-2-aad-Type-3 fuel-asserabiies-as-dessabe¹ia-N, as-Feey-btuslear-Raat-,
Ii iiaiaate-SigaiAeaat-la-Qere-Nbratieas
,"dated Jaauary-1876-.
Deleted tA~ÃW W4 4'll Vaits-4-a ad-2-Eiaerg easy-Qere-Qeeliag-Syste MediAsatieas-fer-Perfeaaaase-liapreveraeat-(Qese wI" ~'w Ma I
I I
I I
July-2-1,1876-. Deleted as-Feay-Musica r-Raat at-lajestiea 44y 4
i i
I I 4'pl pl 449 pursuaaAe4h
'5+p).The-appreved-plaa;whish eeataias-iafeaaatiea-pretested-uad
., 's-eatitied-"Brewas-Ferry Nusiear-Raat-Physisal-Sesuaty-Raa
,"dated-M Mj ~
t i~<<4A aad-Giber-18&882-. Deleted t4 ib W4 I
Units-1-aadWCraergeaey-Qere-Qeeliagysteras-Dew-Pressure-Qeelaahajestiea MediAsatieas-Fer-PerfeRaaase-lisp reveraeat-(Qist dated-Deseiaber-28-,4877-a ad-suppleRieated-by-letter-dated Deseiaber-4~87L Deleted It%it IH ctl ~RC ~i%HI QualiAsatiea-Plaa;iasludiag-araeadraeats-aad-shaages-raade-pursuaat4e pl N
ii 0
liN ideatiAed-as-"Brewas-Feay-Nuslear-Pewer-Statie+Quard-Tmiaiag-8 Qk..., ~44A~
~iaiag-8WualiAsatiea-Raa-shall-be-fellewed-,i~sserdaase-with 4y lt~~
Wh'i N
I d
Amendment No. 202 February 01, 1994
(11)
The licensee shall fullyimplement and maintain in effect all provisions of the Commission-approved physical security, guard training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The plans, which contain Safeguards Information protected under 10 CFR 73.21, are entitled: "Browns Ferry Physical Security Plan", with revisions submitted through May 24, 1988; "Browns Ferry Security Personnel Training and Qualification Plan", with revisions submitted through April 16, 1987; and "Browns Ferry Safeguards Contingency Plan", with revisions submitted through June 27, 1986. Changes made in accordance with 10 CFR 73.55 shall be implemented in accordance with the schedule set forth therein.
74&'i 6'f, M ly Iesated-,but-inside4he-site-exslusien-area
.The4etal-ameunt-ef-lew-level-waste te-be-stered-shall-Bet-exseed4820-su utherizatieri expires-twe-yea
'-this-arneridrnent-and-is-subjest-te-aII
'DO' I"
~'l-,l Deleted (13)
Browns Ferry Nuclear Plant shall implement and maintain in effect all provisions of the approved Fire Protection Program as described in the Final Safety Analysis Report for BFN as approved in the SEs dated December 8, 1988, March 6, 1991, March 31, 1993, November 2, 1995 and Supplement dated November 3, 1989 subject to the following provision:
The licensee may make changes to the approved fire protection program without prior approval of the Commission only ifthose changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.
D.
This amended license is effective as of the date of issuance and shall expire midnight on December 20, 2013.
FOR THE ATOMICENERGY COMMISSION SI A. Giambusso A. Giambusso, Deputy Director for Reactor Projects Directorate of Licensing Date of Issuance:
DEC 201973 See Appendix B for additional License Conditions.
BFN-UNIT 1 Amendment No. 234 July 14, 1998
APPENDIX B Deleted AQRT4QNAL l QNDAPQNS AeieFid-.
ReeilieF AddleFiel-QeediTieee QR4 J4~
N '4 fester-Appends-te4eensee-centre Hed desurnents.IfnplemeRtatieR-ef-this-arnendfneat shaH-IRstude4he-releeatieR-ef-these-requirements te4he-apprep Rate-desuete nts-,as-desspibed-iR the4eeRsee'-s-appHeatien-dated-September-6-,
N QQ,~
MR-,M I AQ 0-1a; AprH-14,20-,a..., Rd-27,and JuRe-2,5,14-and48,1888-,evaluated-iR4he NRQ-staff'-s-Safety-EvaluatieR-enslesed-with4his arnendmeRt-.
234 T4e4eensee-shall-review4he-Teehaieai Rp M ti~ 1 1 a!L 1
Wl &14 ehaRges-,veFify-that4he-required-analyses-and aa li ~~I ae ~ NRc H
A I
IHt'-S-appHes-.
This-ameRdrneRt-is effective-irnraediately aRd-shaH-be eRteRRg-the-fnede-fer whieh4he-T4 sppHes-.
BFN-UNIT 1 Amendment No. 234 July 14, 1998
TENNESSEE VALLEYAUTHORITY DOCKET NO. 50-260 BROWNS FERRY NUCLEAR PLANT UNIT2 FACILITYOPERATING LICENSE License No. DPR-52 1.
The Atomic Energy Commission (the Commission) having found that:
A.
The application for license filed by the Tennessee Valley Authority (the licensee) 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 and all required notifications to other agencies or bodies have been duly made; B.
Construction of the Browns Ferry Nuclear Plant, Unit 2 (the facility) has been substantially completed in conformity with Construction Permit No. CPPR-30 and the application, as amended, the provisions of the Act and the rules and regulations of the Commission; C.
The facilitywill operate in conformity with the application, as amended, the provisions of the Act, and the rules and regulations of the Commission; D.
There is reasonable assurance:
(i) that the activities authorized by this operating license can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the rules and regulations of the Commission; E.
The licensee is technically and financially qualified to engage in the activities authorized by this operating license in accordance with the rules and regulations of the Commission; F.
The licensee has satisfied the applicable provisions of 10 CFR Part 140, "Financial Protection Requirements and Indemnity Agreements," of the Commission's regulations; BFN-UNIT2 June 28, 1974
n
G.
The issuance of this operating license will not be inimical to the common defense and security or to the health and safety of the public; H.
Afterweighing the environmental, economic, technical and other benefits of the facility against environmental costs and considering available alternatives, the issuance of Facility Operating License No. DPR-52 is in accordance with 10 CFR Part 50, Appendix D, of the Commission's regulations and all applicable requirements of said Appendix D have been satisfied; and I.
The receipt, possession, and use of source, by-product and special nuclear material as authorized by this license will be in accordance with the Commission's regulations in 10 CFR Parts 30, 40, and 70, including 10 CFR Sections 30.33, 40.32, and 70.23, and 70.31.
2.
The Atomic Safety and Licensing Board having dismissed the proceeding relating to the licensing action in a "Memorandum and Order," dated November 27, 1973, Facility Operating License No. DPR-52 is hereby issued to the Tennessee Valley Authority to read as follows:
A.
This license applies to the Browns Ferry Nuclear Plant, Unit 2, a boiling water reactor and associated equipment (the facility), owned by the Tennessee Valley Authority.
The facility is located in Limestone County, Alabama, and is described in the "Final Safety Analysis Report" (Amendment 9) as supplemented and amended (Amendments 10 through 55), the licensee's Draft Environmental Statement and Supplement thereto dated July 1971, and November 8, 1971, respectively, and the licensee's Final Environmental Statement dated September 1, 1972.
B.
Subject to the conditions and requirements incorporated herein, the Commission hereby licenses the Tennessee Valley Authority:
(1)
Pursuant to Section 104b of the Act and 10 CFR Part 50, "Licensing of Production and Utilization Facilities," to possess, use, and operate the facilityat the designated location in Limestone County, Alabama, in accordance with the procedures and limitations set forth in this license; BFN-UNIT2 June 28, 1974
(2)
Pursuant to the Act and 10 CFR Parts 40 and 70, to receive, possess, and use at any time source and special nuclear material as reactor fuel in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report as supplemented and amended; (3)
Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use at any time any byproduct, source, and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4)
Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use in amounts as required any byproduct, source, or special nuclear material without restriction to chemical or physical form for sample analysis or equipment and instrument calibration or associated with radioactive apparatus or components; (5)
Pursuant to the Act and 10 CFR Parts 30 and 70, to possess but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.
C.
This license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
Maximum Power Level The licensee is authorized to operate the facilityat steady state reactor core power levels not in excess of 3458 megawatts thermal.
(2)
Technical S ecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 259, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
For Surveillance Requirements (SRs) that are new in Amendment 253 to Final Operating License DPR-52, the first performance is due at the end of the first surveillance interval that begins at implementation of Amendment.
For SRs that existed prior to Amendment 253, including SRs with modified acceptance criteria and SRs whose frequency of performance is being extended, the first performance is due at the end of the first surveillance interval that begins on the date the surveillance was last performed prior to implementation of Amendment.
BFN-UNIT2 Amendment No. 259 August 02, 1999
(4)
Deleted-Amendment 253: The licensee is authorized to relocate certain requirements included in Appendix A and the former Appendix B to licensee-controlled documents.
Implementation of this amendment shall include the relocation of these requirements to the appropriate documents, as described in the licensee's application dated September 6, 1996, as supplemented May 1, August 14, November 5 and 14, December 3, 4, 11, 22, 23, 29, and 30, 1997, January 23, March 12, April 16, 20 and 28, May 7, 14, 19, and 27, and June 2, 5, 10 and 19, 1998, evaluated in the NRC staffs Safety Evaluation enclosed with this amendment. This amendment is effective immediately and shall be implemented within 90 days of the date of this amendment.
%i IC'esign-Qriterie~th-resp'-high-eRergy-pipes-eutside-eentainrneat-IR aseerdanse-with-the-senditien-set-ferth-IR4he-Teshnieal-Spesifisatiens-,
SeeheR4:G-.Q~hish-requires-sefRpietieR-e
'R "QensludiRg Re pert-eR4he-Effects-ef-Restula~pe-Failure-Outside-ef-Qentainmeat-feF the-BrewRs-FerFy-Nuslear-Plant-URI~Rd-"aRd-related-te-URIt-2-prier-te startup-ef-URM-fellewing4he4rst-refueling-eutage-.
Amendment 254: TVAwill perform an analysis of the design basis loss-of-coolant accident to confirm compliance with General Design Criterion (GDC)-19 and offsite limits considering main steam isolation valve leakage and emergency core cooling system leakage.
The results of this analysis willbe submitted to the NRC for its review and approval by March 31, 1999.
Following NRC approval, any required modifications will be implemented during the refueling outages scheduled for Spring 2000 for Unit 3 and Spring 2001 for Unit 2. TVAwill maintain the ability to monitor radiological conditions during emergencies and administer potassium-iodide to control room operators to maintain doses within GDC-19 guidelines.
This ability will be maintained until the required modifications, ifany, are complete This amendment is effective immediately.
(5)
The-fasAity-may-be-rnedified-by-plug gin g4he-bypass-flmv-heies-I+4he-Iewer a
lyA ly Vibratlens-gQGQQ-2409+-,Qst ster-shal&et-be-eperated
'Ml pie'~M I
I pa~i~
witheut-further-autherhatien-by4he-NRQ-.Amendment 254: Classroom and simulator training on all power uprate related changes that affect operator performance will be conducted prior to operating at uprated conditions.
Simulator changes that are consistent with power uprate conditions willbe made and simulator fidelitywill be validated in accordance with ANSI/ANS 3.5-1985.
Training and the plant simulator BFN-UNIT2 Amendment No. 248 July 08, 1997
will be modified, as necessary, to incorporate changes identified during startup testing. This amendment is effective immediately.
(5)(a)
Deleted fuel-assemblies-as-dessribed-i, s-Ferry-wuslear-FIaht-,
URIts<d~afety-ARalysis-Repert-for-PlaRt-MedifleatieRs-te-EhrAIAate liig 4'l~~ ,IllQR441N-,%WW~~
Pl WA JaRuary-'I876-.Deleted (7)
T46486iltty-fRa~&fROdifiedW&4essRbBEWR~OWR&FBFPpHUslea&PIaRt Vwts4-aRd4-GrRergeR~~eeli re-QeelaRt Nl p
lf IPSjl sUbrAItted-by-appIIsatieR-dated-Deserhbe&,&75-aRd-supplerReRts-dated Februa I
I I
I I
4~
BFN-UNIT2 Amendment No. 248 July 08, 1997
(8)
'lt Qemmissien-appreved-physisai-sesunty-plarHnsluding-amendments->vade pursuant-te-the-auth
~pproved-pla~hlsh sentainmnfeRnatien-pretested-end ntNed-"Brewns-Ferry Nuslear-Plant-Physlsa&esunty-Plan
,"dat, etter-dated dune4-1,1@~nd-re4siens-submitted-by-I~
an &4Aeber4.9,'N82-.Deleted (9)
The-fasility-may-be-rRedified-as-dessribed-in-"Brewns-Ferry-Nuslear-Plant Uaits4-andWGrnergen tnjestien-Medifisati prevernen+Qeteber-1877)-"
MJ
@Ph 4~, St.
(10)
The-IIGensee-shall-fellew-aII-provislens-Gf-the-NIRQ-approved-Quaff-TMInlng-8c Qualifisatien-Plan-,insiuding-arnendrne r44e 49<FR@
. 4fpl~l p~m Q
IdentiA rry-Nuslear-Power-Statics-Guard-Training-8 CA~
hy.p
~;0 e~"
Training-L-Qualifieatien-Plan-shall-be-fellewed-,i~sserdanse-with The licensee shall fullyimplement and maintain in effect all provisions of the Commission-approved physical security, guard training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The plans, which contain Safeguards Information protected under 10 CFR 73.21, are entitled: "Browns Ferry Physical Security Plan", with revisions submitted through May 24, 1988; "Browns Ferry Security Personnel Training and Qualification Plan", with revisions submitted through April 16, 1987; and "Browns Ferry Safeguards Contingency Plan", with revisions submitted through June 27, 1986. Changes made in accordance with 10 CFR 73.55 shall be implemented in accordance with the schedule set forth therein.
BFN-UNIT2 Amendment No. 221 February 01, 1994
existirig-severed-paviliea4hat-is-situated-eutside4he-sesurity-ferise-,as presently-lesated-,but-inside4he-site-ex4usien-area
.The-tetai-arneuRt-ef is-subjest-te-aII4he-senditiens-and-restristiens-iRWAVs-applisatieR-dated January-2-1-,4880-.Deleted (13)
Qemmissiea-Qrder-dated-Ma Alt I startup-io-Qy4e-6-."Deleted edified-as-fellows :in II II II (14)
Browns Ferry Nuclear Plant shall implement and maintain in effect all provisions of the approved Fire Protection Program as described in the Final Safety Analysis Report for BFN as approved in the SEs dated December 8, 1988, March 6, 1991, March 31, 1993, November 2, 1995 and Supplement dated November 3, 1989 subject to the following provision:
The licensee may make changes to the approved fire protection program without prior approval of the Commission only ifthose changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.
D.
This amended license is effective as of the date of issuance and shall expire midnight on June 28, 2014.
FOR THE ATOMICENERGY COMMISSION S/
A. Giambusso A. Giambusso, Deputy Director for Reactor Projects Directorate of Licensing
Attachment:
Appendices A & B - Technical Specifications Date of Issuance:
JUN 28, 1974 See Appendix B for additional License Conditions.
BFN-UNIT2 Amendment No. 253 July 14, 1998
APPENDIX B Deleted I
4 requirements-insluded-iR-Appe Ad~nd4he feR-Rer-AppeRd'heHed Q.-sml shall-inslude4he-FeleeatieR-ef-these-requirements IQW
'SS 44~
- 4QQQ, I I I
N SQQSQQI I I-14; Nevember-5-a 29-,and-30&497,danua~3-,IVlareh-'I~pFH 44, 'SS Qas-,s~
Q-,S.144 444;44444 QQ Ws SSQ Q ~ I It-8 amendmeht-.
This-a~meRt-is effesbve-immediately and-shaH-be W/A-wHI-pepferm-aR-aRalysis-ef-the-desigR-basis less-eWeelant-assideRt-te-seRfirm-semplia Bee effsite-limits-sensideFing-maiR-steam-iselatieR I
wl w III system-leakage.The-results-ef-this-analysis-wiH I Is~I I
I I IQQII~s 44IQQ appreval-,aRy-required-medifieatieRs-wHl-be implemeRted-duping4hepefueHRg-eutages
~ I~I~~
SQQISM~I I i~
I IISQ4 me niter-Fadielegieal-eenditiens-duRRg emeFgeRsies-aRd-admiRister-petassium-iedide-te seRtFel-Feem-epeFaters-te-maintaiR-deses-withiR QQQQQW Is II maiRtained-uRtil4he-required-medNeatieRs-,if aRy,are-eemplete-.
effeeQve-immediately-.
BFN-UNIT2 Amendment No. 254 September 08, 1998
APPENDIX B Deleted%9 Qfassreem-and-si~iater-traiRiRg-ea-aH-peweF IbdI g~p 4I4 hI I+ ~p lii aWprated-eenditiens.Simulater-ehaRges4hat a
rate-seRditieas-vN
'be->vade-and-sifRulater-fid asseFdaR and-Qe-phRt-simuiater-wN-be-unedified-,as 8
6 44M duFing-staMp-testiRg-.
BFN-UNIT2 Amendment No. 254 September 08, 1998
U J
TENNESSEE VALLEYAUTHORITY DOCKET NO. 50-296 BROWNS FERRY NUCLEAR PLANT UNIT 3 FACILITYOPERATING LICENSE License No. DPR-68 1.
The Nuclear Regulatory Commission (the Commission) having found that:
A.
The application for license filed by the Tennessee Valley Authority (the licensee) 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 and all required notifications to other agencies or bodies have been duly made; B.
Construction of the Browns Ferry Nuclear Plant, Unit 3 (the facility) has been substantially completed in conformity with Construction Permit No. CPPR-48 and the application, as amended, the provisions of the Act and the rules and regulations of the Commission; C.
The facilitywilloperate in conformity with the application, as amended, the provisions of the Act, and the rules and regulations of the Commission; D.
There is reasonable assurance:
(i) that the activities authorized by this operating license can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the rules and regulations of the Commission; E.
The licensee is technically and financially qualified to engage in the activities authorized by this operating license in accordance with the rules and regulations of the Commission; F.
The licensee has satisfied the applicable provisions of 10 CFR Part 140, "Financial Protection Requirements and Indemnity Agreement," of the Commission's regulations; BFN-UNIT3 July 02, 1976
G.
The issuance of this operating license willnot be inimical to the common defense and security or to the health and safety of the public; H.
The issuance of Facility Operating License No. DPR-68 is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied; and I.
The receipt, possession, and use of source, byproduct and special nuclear material as authorized by this license willbe in accordance with the Commission's regulations in 10 CFR Parts 30, 40 and 70, including 10 CFR Sections 30.33, 40.32, 70.23, and 70.31.
2.
The Atomic Safety and Licensing Board having dismissed the proceeding relating to licensing action in a "Memorandum and Order," dated November 27, 1973, Facility Operating License No. DPR-68 is hereby issued to the Tennessee Valley Authority to read as follows:
A.
This license applies to the Browns Ferry Nuclear Plant, Unit 3, a boiling water nuclear reactor and associated equipment (the facility), owned by the Tennessee Valley Authority. The facility is located in Limestone County, Alabama and is described in the "Final Safety Analysis Report" (Amendment 9) as supplemented and amended (Amendments 10 through 65), and the licensee's Draft Environmental Statement dated July 1971 and supplement thereto dated November 8, 1971, and the licensee's Final Environmental Statement dated September 1, 1972.
B.
Subject to the conditions and requirements incorporated herein, the Commission hereby licenses the Tennessee Valley Authority:
(1)
Pursuant to Section 104b of the Act and 10 CFR Part 50, "Licensing of Production and Utilization Facilities," to possess, use, and operate the facilityat the designated location in Limestone County, Alabama in accordance with the procedures and limitations set forth in this license; (2)
Pursuant to the Act and 10 CFR Part 70, to receive, possess and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended; BFN-UNIT3 July 02, 1976
(3)
Pursuant to the Act and 10 CFR Parts 30, 40 and 70 to receive, possess and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4)
(5)
Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form for sample analysis or instrument calibration or associated with radioactive apparatus or components; Pursuant to the Act and 10 CFR Parts 30 and 70, to possess but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.
C.
This license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
(1)
Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 3458 megawatts thermal.
(2)
Technical S ecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 218 are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
For Surveillance Requirements (SRs) that are new in Amendment 212 to Final Operating License DPR-68, the first performance is due at the end of the first surveillance interval that begins at implementation of Amendment.
For SRs that existed prior to Amendment 212, including SRs with modified acceptance criteria and SRs whose frequency of performance is being extended, the first performance is due at the end of the first surveillance interval that begins on the date the surveillance was last performed prior to implementation of Amendment.
(3)
Qeleted-.Amendment 212:-The licensee is authorized to relocate certain requirements-included in Appendix A and the former Appendix B to licensee-controlled documents.
Implementation of this amendment shall include the relocation of these requirements to the appropriate documents, as described in the licensee's application dated September 6, 1996, as supplemented May 1, August 14, November 5 and 14, BFN-UNIT3 Amendment No. 218 August 02, 1999
December 3, 4, 11, 22, 23, 29, and 30, 1997, January 23, March12, April 16, 20, and 28, May 7, 14, 19, and 27, and June 2, 5, 10 and 19, 1998, evaluated in the NRC staffs Safety Evaluation enclosed with this amendment. This amendment is effective immediately and shall be implemented within 90 days of the date of this amendment.
Amendment No. 218 August 02, 1999
Ck
(4)
(5)
(6)
The-Iisensee-shall-maintain-iR-effect-and4Aly-impiement-aII-previsiens-ef-the Qe
'ysisal-sesurity-plan-insIUdiRg-amendments-made pursuant4e4heautherity-ef-4Q-CAMO-.54(p).The-appreved-plan-,vanish sentains-infermatien-pre Nuslear-Piarit-Physisal-Sesurity-Pian
,"dat, etter-dated 4y-fVAl CA a,
. Amendment 214: TVAwill perform an analysis of the design basis loss-of-coolant accident to confirm compliance with General Design Criterion (GDC)-19 and offsite limits considering main steam isolation valve leakage and emergency core cooling system leakage.
The results of this analysis will be submitted to the NRC for its review and approval by March 31, 1999.
Following NRC approval, any required modifications will be implemented during the refueling outages scheduled for Spring 2000 for Unit 3 and Spring 2001 for Unit 2. TVAwillmaintain the ability to monitor radiological conditions during emergencies and administer potassium-iodide to control room operators to maintain doses within GDC-19 guidelines.
This ability will be maintained until the required modifications, ifany, are complete. This amendment is effective immediately.
74&i Ik Il 11m~
C pp QualifisatieR-Ran-,insluding-amendments-and-sha Ages-made-pursuant-te 40FR40%4(p)~h&Gppreve~Ua~IRIRg~ualifieatieRWarH5 identified-as-"Brewns-Ferry-Nuslear-Rewer-Station-Guar4 Trainirig-8 Qualifieatien-RaR,"dated'g,,
'sed-by-pages-dated
-,9 W;4S, 4IA~
may-subsequentiy-be-revised-IR-asserdanse-wit lj lthQ
[MAL~~14 4 I 4,'~N 4DCF~N W-,C4hy I~
. A 14:
Classroom and simulator training on all power uprate related changes that affect operator performance will be conducted prior to operating at uprated conditions.
Simulator changes that are consistent with power uprate conditions will be made and simulator fidelitywill be validated in accordance with ANSI/ANS 3.5-1985.
Training and the plant simulator will be modified, as necessary, to incorporate changes identified during startup testing. This amendment is effective immediately.
The licensee shall fullyimplement and maintain in effect all provisions of the Commission-approved physical security, guard training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The plans, which contain Safeguards Information protected under 10 CFR 73.21, are entitled: "Browns Ferry Physical Security Plan", with revisions submitted through May 24, 1988; "Browns Ferry Security Personnel Training and Qualification Plan", with revisions submitted through April 16, 1987; and "Browns Ferry Safeguards Contingency Plan", with BFN-UNIT3 Amendment No. 200 November 02, 1995
~
revisions submitted through June 27, 1986. Changes made in accordance with 10 CFR 73.55 shall be implemented in accordance with the schedule set forth therein.
r (7)
Browns Ferry Nuclear Plant shall implement and maintain in effect all provisions of the approved Fire Protection Program as described in the Final Safety Analysis Report for BFN as approved in the SEs dated December 8, 1988, March 6, 1991, March 31, 1993, November 2, 1995 and Supplement dated November 3, 1989 subject to the following provision:
The licensee may make changes to the approved fire protection program without prior approval of the Commission only ifthose changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.
BFN-UNIT3 Amendment No. 200 November 02, 1995
(8)
The-licensee-is-autherized-te-ternperaNy-stere-Iew-level-radieastive-waste-iR-aR p
'ti I
'IVf ~ly te-be-stered-shall-Re~ed-1320-series-ef-tetai-astivity
.This-aChe~eR expires-twe-yeaFs-frern-the-effestive-date-ef-this-arnendrnent-and-Is-subjest-taeN the-senditiens-and-restnstiens-in~s-applieatiert-date44anuary-2-1-,
4RK4Deleted D.
This-lisense-Is-subjeet4e-the4ellewlng-additienal-senditiens; iW W4 Pl N~'Pl pestuiated-less-ef-seeiant-assident
.Sheuld-the-Iisensee-deteRRIRe4hat4he results-ef-the-plant-unique-analysi~re-Bet-iR-senfeRRanse-with-the 4JD p
ytlN'~
'W te-the-Qernrnissien-.
This-aetio n-plan-will-Inside-as-a-rninirRUm-the-feiiewing-Infermatien-:
%-44 I '
sritena-.
~f-a-plant-hardware-rnedifisatien-is-fRade-,the~septanee-snteria-will4e dessribed-enmplant-unique-basis-.
Cl~'
'pl H' 9
hl~
dynarnie-leads-assesiated-with-a-pestuiated-I reas ef-design-fecund-Bet-raeeting-the-m'-iginai-d f 44 the-eriginal-design-safety-rRargiRs-.
BFN-UNIT3 Amendment No. 175 February 01, 1994
IRl-N' ~
pkN sea4HRment-leRg-teRR-preg ram-related-te-relief-valve-eperatien-,trike-sash medifisatiens-eR-a4imeiy-basi deslg en-subjested-te-relief-valve~peratieR-.Deleted (3)
T4e-fasility-may-be-medified-as-dessA bed-i Rt-URit-3 8
ttQ lgJj'"
peFt-and-iR-TVA-'s suppiementaI-letter-ef-Desember1~878-.
Deleted (4)
Gem edified-as-felkwvs-:
t~shment4,fer-item-II-.FA-4-and-II-.FAMshang VRit-2-staFtup-ia-Qy4e-6."Deleted E.
This amendment license is effective as of the date of issuance and shall expire midnight on July 2, 2016.
FOR THE NUCLEAR REGULATORYCOMMISSION S/ R. C. DeYoun for Roger S. Boyd, Director Division of Project Management Office of Nuclear Reactor Regulation
Attachment:
Appendices A 8 B-Technical Specifications Date of Issuance:
JUL21976 See Appendix 8 for additional License Conditions.
BFN-UNIT3 Amendment No. 212 July 14, 1998
APPENDIX B Deleted Amend-.
Number-AdditienaMenditiens 44 44 I
t 4I requiFements-insluded-in-Appendix-A-and4he feRRer-Append'esuments.lmplementatien-ef-this-amendment shaH-instude4he-releeatien-ef-these-requirements I 4 QQ~
- 4QQQ, I Ql t 44444444 l-l4-,
I QQ,~
QQQ44~
46,20,a...,nd-27,and-June 2-,5-,40-and48&MG-,evaluated-in4he-NRQ staff'-s-Safety-Evaluatien-enelesed~th4his amendment-.
Wis-amendment-is effestive-immediately and-shall-be days-ef-the-date-ef-this
%N-wHl-peFferm-an-analysis-efthe-design-basis less-ef-seelant-aseid en'-se nba-a-mmpHanse QII IQ '~QQ44C effsite-limits-sensideping-main-steam-iselatien valve-leakage-and-emergency-mre-see Hng system-leakage.The-results-ef-this-analyst-wN 4
t 44444~
QQ IQI M~A WQIQQ appreval-,any-requiFed-medifieatiens-vNl-be plemented-duping4he-refueHng-eutages ssheduled-fer-Spping-2000-fer-Unit-3-and-SpFing
~QQ.-QQI I
I i~
QQQQQ meniter-radielegieal-senditiens-during emergencies-and-administer-petassium-iedide-te sentrel-reem-eperateFs-te-maintain-deses-within QDQI-guide Hnes.This-abHity-vol-be WQ any,aFe-semplete.
effestive-immediately-.
BFN-UNIT3 Amendment No. 214 September 08, 1998
APPENDIX B Deleted QIassreeFR-and-sirnulater-traiRIRg-en-aii-peuer.
~I~I t S perfeFrnaR aWprated-eendNeRs.Sirnuiater-shaRges-that are-eensisteRt-with-peter-uprate-eendNeae-vN) be->vade-aRd-sin~r-Ade&pu4I-be-vaiidated-iR i~i I and4he-plaRt-sirnuiater-vNi-be-mediAed-,as Reeessa~e-IReerperate-changes-IdentiAed S..
BFN-UNIT3 Amendment No. 214 September 08, 1998
~ j ~
1
ENCLOSURE 1
TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)
UNITS 2 AND 3 TECHNICAL SPECIFICATIONS
( TS )
CHANGE 399 INCREASED MAIN STEAM ISOLATION VALVE (MSIV) LEAKAGE
RESPONSE
TO REQUEST FOR ADDITIONAL INFORMATION (RAI)
DATED NOVEMBER 23 ~
1 999 Below are responses to the nine NRC items from the subject RAI on TS-399.
The NRC questions are repeated along with the TVA responses for each item.
TS-399, which was submitted on September 28,
- 1999, proposed changes to the Unit 2 and 3
TS to increase the allowable leakage rate criteria for the MSIVs and requested an exemption to specific portions of 10 CFR 50, Appendix J to allow the exclusion of MSIV leakage from the summation of containment leak rate test results.
Enclosure 2 contains supporting calculations for the condenser seismic assessment associated with RAI Item 7.
Enclosure 3
provides addii ional details regarding RAI Item 8 which addresses specific NRC staff questions on dose calculation methods.
Commitments made in these responses are presented in Enclosure 6.
NRC ITEM 1 Section 5.2 of the March 3, 1999, safety evaluation for NEDC-31858, states that a secondary ALT path to the, condenser, having an orifice, should exist.
Your application states that in the event that FCV-1-58 were to fail to open, the leakage flow would split, with part of the flow going to the condenser via a 0.1875 inch diameter orifice in a normally open bypass around FCV-1-58, and the remainder going to the condenser via normal leakage paths through the main steam stop/control valves and through the high pressure turbine.
It is noted that NEDC-31858 para.
6.1.1(2) states that the ALT flow path should, based on the radiological dose methodology, be at least 1-square inch in internal cross sectional area.
Please describe the effect on offsite dose and control room habitability, of this single failure.
In particular, will dose consequences remain acceptable in the event of single-failure of FCV-1-58?
TVA RESPONSE TO ITEM 1 The BFN alternate leakage treatment (ALT) flow path is shown in Figure 3-1 of Attachment 4 of the September 28,
- 1999, TS-399 submittal.
The ALT path is from the outboard side of the MSIVs through Flow Control Valve (FCV)-1-58 to the condenser and satisfies the sizing requirements of NEDC-31858 paragraph 6.1.1(2) which states that the ALT flow path should, based on the radiological dose methodology, be at least 1 square inch for internal cross sectional area.
The orificed bypass path around FCV-1-58 shown in Figure 3-1 addresses Section 5.3 of the NRC safety evaluation dated March 3,
- 1999, which states that a
secondary path to the condenser, having an orifice, should exist.
This secondary path is considered a contingency alignment in the event of the unlikely failure of FCV-1-58 and is not sized to meet the 1-inch path provision discussed in the NEDC specified for the credited ALT path.
- Moreover, NEDC-31858 does not prescribe that a secondary ALT path be available which is fully redundant to the credited ALT path in terms of sizing.
\\
The failure of FCV-1-58 is unlikely to result from a loss of offsite electrical power.
For example, 2-FCV-1-58 is powered by 480-V Reactor Motor Operated Valve (RMOV) Board 2C.
RMOV Board 2C is normally aligned to 480-V Shutdown Board 2B which is Division II essential power.
The alternate feed to RMOV Board 2C is 480-V Shutdown Board 2A which is Division 1 essential power.
These 480-V Shutdown Boards have separate Emergency Diesel Generators (EDGs) as back-up power supplies through their respective 4160-V Shutdown Boards.
If the normal feeder (480-V Shutdown Board 2B) to RMOV Board 2C is lost, it can be transferred to its alternate power supply (480-V Shutdown Board 2A) by remote breaker operation.
Therefore, it is an easy operation to transfer 480-V RMOV Board C
to its alternate emergency power supply.
As noted above, the two 480-V Division I and II Shutdown Boards both have their own (separate)
EDG power supplies.
The power arrangement for 3-FCV-1-58 is similar.
Refer to the Final Safety Analysis Report (FSAR) Figures 8.4-1.b and 8.4.2 for a diagram of this electrical distribution arrangement.
For reasons stated above, it is highly unlikely that power will not be available to FCV-1-58 in the event of loss of offsite power.
As discussed in the response to NRC RAI item 4, FCV-1-58 will be periodically tested as part of the Inservice Test Program (IST) to ensure the valve is operable.
In addition, the functionality of the ALT path has been made highly reliable through the efforts to ensure the line is seismically rugged as discussed in the TS-399 submittal.
El-2
TVA considers that the proposed ALT path configuration using FCV-1-58 is consistent with the NEDC criteria to provide a
reliable ALT path.
TVA is also providing a secondary orificed contingency path in the unlikely event of a failure of FCV-1-58.
With the 0.1875-inch orificed path around FCV-1-58, it is calculated that the majority of MSIV leakage would still be directed to the condenser with a smaller remainder through the closed Main Steam Stop/Control Valves to the high pressure turbine.
The Main Steam Stop/Control Valves are currently in the preventative maintenance program.
As such, one Main Steam Stop and one Control valve is refurbished each outage.
Consequently, the Main Steam Stop/Control valves are refurbished once every 96 months.
These valves are tested each refueling outage for leak tightness and have historically been highly reliable.
Therefore, even in the unlikely event of the failure of FCV-1-58, the bulk of the MSIV leakage would still be routed to the condenser,
- hence, reducing potential control room and offsite doses.
NRC ITEM 2 Your application indicates that sealing steam supply valve, PCV 1-147, will be modified so that it fails closed instead of open.
Assuming that fails-open was the original "safe" fail
- position, please confirm that the new fail position will not adversely affect the capability to mitigate design basis accidents and other postulated events.
TVA RESPONSE TO ITEM 2 Pressure Control Valve (PCV) 1-147 is used during reactor startup to provide steam seals to the main turbine.
At higher reactor powers (above approximately 25% power),
the BFN turbine is self-sealing and PCV-1-147 is maintained closed by the valve controller.
The existing failure position (open) of PCV-1-147 presents an operational problem in potentially overwhelming the capacity of the seal steam subsystem to "unload" (self-regulate) the seal steam header pressure.
Therefore, in the event of an "open" failure, to continue normal power operation, it would likely be necessary to supplement the automatic seal steam unloader valves, PCV-1-148A and B, by opening the manual unloader valve, FCV-1-149, and/or by closing the high pressure steam supply
- valve, FCV-1-146.
The new failure position (closed) would present an operational problem only at low reactor powers (below approximately 25 percent power).
This would result in a slow loss of condenser vacuum if not corrected.
Low seal steam pressure is alarmed in
0
the control room and the associated Alarm Response Procedure e'd,rects the operator to open the steam seal bypass valve (FCV 145) t:o restore steam seal pressure.
This is a simple t:ask that can be performed from the main control room, and there is ample time to respond before condenser vacuum is lost.
From the above discussion it is seen that the failure of PCV-1-147 to either an open or closed position results in a operational problem dependent on the power level of the reactor.
Either end state is readily remediable by operator action.
Since the reactor is almost always at high power except for brief periods of start-up and shutdown operations, the new fail-closed mode is preferable from,an operational and safety point of view.
PCV-1-147 is not a safety-related valve and its operation is not currently assumed in the mitigation of design basis accidents (DBA) or transients.
Therefore, it is concluded that the new fail-closed mode to maintain the ALT boundary is satisfactory and does not adversely affect normal reactor operation.
NRC ITEM 3 Your application indicates that check valves are to be added to preheater steam lines to ensure ALT boundary integrity.
Please describe any proposed measures surveillance tests for these valves.
Does the use of these valves create a single-failure concern?
TVA RESPONSE TO ITEM 3 The subject check valves will be located in the steam supply lines to the Offgas Preheaters as shown in Figure 3-1 of Attachment 4 of the September 28,
- 1999, TS-399 submittal.
These new valves are within the current scope of the BFN American Society of Mechanical Engineers (ASME) Inservice Test (IST)
Program.
These check valves will be inspected and tested in accordance with the requirements for ASME Class 2 valves.
As
- such, these valves are nominally required to be exercised to their safety position (closed) once each quarter.
If quarterly or cold shutdown testing is not practical, the IST program allows that check valves may be disassembled and inspected each refueling outage as an alternative.
TVA has concluded that it is not practical to exercise these valves on a
quarterly or cold shutdown basis.
Position 2 of Generic Letter (GL) 89-04, Guidance on Developing Acceptable Inseryice Testing
- Programs, allows identical check valves to be grouped together (four valves per group maximum) and disassembled on a rotating basis (one valve each refueling outage) when normal testing is not practical.
Therefore,'ection XI surveillance testing will
0
consist of disassembly and inspection on a rotating basis (one t Heck valve each refueling outage) in accordance with Position 2
of GL 89-04.
The valves will also be verified to open after disassembly and inspection by proper operation of the Offgas Preheaters.
Regarding single failure considerations, these check valves are particularly well suited for this application of providing a boundary for the ALT path.
They are highly reliable and provide positive isolation through their design.
Alternate configurations such as fail-closed air operated valves and motor operated valves were considered, but were rejected in favor of the use of check valves.
Use of check valves is considered more reliable than air valves since operation of the check valve depends only on the system process (differential steam pressure),
and not the external devices such as controllers, solenoids,
- switches, etc.
In
- addition, a fail-close pneumatic valve would have a potential to negatively interfere with normal operations.
Motor operated valves (MOV) would be dependent on electrical power availability and relay logic.
- Hence, in this application, the use of check valves is considered the best choice of components which minimizes potential interferences with plant operation while providing high reliability for retention of the ALT path boundary.
As noted above, these check valves are within the scope of the BFN IST program and will be inspected and tested as described to provide assurance of proper component operation.
NRC ITEM 4 In allowing nonseismic piping to perform an engineered safety feature (ESF) function, it is expected that licensees will include the ALT system in the ASME Section XI inservice inspection (ISI) and inservice testing programs, and perform augmented ISI and motor-operated valve inspections in a manner consistent with ongoing ASME and approved risk-based programs applicable to ESF piping systems.
Please confirm if this is your intention.
Also, your application states that the most limiting single active failure would be failure of valve FCV-1-58 to open.
Please describe any augmented periodic testing (i.e.,
Generic Letter (GL) 89-10/GL 96-05 diagnostics) that will be performed on this valve.
TVA RESPONSE TO ITEM 4 The piping and components within the boundaries of the MSIV ALT path are considered to be within the scope of the BFN Section XI
,a~4
..1 0
- and, accordingly, will be inspected and
~~ested in accordance with the IST/ISI programs.
Additional detail is provided below for certain aspects of the program pertaining to the RAI questions.
The IST program will test the power operated valves within the ALT path boundary on a periodic basis.
The specific test requirements will be based on the function of the individual valve (e.g.,
passive versus active).
Testing of the check valves (considered active check valves) to the offgas preheaters is discussed in RAI Item 3.
Certain valves that serve as part of the ALT path boundary (for example, Main Turbine Stop and Bypass valves) are speci'fically excluded from the IST program in accordance with Regulatory Guide 1.26, Quality Group Classifications and Standards for Water-,
Steam-,
and Radioactive-Waste-Containing Components of Nuclear Power Plants.
Currently, some of these excluded valves are tested during power operations to ensure their functionality and are in the preventive maintenance program for periodic refurbishment.
These valves will be included as part of the IST program, but as non-Code valves.
The ALT path boundary piping does not meet the criteria for inclusion in the augmented Intergranular Stress Corrosion Cracking weld inspection program.
This piping is, however, part of the Flow Accelerated Corrosion (FAC) program which periodically monitors pipe wall thickness degradation.
FCV-1-58 was considered for inclusion in the augmented MOV test programs such as GL 89-10, Safety-Related Motor-Operated Valve Testing And Surveillance, and GL 96-05, Periodic Verification of Design-Basis Capability of Safety-Related Power-Operated Valves.
The design basis for establishing the ALT path is a Loss-of-Coolant Accident (LOCA) with assumed major core damage and the MSIVs closed.
With the MSIVs closed, the ALT path boundary is physically isolated from the reactor vessel and primary containment except through leakage through the MSIVs.
The ALT path piping will depressurize through the orifice around FCV-1-58.
In order to establish the ALT path, FCV-1-58 will not have to open against a large differential pressure and the post-accident system conditions will be less severe than the conditions which the FCV-1-58 valve would experience during IST testing during normal power operations (with full reactor pressure).
Therefore, the FCV-1-58 will not be included in the GL 89-10/96-05 MOV program since the periodic IST program testing on this valve is considered to be adequate to ensure its functionality.
E1-6
NRC ITEM 5 Section 4.1.2 of your EQE Report identifies the load combinations and stress allowables utilized in seismic assessments.
Please provide a discussion of the extent to which the criteria used are consistent with the licensing basis requirements for other engineered safety features.
TVA RESPONSE TO ITEM 5 The load combinations and stress allowables utilized in the seismic assessments for the resolution of outliers and the evaluation of ALT piping, related components, and supports as presented in Section 4.1.2 of the EQE Report (Attachment 4 of the September 28,
- 1999, TS-399 submittal) are consistent with plant licensing basis requirements used to address Class II piping, and pipe supports and components for pressure boundary integrity and position retention at BFN.
These seismic evaluation criteria are contained in TVA Design Criteria BFN-SO-C-7306, Qualification Criteria for Seismic Class II Piping, Pipe Supports and Components, which is Reference 9 of the EQE report.
The objective of the subject seismic assessments was to provide assurance that the ALT pathway would maintain pressure boundary integrity and would not be adversely affected by such factors as (1) differential displacements of structures, equipment, and piping (2) pipe support integrity issues and (3) seismic interaction issues such as the impact of piping with equipment, structural features, and other piping.
Additionally, valves that are classified as active in establishing the ALT path must be functional following the Design Basis Earthquake (DBE) and were evaluated in accordance with the General Implementation Procedure (GIP) methodology as referenced in Section 4.1.2 of the EQE report.
Qualification in accordance with GIP provides reasonable assurance the required valves will be functional.
The loading combinations and stress allowables utilized in the design or assessment of Class I systems (ESFs piping, pipe
- supports, components, etc.)
are described in Appendix C of the BFN FSAR, Structural Qualification of Subsystems and Components.
These requirements are specified in TVA Design Criteria BFN-50-C-7103, Structural Analysis and Qualification of Mechanical and Electrical Systems, for Class I piping and tubing, in TVA Design Criteria BFN-SO-C-7107, Design of Class I Seismic Pipe And Tubing Supports, and in TVA Design Criteria BFN-50-C-7105, Pipe Rupture, Internal Missiles, Internal
- Flooding, Seismic Equipment Qualification and Vibration Qualification of Piping, for Class I and Class II equipment.
The load combinations and stress allowables for ESFs were developed El-7
p~
to assure not only pressure boundary integrit:y and position
~etention, but also for full functionality of equipment following a
DBE.
In summary, the load combinations and stress allowables used for the ALT path seismic assessment discussed in the EQE Report are based on assuring that the system will maintain pressure boundary integrity and position retention and, in some cases for valves, maintain functionality.
Since the main steam piping system housed in the Turbine Building was not originally designed t:o include seismic loading, a seismic verification walkdown to identify potential piping concerns was performed of the leakage pathway to provide assurance that pressure boundary integrity and position retention would be maintained.
The load combinations and stress allowables in Section 4.1.2 are the bases used to
- resolve, by calculation, or maintenance/modifications, all identified outliers.
These resolutions are summarized in Tables 4.1 and 4.2 of the EQE Report.
NRC ITEM 6 Referring to Page 10 of the EQE Report, and noting that different Class I buildings at Browns Ferry Nuclear Plant have different vertical soil amplification factors, please explain the basis for the specific scaling factors selected for the Turbine Building.
TVA RESPONSE TO ITEM 6 The methodology to determine the soil amplification factors for the various Class I structures at BFN is defined in TVA Design Criteria BFN-50-C-7102, Seismic Design, which requires that structures founded on soil consider soil amplification.
The soil amplification factors for applicable Class I structures are shown in BFN FSAR Chapter 12, Structures and Shielding.
The horizontal soil amplification factors range from 1.0 for rock-founded structures such as the Reactor Building to a maximum of 1.6 for soil-founded structures such as the Diesel Generator Building (DGB).
Similarly, the vertical soil amplification factors range from 1.0 to 1.3.
Seismic demand for equipment in a particular structure is determined by scaling the site design basis response spectrum, i.e.,
the Housner spectrum for 5% damping and anchored at 0.2g, by the appropriate horizontal and vertical soil amplification factors.
Since the Turbine Building is designated as,a Class II structure in the
- FSAR, no soil amplification provisions were originally specified and no dynamic seismic analysis results were available to define seismic demand on t:he structure or components.
It was determined that: the soil amplification factors for the DGBs would be most representat:ive for the Turbine Building.
The foundation E1-8
materials are similar as are the foundation depths.
In addition,
@he DGB horizontal soil amplification factor of 1.6 is known to be conservative, so this conservatism will be extended to the specification of the seismic demand for the equipment in the Turbine Building for the seismic evaluation.
The primary foundation difference is that the Turbine Building is supported on steel H-piles to bedrock.
However, it is considered that the primary effect of the pile foundation would be to increase the foundation stiffness in the vertical direction relative to a similar foundation without piles.
Therefore, the horizontal soil amplification for the Turbine Building would have a more significant effect than that of the vertical in the overall seismic verification efforts.
Accordingly, seismic demand for equipment in the Turbine Building and for the seismic assessment of components is based on the same horizontal soil amplification factor of 1.6 and vertical soil amplification factor of 1.1 as was used for the DGBs., These factors were used to scale the BFN design basis DBE response spectrum (0.2g Housner spectrum, 5% damping) to determine seismic demand.
NRC ITEM 7 In Table 4-8 and Figures 4-2 thru 4-5 of the EQE Seismic Evaluation Report, only Moss Landing Units 6
7 condensers are provided for comparison with the Browns Ferry condensers.
This is too limited to support a finding that the earthquake experience database demonstrates the seismic adequacy of Browns Ferry's condensers.
Please provide additional condenser data.
As stated in the staff's March 3, 1999 safety evaluation, there is no standard at the present
- time, endorsed by NRC, that provides guidance for determining the required number of piping and equipment items, that should be referenced in the earthquake experience database when utilizing the BWROG methodology.
Therefore, you are responsible for ensuring the sufficiency of the above data submitted for staff review and determination.
If sufficient data are not provided for the condenser, the NRC may require that the condenser be analytically evaluated against all the pertinent operating and design loadings, in accordance with the plant's design basis methodology and criteria.
TVA RESPONSE TO ITEM 7 The BFN condenser design attributes are shown to fall within the bounds of the Moss Landing database site as discussed in Section 4.4 and depicted on Table 4-8 of the EQE Report.
To provide additional assurance that BFN condensers would maintain structural integrity, a specific analysis was performed on the
0
condenser subject to BFN seismic demand.
Results of the analyses
'demonstrate that the condenser shell stresses are small, with maximum stress ratios based on American Institute of Steel Construction (AISC) allowables of 0.12 for combined axial and bending and 0.10 for shear (Reference section 4.4 of EQE Report).
Additionally, the condenser anchorage was also compared with the performance of condensers of the database site.
The anchorage was demonstrated by seismic experience and by analytical methods to be acceptable.
Maximum stress ratios from the condenser support anchorage evaluation including BFN seismic
- demand, based on AISC allowables, are 0.70 for bolt tension in the perimeter support feet and 0.86 for shear in the center support built-up section
(
Reference:
Section 4.4 of EQE Report).
Based on the above, it was concluded that the BFN condensers were acceptable.
Refer to Enclosure 2 for a copy of the calculations used to determine the stress ratios given above.
NRC ITEM 8 The radiological analysis description provided in the application does not provide an adequate basis for the staff to determine whether or not those analyses are acceptable.
The staff notes that the reported increase in doses appears to be inconsistent with the proposed eight fold increase in the allowable MSIV leakage.
Please provide the analysis assumptions,
- methods, and input parameters used in your calculations, in sufficient detail for the staff to resolve the apparent inconsistency and, if deemed necessary by the staff, to perform independent calculations to confirm your reported results.
Your response should identify any changes made to the assumptions,
- methods, and inputs used in analyses previously approved by the NRC for Browns Ferry Units 2 and 3.
TVA RESPONSE TO ITEM 8 Refer to Enclosure 3 for a response to this item and to additional NRC staff questions on dose methodology.
After further review, we agree that the linearity assumption used in TVA's initial calculation is not always conservative.
Therefore, TVA reperformed the MSIV leakage dose calculations rather than use extrapolation factors to determine the MSIV leakage contribution to dose.
These were completed in accordance with the NEDC methodology as reviewed in the NRC SER.
This recalculation resulted in a reduction of the requested MSIV allowable leakage rate.
Therefore, TVA is providing an amended TS change request as part of this response (See Enclosures 4 and 5).
E1-10
NRC ITEM 9 V
Your application requests an MSIV leakage be included the addition to the 0.6 L, limit penetration leakage).
Is it consistent with NEDC-31858?
exemption?
exemption from the requirement that overall Type A leakage limit (in for the sum of Types B and C
your understanding that this is Is there a valid need for this TVA RESPONSE TO ITEM 9 10 CFR 50, Appendix J testing ensures primary containment leakage following a design basis LOCA will be within the allowable leakage limits specified in plant TS and assumed in the safety analyses for determining radiological consequences.
For BFN, the acceptance criteria for the Type A test Containment Integrated Leakage Rate Test (CILRT) is 0.75 L for return to power following performance of the CILRT.
This limit is shown in BFN TS 5.5.12, Primary Containment Leakage Rate Testing Program.
The 0.75 L
acceptance criteria allows for a 25% margin for degradation during plant operation.
The, CILRT currently includes leakage through the closed MSIVs.
The proposed increase in MSIV leakage, if not excluded from the 0.75 L, acceptance criteria for the CILRT, could account for approximately 18% of the 0.75 L, acceptance criteria and significantly reduces the margin available for all other primary containment leakage paths.
Inclusion of MSIV leakage in the CILRT would effectively reduc'e the CILRT acceptance criteria to approximately 0.62 L,.
In analyzing the use of the ALT path, the radiological consequences of MSIV leakage are being determined separately from other primary containment
- leakage, since MSIV leakage is released directly into the Turbine Building, which is not treated by the Standby Gas Treatment System.
The MSIV leakage rates are measured as part of the 10 CFR 50, Appendix J Program, to verify this leakage will not exceed the proposed maximum leakage in the TS and assumed in the safety analyses for radiological consequences.
Therefore, since the effects of MSIV leakage are being explicitly accounted for in the dose analysis, it is appropriate that MSIV leakage be excluded from the Type A testing results.
Exclusion of MSIV leakage from the Type A test acceptance criteria is necessary to provide adequate margin for leakage of the remaining primary containment leakage paths tested during the CILRT.
This exclusion is justified because of the separate treatment of MSIV leakage as previously discussed.
The radiological consequences of primary containment leakage and MSIV leakage will continue to be maintained within allowable limits
and the intent of 10 CFR 50, Appendix J will continue to be hatisfied.
NEDC 31 858PI Section
- 6. 3. 2. 1, discusses the need for Appendix J exemptions for both Type A and Type C tests.
Therefore, the exemption request is consistent with the NEDC.
E1-12
a~s -0
ENCLOSURE 2
TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)
UNITS 2 and 3
PROPOSED TECHNICAL SPECIFICATIONS (TS)
CHANGE TS-399 INCREASED MAIN STEAM ISOLATION VALVE (MSIV) LEAKAGE
RESPONSE
TO REQUEST FOR ADDITIONAL INFORMATION (RAI)
DATED NOVEMBER 23 I 1 999 CONDENSER SEISMIC STRESS CALCULATION
~
Excerpt from calculation CD-N0001-980038, R1 Main Steam line Ruggedness Shows Condenser Anchorage Calculation
~
Excerpt from calculation CD-N0001-990113, RO Seismic Evaluation Report Shows Condenser Shell Calculation
'h J'n I
0
gA Record TVANCALCULATIONCOVERSHEET Title Main Steam Seismic Ruggedness Evaluation Page i
of Plant BFN Unit 3
Preparing Organization CE-Civil Branch/Project Identifters CD-N0001-980038 Key Nouns (For RIMS)
Seismic, Com onent Qual, Pi in, Pi'e Su ort Each time these calculations are issued, preparer must ensure that the original (RO)
RIMS accession number is filled in.
Applicable Design Document(s)
BFN-50-C-7100, BFNZO-C-7102 BFN-50-C-7306 Rev RO R
)
(for RIM RIMS Accession Number RI14 9809 15 10 6 14 9 9093 SAR affected:
QYes HNo Section(s):
UNID System(s) 001, 006, 008, 012, 043, 071, 073, 303 Rev 0 R
'I Design Change Document No.
T40871A TAIICBi P
par Qualmy Related?
Yes No H
CI Yes No 0
H Safety ratatedT These calculations contain Yes No unverified assumption(s) that R
i d
0-/z-E wed 0-/'Z-)If p cA mar oc sefresf
&Iaglf Vlf These calculations contain spa@at requirements and/or These calculations contain a design output attachments Yes No 0
H Yes No 0
H Ap ov Date Calculation Revision.'ntire Catcutafion Selected pages Not Applicable El00 Statement of Problem:
The Main Steam piping dovmstream of the outboard MSIVs Is needed to be capable ofwith standing an earthquake so that any leakage through the MSIVs from the Reactor side can be contained and diverted to the main condensers.
A walkdown was performed by EQE to verifythe seismic adequacy ofthe piping. Problems found during the walkdovm were identified as outliers. The outliers which were found acceptable as4s by performing detailed engineering evaluations are documented in this calculation. Also included in this calculation is the EQE report documenting the outliers and the final resolution for each item. The outliers that required a plant modification as a solution are documented in calculation CD-N0001-980039.
Abstract This calculation is a collector for the reports, calculations, etc. done by EQE International for the seismic ruggedness verification of the BFN Main Steam piping In the Turbine Building. The scope of the seismic ruggedness walkdown performed by EQE was based on preliminary isolation boundary locations as shown in Table 2-2 and Figure 2-1 of the 'Summary Report" contained in this calculation as Attachment A.
A review ofthis preliminary isolation boundary has shovm that the valves at these locations may not close as desired.
Once the isolation boundary review is complete, any additional walkdovms determined necessary to ensure the seismic ruggedness of the Main Steam piping willbe done in support of DCN T41019A.
Catcut<kion C4xmiklcckm~ -'Q tg'i Note: This calculation is to provide a retrievable source for the summary report and supporting calculations performed by EQE International, Inc. TVAsignatures are not attesting to the technical accuracy of the included documents of this calculation.
Tot~i p cg
. &t4-&e pgq (tent 0
Microfilmand return calculation to Calculation Library.
Address:
0 Microfilmand return calculation to:
0 Microfilmand destroy.
TVA40532 (08-97)
Page i NEDP-2-1 [0845-97)
I I I l
gggig
~
l.
i'll
~
0 r
~
0 Ile 0
T~
~
~
h~ CALCULATIONCOVER SHEET Calculation No.
200621-~09 Project:
TVABFN-3 MSIVOUTLIER RESOLUTION Calculation
Title:
CONDENSER SEISh1IC EVALUATION
References:
Appendices:
SEE SECTION 3.0 NONE Total Number of Pages (Including Cover Sheet):
Revision Number Approval Date Description of Revision Originator Checker Approver 9/08/98 ORIGINALISSUE S.P. HARRIS
'j J.O. D)ZON
,0.
J.O. DIZON J~Q'
'ttachment HL aa(
-R 38am. D Sheet Ho. ~g Of
MEINTERNATiONAL, CALCULATIONSHEET JOBNO.
QQQ$~1 JOB ABF -SMSV R
S I
CALC NO. ~~
SUBJECT Condenser Seismic Evaluation SHEETND~
BY~SDATE~i~l9 E CHK~0 DATE~908/98 TABLEOF CONTENTS 5.0 PURPOSE AND SCOPE......
2.0 METHODOLOGY
~ ~ ~
~PAG 3
3.0 REFERENCES
~ ~ ~
~ ~
~ ~ ~ ~ &~ ~ ~ ~ ~ ~ ~ ~
~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~
4.0 TURBINE CONDENSER CHARACTERISTICS COMPARISON.......,..........
5.0 ANCHORAGE EVALUATION...
6.0 CONCLUSION
S.
Figure 4-5 Size Comparison....
f!gure 4-2 Weight Comparison.
Figure 4-3 Height Comparison.......
Figure 4-4 Plan Dimension Comparison.............
Figure 4-5 Schematic Plan of Condenser Anchorage.......,.
Figure 4-6 Anchorage Comparison Transverse to Turbine.............
figure 4-7 Anchorage Comparison Parallel to Turbine.
Figure 4-8 Design Basis Spectra Comparison with Database Sites,..............
figure 5-I Anchor Design...
~..
Table 5-1 BFN Condenser Weight and C.G.
Table5-2 CondenserLoads Table 5-3 Anchor Demands...
18 20 xhbfnPmSiv'I9:alc09.dOC Attachmeiit Ho.
calC. E9 + O.'Lg - ZdldÃEEE.~-
Sheet No, of
I:~~Jg EOE NTERNATIONAL CALCULATIONSHEET JOB NO. ~6/~1.
JOB TVABF -
V TLIER R LUTI N CALC MO. ~op SUBJECT ondenser Seismic Evaluation SHEET HO 3
BY SPH DATE~I~3 OHK~OD DATE~9/a///8 1.0 PURPOSE AND SCOPE The purpose of this calculation is to verify the Browns Ferry (BFN) condenser seismic capacity and resolve Outlier 13-1, as identified during the BFN-3 MSIYSeismic Yerification Walkdown (Reference 2).
2.0 IVIETHODOLOGY The SFN Unit 3 condenser is evaluated using seismic experience data from past earthquakes and ongineering analysis.
Seismic capacity versus demand is evaluated by comparing the BFN Unit 3 condenser with condensers in the seismic experience database that have experienced strong ground motions in excess of the BFN-3 Design Basis Earthquake (DBE). Condenser size, construction, design and anchorage characteristics are summarized and compared with parametors of earthquake experience condensers.
Anchorage evaluatIon methodology used is consistent with that described in the Generic Implementation Procedure (SIP, Reference 1) and standard structural engineering practices.
xAbrnpmsiv'calc09.doc llttacltmettt L'O.
Clic. >l//.l -
08> - Pii EV.+
E//.ae En.
J 4 at
EQE iNTERNAT[ONAL CALCULATIONSHEET JOB NO. 7QgQ?/~0 JOB A B M
V I
L N
CALC NC. ~00 sUBJEGT Cond nss sis io EY Usfi SHEET NO.
4 BY BPN DATE
/B~g CHK~f DATE~00/gg
3.0 REFERENCES
1.
"Generic implementation Procedure (GIP) for Seismic Verification of Nuclear Plant Equipment", Rev.
2A, March 1993, prepared by Vlinston &Strawn, EQE, et al., for the Seismic Qualification UtilityGroup (SQU6).
2.
BBFNP - Unit 3 MSIVSeismic Verification Vff'afkdown RePort", ECIE RePort No 50147.08-R-001, Draft, September 30, 1995.
3.
BWROG Report for increasing MSIV Leakage Rate Limits and Elimination of Leakage Control Systems, General Electric NEDC-31858P, Rev. 2, September 1993.
4.
AISC Manual of Steel Constructfon, 8th Edition S.
Seismic Verification of Nuclear Plant Equipment Anchorage (Revfsion 1), Volume 1: Development of Anchorage Guidelines, EPRl NP-5228-SL, June 1991.
e.
TVAand Vendor Documents:
- a. Misc. Steel Turbine Foundation Embedded Parts Sheet 2
- b. Outline of Shell 3A for 666,000 Sq. Ft,...Surface Condenser
- c. Outline of Shell 38 for 666,000 Sq. Ft....Surface Condenser
- d. Outline of Shell 3 for 666,000 Sq. Ft......Surface Condenser
- e. Instructions for the Care &Operation of Surface Condenser &Accessories
- f. Arrangement of Condenser Supports 8, Anchor 48N840 R12 3-93-621-5-1 A 3-93-621 2A 3-93-621-5-3A BFN-VTD-F175-0050 93-505-3-190 Rev E 7.
TVABrowns Ferry Nuclear Plant Final Safety Analysfs Fleport (FSAR).
s.
TVACalculation No. CD-N0001-980039, "Main Steam Seismic Ruggedness Verification'.
xhbfnPmsivfff:afc09.dof'.
Attachment ffo.
Cafe. No. '
Dg/-9J 03gnp.g Sheet No.
sf
pO,
EOE INTERNATIONAL
~~~~~
JOB NO. ggQQ1.0$
JOB TV CALO NO. ~C-00 SUBJECT CALCULATIONSHEET MSIV 7
)
RE TION Condenser Seis lc v
SHEET NO 8
BY SPH OATE~gjf'98 CHK JOD DATE ~9,'08'98 4.0 TURBINE CONDENSER CHARACTERISTICS COMPARISON The High Pressure condenser has a heat transfer surface area of 222,000 ft. In Table 3-], the design attributes of the condenser is compared with the two sites in the earthquake experience database that have condensers most representative of BWR type condensers:
Moss Landing Units 6 5. 7, and Ormond Beach Units 1 8 2.
The shell of the condenser is constructed of ASTM A-265 Gr. C steel plate. The database condenser shells are ASTM A-285 Gr. C steel plate. Some of the overall heat transfer area, weight, and footprint of the condenser are enveloped by the database condensers, as shown In Figures 4-1 through 4-4.
in summary, the condenser design and anchorage are similar to those at facilities in the earthquake experience database that have experienced earthquakes in excess of the Browns Ferry design basis OBE (See Figure 4-8), Appendix D, Section 4.1, of NEDC -31858P (Reference 3) contains details of the earthquake experience for condensers.
Specific data used in the evaluation are as follows:
4.1 BFNP-Unit 3 Condenser Design Basis 4.1.1 Design Code Heat Exchanger institute (HEI) Standards 4.14 Hydrostatic Test Requirements Shell - Completely filled with water xhbfnpmsivhcnlc09.doc eehrlieIIt Ro.
dO/
03 II@.
Sheet Ho,
EGE iNTERNA.IONAL JOB NO. $000021. Il JOB CALC NO.
C-Ogg SUBJECT CALCVLATlONSHEET H-M 1V i
R RESOL I
Condenser Seismic Evaluation SHEETNO~
BY SPH DATE~f~QQ CHKegg DATE 9f08/98 4.1.3 Anchorage The condenser anchorage is shown schematically in Figure 4-6. The condenser has six plate supports with (2 or 3) 2" or 2a " diameter anchor bolts each.
Each anchor bolt has greater than 5'ominal length with approximately 48" embedment in the turbine building foundation. The supports are designed to rcsi-t vertical operating loads. Thermal growth of the condenser occurs from the fixed point near the center of the base.
The sliding plate supports have slotted holes allowing thermal growth radial to a fixed center support pad. The center support consists of a built-up H section, embedded 4'nto the turbine base mat and welded to the condenser bottom.
Size Comparison of Hrowns Feny Unit 3 Condenser with Database Condensers BFNP-3 Ormond Beach 21,0,000 Mass Landing 0
50,000 100,000 150,000 2CC.300 250.000 3C0,000 350.000 400,000 4M.COO Heat Transfer Area (ft2)
Figure 4-1 Size Comparison of Browns Fer~ Unit 3 Condenser with Database Condensers xhbfnpmsihcalc09.doc Attachment Ho.
T Gale. Ho. 8 0
4 REV.+
Sheet No.
of
EOE INTERNATIONAL JOB NO. gggggl.01 JOB A
CALC NO. ~P SUBJECT CALCULATIONSHEET V
IER R TION Condenger SeiSIn(C EV
(
tion SHEET NO.
'7 EY~PDATE OOII/98 CHK~~DATE OIOI/OIS Comparison of BFN-3 Condenser with Database Condensers BFNPC 2,070,000 OrnTond Beach 1.767.500 0,115, 0
500s000 1 i000s000
'.500eppp 28000o000 2+00,000 31000I000 3 6009000 Weight (ibs)
Figure 4-2 Weight Comparison of BFN-3 Condenser with Database Condensers Comparison of BFN-3 Condenser with Database Condensero BFNPN 47 Orrncnd BeaCh Moss Landing 47 0
6 10 15 20 25 30 35 40 45 60 Height (fest)
Figure 4-3 Height Comparison of 8FN-3 Condenser with Database Condensers x:IIbfnpmsiv>calc09.doc I(ttachiIIoftt HO.
+000-JOSREII 0 Shoat No.
F of
EQE Itd"rERIIIATIOIIIAL JCB No. S((0337.07 JCB CALO NO.
C-0(OCI SUBJECT CALCULATIONSHEET
~ MIV I
R ON Condenser Seismic Evaluation SHEETNO~
BY~h(
DATE ~nfgl 9 OHKatoD DATE 9~lo aE
,'dt~+PI g
I El
.IK Rl I
Il I stnss landing 3 3 7 (eeft x 33 tl) almond Eoach (52ftx 27ft)
Q BrnwnsFsnTUnit3 (enftx3eftf Anctttfrhntts entete sent(cd BOSS dt ected Item coTot Bnc!Norptsto ISUrspong A1 C)
AnchDr tf(NBhe(Ul sleet(ed hots Dsfpsftctcctkt (5ftppCttg 6 4 DI fffCSd Encncr pthto Figure 4W Comfhartson of Browne Ferry uni!3 condenser Ran ctmensicns wiIh Databaso Ccndensers Figure 4-5 Schematic Phn V!ow of Orcfwns Ferry Unit3 Contfamscr Anchorage x."(bfnpmsthcatc09.doc Attachment NIL Gale. Ng.(P k8/jig/ g ltElf Q Sheet f(ts J
ef
EQE INTERNATIONAL JOB NO. gQQg21.01 JOB TV GALC ND. r0009 SUBJECT CALCULAT(ONSHEET
~
MSlV l
I RRE 0 UTION Condenser Seismic Evaluation SHEET NO.
9 BY SPH DATE 8/31/98 CHK~JD DATE~~8 Comparison of BFN-3 Condenser Anchorage with Database Condensers 0.0002 0
E 0.0001 0
Vl Moss Landing El Centro BFNPN Supper Bound 8 Lower Bound Comparison of BFN-3 Condenser Anchorage with Database Condensers E
0.0002 D
O 0.0001 0
L ra Moss Landing El Centro BFNP-3 0 Upper Bound tent Lower Bound Note (1): Shear Area(in )/Demand(condenser weight x g level)
Figure 4-6 and 4-7 Comparison of BFN-3 Condenser Anchorage with Database Conderisers xAbfnpmsivMalc09.doc Att2chrnerit Q.
Cafe. Ho.g4 rf/g gggiiot Sheet Na.
/V of
EQE INTERNAT(ONAL CALCULATIONSHEET JOB NO. 20/66 gf JOB TVA FN M IVO RRE lON CALC NO. ~9 SUBJECT n
r Seismi Evaluation SHEETND~
SYLPH DATE~ELIIE8 DHK JDD DATE~/E//9I The condenser anchorage was compared with the performance ot similar condenser ln the earlhquake experience database.
The shear area of the condenser anchorages, divided by the seismic demand was used to compare condenser anchorage with condensers in the earthquake experience database (See Figures 4-6 and 4-7). The values for the BFN-Unit 3 condenser are as follows:
Parallel to Turbine Generator Axis Shear Area (ln )/Seismic Demand Lower Bound 0.0000604 Transverse to Turbine Generator Axis 0.0000966 The condenser anchorage shear area to seismic demand is substantially greater than the selected database sites (see NEDC 31858P, Reference 3, Figures 4-10 and 4-11 of Appendix D)
, 4.1.4 Manufacturer:
Foster Wheeler 4.1.5 Surface Area, Weight, Dimensions Surface Area: condenser has 222,000 fP Weight: condenser weighs 2,070,000 ibs operating.
Dimensions: condenser is 50'ong, 32'ide and 47' high.
4.1.6 Type: Base supported, rectangular.
4.1.7 Shell Material Material: ASTM A-285 Gr. C steel 7/8 thick xhbfnpmsivknlc09.doc
/tttacftmprit gp.
Cplp. Mp,P
/Qyp piggy
EQE INTERNATlON/8t.
CALCULATIONSHEET JOB NO. 2~00 ~11 JOB TV BF IER R CALC NO.~~
SUBJECT Condenser i
c Evaluati SHEET NO~I BY~~DATE~8/3t/ 8 CHK~JO DATE~9/08/9 4.1.8 Tube Design Material: ASTM 8-111 inhibited Admiralty Tubes: ASTM B-111 inhibited Admiralty, 7/8"-1 8 BVIG.
The effective tube length is 49'T/4".
The condenser design and anchorage are similar to those at facilities in the earthquake experience database that have experienced earthquakes in excess of the BFN design basis DBE and the BFN condenser is bounded by the comparable attributes of '.he database condensers.
x:hbfnpmsiv4alc09.doc Attachment No.
I;>ic. Ifo. k-4+Cl-IOMIIElt.0 Sheet Ho. XZ ef
I~+
EQE INTERNATIONAL JOB NO. ~200 ~1 JOB CALC NO. ~Qg SUBJECT CALCULAT!ONSHEET I
OUTLlER R UTi N C
r Seismic Evaluation SHEET NO.
8Y SPH DATE~8/31/g CHK JQD DATE g~lo 1//t 1.2 C0 Cl g 0.8 Qa 0.6 0.4 0.2 0
I
~
~ ~ Voo
~
P
~ e 4o r
I
~ /
~ o ~ ~ ar
~
~ ~ woe ~
I I
~ 4
~
I I
No~sN/1:
'pa values
~ Moss Landing Valley Steam
~ El Centro I
~ Coohvater
- ~ +
~ - Bulk Mail
~RID Dell
~ Humbolt '75(1)
~ Humbolt 80(1)
-"+--Humbolt '92(1)
---~ -- Glendale(13
~ Browns Ferry Frequency 100 Figure 4-8 comparisons of Database Site Spectra to Browns Ferry Unit 3 DB= Ground Spectra x:tbfnpmsiv'ealc09.doc Attachment Ho.
Calo. fto.
P700b - Bd9ZRg,+
Slteat Ho.
of
EOE INTERNAo.ORAL CALCULATIONSHEET JOB NO. ~8~1~
JOB TVA I=NN M IV I
l N
CALCND. C-!!LS SUBJECT cndsnss/ Ssis/nic Ev 1
SHEETNC~
BY~0P DATE~s/31 3
CHK JQD DATE
! //Cff/3 6.0 ANCHORAGE EVALUATION e
ine In ut Seismic An evaluation of the condenser support system was performed using seismic demand determined by the GIP method for the turbine building at the condenser foundation elevation.
Condensers are large structures that are by design very stiffto withstand operating vacuum and hydro-static test loads. The condenser shell design utilizes a steel shell, stiffened with integral plate stlffeners and structural members.
The predominant seismic response of the condenser Is expected to be rigid body dynamics. The calculations are performed for the frequency range of the zero period acceleration (ZPA).
The condenser Is mounted on the base mat at elevation 662 feet. Design spectra were not available for the Turbine Building at this elevation. Accelerations for the condenser analysIs were based on the site design spectra zero period acceleratIon of 0.2g. This value was Increased by a factor of I.6 In accordance with the BFN FSAR (Reference 7) to account for site amplification for soli-founded structures (see also Reference 8).
Determl e ve oments and Base Sh ar Established weights, center of gravity, overturning moments and base shears for the condenser are calculated and shown in Table 5-'I. Overturning moments and base shears are established using conservative methods for combination of directional and load components.
The condenser is symmetric about two axes and eccentricities are expected to be very small.
Dynamic frequencies for the condenser are complex due to the stiffened plate nature of the construction and the complicated internals, and are not known. The condenser was assumed to behave in a rigid manner and respond to the ZPA.
Overturning moments are calculated assuming that plane section remain plane and materials are assumed to be elastic. Mcments and base shears are shown in Table 5-2.
x:fibfnpmsiv'I/ealc09.doc Attachriiellt Ho.
Caic. Hp.
Sheet Hp.
N Am. ri of
+t
EQE:NTERNATlONAL JOB NO. ~20 ~01 JOB A
CALC NO. g:-0(~
SUBJECT CALCULATlONSHEET IV Condenser Seismic Ev GIREET NO 14 BY SPH DATE 1~/3 gB CHK~CD DATE~9Rl8/9 Su ort Forces and Ca acities Support forces and capacities are calculated and shown in Table 5-2. Total tension forces due to overturning moments in N-S and E-W directions of earthquake excitation are combined by the Square Root of the Sum of the Squares (SRSS) rule and are added absolutely to the operating loads. Operating loads are taken as the sum of vacuum loads and dead load forces on the anchor pads from Reference 6f.
Tension forces are resisted by the six support pads around the condenser perimeter. Base shears are calculated separately for N-S and E-W directions of earthquake excitation and compared with shear capacities.
Base shear ls resisted by the anchor support at the condenser center.
BoitCa aci
-Tension Su ortsA B C& D The condenser anchorage is shown schernaticaliy in Figure 4-5. The condenser has six plate supports with (2 or 3) 2" or 2Tik "diameter anchor bolts each. The supports are designed to resist vertical operating loads.
Thermal growth of the condenser occurs from the fixed point near the center of the base. The sliding plate supports have slotted holes allowing thermal growth radial to a fixed center support pad.
Each of the six perimeter bolted anchorages (Supports A, B, C & D), located at around the condenser perimeter, consists of either (2 or 3) 2-or 2'-inch bolts. Each anchor bolt has greater than 5'ominal length with approximately 48" embedment in the turbine building fcundation per Reference 6a. The bolts are in slotted holes on 15 Inch centers.
The ACI allowable pullout on a cast-in-place anchor is based on the area of the bolt, the bolt material, the concrete strength, and the embedment lenath.
Bolt materia) is assumed to be A-36 steel and the atlowabie
tensile load is based on the nominal bolt area times an allowable'stress.
The allowable stress is 1.7 times the working stress design allowable given in Part 1 cf the AlSG (Reference 4) and is equal to 34,000 psi.
Some of the bolts are 2", some are 2YZ. The pullout capacity of all the bolts is calculated assuming a 2" nominal. The pullout capacity of the bolts is as follows:
Bolt Diameter, D =2.0inch Bolt Area, A = 3.54 inch 2 Bolt Capacity, C (4) 3.14 x 34ksi = 427 Kips xbhfnpmsivlcntc09.doc Attachmellt ffa, Calc, Ho.
- QN-gpss REy NI/89I N9. j
Eoa iNTQRlVA'noNAL gear~~
CALCULATION$HEET JOB NO. 20/0221.01 JOB TV B
SIV 0 R RES I
N CALCNO. ~COO SUBJECT Ccndense/Seismic ueiueii SHEET NO~I UPDATE~/21/.
C HK~DDATE~9/08/
The ACl code formula for pullout capacity of cast-in-place bolts due to failure of concrete is:
C 4)~f'c n(L+D)L where: L 48 Inches (min assumed)
Q= 0.65 f'c 3,500 psi; concrete compressive strength 0 = 2 inches C
1,159 Kips per bolt minimum (greater for 2~/a" bolts)
The projected shear cone for a bolt with 48 inch of embedment is as follows:
- =~*
4 A cone-7,543 in 2 The Installed anchorage configuration has only 15 inch spacing between the bolts. Using the GIP Tension bolt reducllon formula (Reference 1 ):
A'COne=m r2-Th(r28-rs'sin(8/2))
where:
L = bolt embedment = 48 in.
D = bolt diameter = 2 in.
s' actual bolt spacing = 15 in.
r = (2L+D)12 = 49 In.
B = 2 cos (s'/(2L+D)) = 2.8343 rad.
- A'cone = 4,504 in C'=(A'/Acone) xC C'= 692 Kips > 427 Kips per bolt minimum Bolt tren th controls The perimeter supports are primarily designed for tensile loads but can carry some shear loads once sma! I boit hole gaps are overcome.
xAbfnpmsivkalc09.doc Attachment No.
Calo. Iio.
Sheet No.
cT
~ a.~W 0
gp~
ROE INTERNATIONAL Fsh JOB NO.
E~~OO 5~0 JOB CALC NO.
~C-~
SUBJECT CALCULAT1ONSHEET f4-IV I RR 0
OM Condenser Seismic Evaluation SHEETHO~
BY SPH DATE~~/9s CHK~OS DATE 9~10 EE C n r
nchor rtCa a
hear The center support consists of a built-up H section, embedded 4'nto the turbine base mat and welded to the condenser bottom as shown in Figure 5-1. Allseismic shears are assumed to be resisted by the H section.
F0.4 Fx 4/3 (AlSG earthquake increase) = 19.2 Ksl for A-36 steel A=20x2=40 in'-S A,= 16 x 2 x 2 = 64 in E-W V,~~40x19.2=768Kips N-S (Transverse)
V,~
64 x 19.2 = 1,229 Klps 2-IN (Longitudinal)
The existing condenser anchorage system capacity is substantially greater than the demand pf combined operational and lateral seismic DSE forces.
DON>k ECTeI NRRDR (faA PNI)
Figure S-1 Anchor Design xhbfi>po>siv'tcalc09.doc Attachment
- tiD, Calc. Ho.f Of-IRg +
S999t Eo,~+
ot
~@
EQE INTERNATIONAL CALCULATtONSHEET
. utl aL CALC NO. ~00'UBJECT Condenser Seismic Evaluation SHEET NO.
17 BY~~DATE~~rL8 CHK~Jt DATE~o/oo Table 5-1 BFNP Condenser Wefght & C.G.
Section No.
No. of Weight Each Sec Vons (Ib)
Weight y
(lb)
C.G. (in)
Condenser Shell 1,500,000 1,500,000 190 285,000,000 Water fn Tubes Water fn Hot Well 1
327,946 242,575 327,946 61 242,575 18 26,563,6'f 0 4,366,343 Totafs 2,070,520 C.G.
12.72 (ft) 315,929,952 OperaVng wt from Foster Wheeler Dwg.
No. 0-93-505-3-190 Rev. E (Ref. 6f) 2,070,000 (lbs) xhbfnpmsivthcafc09.doc Attachment Htt, Cafe. N0.2 0 0 - Z4'EV.Q Sheet tto J
.of
ca\\op ereevee EQE INTERNATIONAL CAI CULATIONSHEET SHEET No,ts JOB NO. 200621,01 JOB TVABFN-3 MSIVOUTLIER RESOLtjTIOM DALcNQ, ~coe BUBJEGT CondeneerSelemlcEvalcallon BY SPH DATE CHK JOD DATE 8/31/98 9/08/88 Table 6-2 Condenser Loads
'Vertical Loads
+'
Tsnsion DBE Acc.f~l~c Total NI Total: W~:
2070 kips 2070 kips Arm:
12.72 Lateral Eatthquake Accelerations 0.32 '='.2 x
1.6 Vertical Earthquake Accel.
Soil Arnplification Moment:
kfp-ft DBEAcc.vw 0.20 g
IL 8426 kip-ft kip-ft otaI: I/I/tatst:
2070 kips OBE Shear:
662A kips DBE Moment:
8426 Idp-ft Total Uplift:
414 klps Support l.ccation ate Bolt Longitudinal Anchor Load e1e
- Bolt, Anchor Load Max Bott Transverso
~1
~
Bolt Anchor Load Anchor Load Max Bott Anchor Load Total SRSS Bott Active Load Support (kips}
Active (ktps) Suppor t
Load (kips)
(kips}
Load (ktps)
Active Support Load (kips}
(kips)
Active Support (kips}
l.oad (kips)
(kips)
Load (klps}
X1 Y1 21 X2 Y2 22 F1 662 662 682 682 Totals:
1.00 0
x tbfnpulsIAcalc09.doc 1.00 0
1.00
EOE INTeRN161'rONAL CALCULATIONSHEET JOB NO. ggggP1.O1 JOB F
SIV0 l
R UTI CALC NO. ~C-!Qg SUBJECT Condenser Seismic Evaluation SHEET MD~1 BY SPH DATE~~11 9 CHK~CD DATE ~fEEE Table 5-2 (Cont.)
Support Location OP Operating Load (kips)
Arm (ft)
Longitudinai Active Support Tension (kips)
Arm (ft)
Transverse II 1 ~
Active Support Tension Due to Seismic Overturning Tension (kips) 91 9 Actrve Support Tension (kips)
Tension Due to Vert Vertical A1 8
A2 C1 D
C2 F1
-70
-70
-70
-70
-70
-70 39.34 39.34 39.34 39.34 39.34 39.34 86 43 0
86 43 0
27.67 27.67 27.67 27.67 27.67 27.67 203 203 203 0
0 0
69 69 69 69 69 69 Totals:
~420 414 Support Locatfon Diagram C1
~.-re!2 7
27.67 I
l Transverse 7
19.67 19.67 6
Longitudinal
~Ass r~mggps:
Anchors A, B, C 8 D take no shear. Allseismic shear fs taken by the fixed support F Anchors A, B, C 8 D take tension/compression from rocking and operating loads x:tbfrrpmsiQcaic09.doc
EQE INTERNATIONAL CALCULATiONSHEET JOB NQ. ~00~
JOB TVABFM-3M IV U LIER R SQ TON CALO NO.
~C.O SUBJECT Condenser Seismic Evatn lion SHEETNO~
BY~PDATE~iEJJEE CHK JiiD DATE 9$ WINB Table 5-3 Anchor Oernands Support DBE Loads Shear Tension Per Bolt Load Bolt Stress Shear Tension Bolt Size Holt Area 0 of Bolts Holt Stress FofA46
- steel, Allowable ls 34ksi (kips)
(kips)
(k/boit)
(k/bolt)
Ksi A1 B
A2 C1 D
C2 F1 0
0 0
0 0
0 662 80 60 53 80 60 53 0
27 30 18 27 30 18 2
2 2
2.5 2.5 2.5 3.14 3.14 3.14 4.91 4.91 4.91 17.08 23.67 15.33 2.72 1.14 0
ok ok ok ok ok ok Allbolt tension loads are much less than 427 Kip capacity (minimum 2" Diameter)
Anchor shear load is less than minimum 768 Kip (N-S) capacity x."LbfnpmsiAcalc09.doc Attachment Ho.
Cate. No.
Sheet Ho.~+~
of
Eo E IMERNATtoNAL.
CALCULATIONSHEET JOBNO. /gag~
JOB BFN-MS)VOVTLtERR S L
CAUC No.
c-/icosesUBJEGT Condenser Seismic Esa/uadon SHEET NO 21 BY~PDATE
/~/3 /SQ CHK~JD DATE 9/Oe/99
6.0 CONCLUSION
S Analyses of the BFN Unit 3 condenser and comparisons with database condensers were performed.
The condenser design and anchorage are shown to be similar to those at facilities in the earthquake experience database that have experienced earthquakes ln excess of the BFN design basis DBE. The BFN condenser is bounded by the comparable attributes of the database condensere.
Anchorage calculations performed for the BFN Unit 3 condenser indicate that the Imposed seismic and operational demand of the design basis DBE is less than the anchorage capacity.
xhbfnpmsi Acalc09.doc Attachment No.
T Cs/c. /is..'
'0I NBA'/.+-
Sheet HL~7~.
nf
TVANCALCULATIONCOVERSHEET '
p s ~w t.pg ~ L Seismic Evaluation Report Plant BFN Page 1
Of sea Unit 0
~lo preparing Organization NEICEB Calculation Identifier Koy Nouns (Fcr EDM)
Seismic, Component Qualification, Piping, Pipe Support Each time these calculations are Issued, preparer must ensure that the original (Ro)
RIMSJEDM accession number is Gled in.
CD-NODOSE-990113 Applicable Design Document(s)
BFN.6~.7'l00 BFN.5(LC-2107, BFN-SOW-7306 Rev RO R1 for EDM uso EDM Accession Number RN4 9 9090 8 UNlDSystem(s) 00'I RO R3 R3 Quality Retatcd7 Yes No DCN, EDC, NA HIA safety raiated7 Ifyes, mant Yes No Quality Related yes 0
a Prepared Checked F Caramante These calculations contain unverified assumpbon(a) that must be verified tater7 Yes No 0
~
Design Verified Approved Approval Date Ttiese caicuiagons contain Yes special requirements and/or limitin conditions?
Thoso catculaUons contain Yo a design output 0
R attachment7 Ca!cuiation Classification 0
SAR Affected7 Revision applicablity Yes0 No ~
Endre cate
~
Yes0No0 EnUre cato 0 Selected pgs0 Yes0 No0 Entbe cato 0 Solocted pgs0 Entire cate 0
Selected pgs 0 Microfiche generated Number Ycs No 0
a Statement of Problem The Main Steam piping downstream ofthe outboard MSIV's is desired to be capable ofwith standing an earthquake so that any leakage through the MSiY's from the. Reactor side can be contained and diverted to the main coplensor.
This calculation supports the MSIYleakage tech spec change. at SFN.
Abstract
. ~
~
Change Submittai Seismic Evaluation Report', (200918-R-002); August 1999, By EQE International, Oakland, CA. Additionally, Bounding Calculations (200918-C-002) "Seismic evaluation for the Condensera" and (200918-CZ01) "Seismic verification of the main steam drain piping and supports associated with the MSIYalternate leakage treatment pathway" are contained in this calculation.
H MlcroGm and return calculation to Calculation Libary.
Address: POB-1A-BFN Q
Microfilmand return calculation to:
Q Mlctottimond destroy.
TVA40532 [02-1999)
This Page Added By Revision 0 NEDP-2-1 (02-19-1999]
CD-N000l-990 Il3 Page~~
Attnchtnent~
JOB NO.
2CCB'le JOB BFNMSIVTECHSPEC CHANGE CALC.NO.
Ocee SUBJECT ADOITIONALSEISMICEVALUATIONSFOR
BY
- DATE, CHK DATE g Qd f 1,0 PVRPOSE The purpose of this calculation is to document the results ofthe additional selsrnlc evaluation performed on the BFN condensers.
as part of the seismic adequacy verification ofthe components associated with the MSIVAlternate Leakage Treatment (ALT}pathway.
2.0 SCOPE 6 METHODOLOGY The BFN condensers are the terminai boundary points of the MSIValternate leakage treatment (ALT) pathway, hence, they are necessary to maintain structural integrity following a Design Basis Earthquake (DBE). The condensers are located in the Turbine Building and are not designated as Seismic Class I systems.
As part of the plant specific seismic verification of the non-seismic components using the earthquake experience-based approach as outlined in the BWROG Report (Reference 1), the following reviews are performed to demonstrate that the BFN condensers fali within the bounds of the experience database and/or exhibit adequate seismic capacity:
Review of the condenser design codes and standards, design characteristics and parameters, and support/anchorage configurations.
Verification walkdown to identify potential seismic interaction concerns.
~
Engineering evaluations ofthe condenser and support configurations.
The BFN condensers are evaluated using both seismic experience data from past earthquakes and engineering analysis.
Analytical evaluations of the condenser and support anchorage are performed in accordance with the guidelines in the Generic implementation Procedure (GiP, Reference 5), and the general requirements of the American institute of Steel Construction (AISC, Reference 6), as applicable.
JAbfnpmsiAca!c91802.doc
0
all el CD-N0001-990113 Page~1 Attachment~
JOBNO.
ES!919 JOB BFNNSIVSECHSFECCHANGE cacc.no.
ceca sustecr AoomoNAABE'BNICEvAwASIONBFOB lHE BFN CCNOENSEBS 9
sHEETNO.
J4 CHK DATE
3.0 REFERENCES
"BWROG Report for Increasing MSIVLeakage Rate Umfts and Eflmfnatlon of Leakage Controi Systems", GE NEDC-31868P, Revision 2, September 1993.
Safety Evaluation of GE Topical Report, NEDC-31 858P, Revision 2, "BWROG Report for Increasing MSIVLeakage Rate Umlts and Elirninatfon of Leakage Control Systems", U.S.
Nuclear Regulatory Commission, March 3, 1999.
3.
"Browns Ferry - Unit 2, MSIVSeismic Verification Summary Report", EQE Report No.
200918-R-001, Revision 0, August 1999.
4.
"Browns Ferry-Unit 3, MSIVSeismic Verification Summary Report, EQE Report No.
200621-R-001, Revision 0, September 1998.
5.
Generic Implementation Procedure (GIP) for Seismic Verification of Nuclear Plant Equipment", Rev. 2A, March 1993, Prepared by Winston 8 Strawn, EQE, et al., for the Seismic Qualification UtilityGroup (SQUG).
6.
AISC, "Manual of Steel Construction", Eighth Edition, 1980.
7.
TVACalculation No. CD-N0001-980039, "Main Steam Seismic Ruggedness Verification".
8.
TVACalculation No. CD-N0001-980038, "Main Steam Seismic Ruggedness Evaluation".
\\
9.
ASME, "Boilerand Pressure Vessel Code, Section llf, Division I, Appendices",
1980 Edition.
Jhbfnpmsihcalc91802.doc
JON NO.
800818 CALO.NO.
CINM CD-NOOOI-9901 13 9098~5 Attechmeot JOB BFN MS1VTECH SPEC CHANGE SuBJECT ADDmoNALSEtSMtc EVALVAnQNSFOR
'THE BFN CONDENSERS SHEET N80 BY CATE 9 X CHIC ~~ ONE f gt9 qq 4.0 SEISMIC EVALUATIONS The BFN condensers consist of three single-pass, sinale pressure, radiai flowtype surface condensers.
Each condenser is located beneath each of the three low pressure turbines, and Is structuratfy independent.
Table 1 fists the design data for BFN condensers and for the two experience database sites listed in the SWROG Report (i.e., Moss Landing 6 &7, and Ormond Beach 1 &2). Design characteristic comparisons of the BFN condensers with the above two selected database condensers are presented in details ln Reference 8. These include size (surface area), weight, height, and plan comparisons.
The BFN condenser desfgn data is comparable to the data for these two database sites.
The BFN condenser anchorage was compared with the performance of simifar condenser in the earthquake experience database.
The shear areas ofthe condenser anchorage, in the directions parallel and transverse to the turbine generator axis, divided by the seismic demand, were used to compare with those presented in the BWROG Report (Reference 1), The BFN condenser anchorage shear area to seismic demand is substantially greater than the selected database sites. The condenser support anchorage was also evaluated and the results indicate that the combined seismic DBE and operational demand is less than the anchorage capacity based on the AISC allowables.
Maximum stress rat!os are 0.70 for bolt tension ln the perimeter support feet, and 0.86 for shear in the center support built-up section.
Detaned description ofthe BFN condenser support anchorage and anchorage evaluations are presented in Reference 8.
A composite comparison ofthe ground response spectra of selected earthquake experience database sites with the conservatively estimated BFN DBE ground spectrum (i.e., 0.2g Housner input spectrum at rock outcrop scaled by 1.8 to account for soll amplification) is shown in Figure 1.
In general, the earthquake experience database sites have experienced strong ground motions that are in excess of the BFN DBE at the frequency range of interest (i.e., about 1 Hz.
and above), with the exception ofthe Ormond Beach site. Many of the database site ground motions envelope the conservatively estimated BFN DBE ground spectrum by large factors in various frequency bands withir. the 1 Hz. and above range.
Figures 2 and 3 show the individual comparison plots ofthe conservatively estimated BFN DBE ground spectrum with the Moss Landing and Ormond Beach site spectra, respectively.
J:)bfnpmsivhcafc91802.doc
CD-N0001-990113 PE e, BC AMchment F
JOB hlO.
SN9t8 JOB BW MSIVTECH SPEC CHANGE CALO.NO.
CCSS SUBJECT ADOIRONALSEISMICEVALUATIONSFOR TNEBFN CONDENSERS SHEETNO. b BY DATE T N+
CHK DATE The Orrnond Beach power Plant was affected by the magnitude 6.8, Point Mugu Earthquake in 1973, which was considered to be a relatively moderate earthquake, and was substantially lower than the 5089 Lama Prieta Earthquake (Magnitude 7.1) as experienced in the Moss Landing Power Plant as well as those experienced by most of the other database sites.
To ensure that adequate seismic margins exist in the BFN condensers in the event of a plant DBE, additional seismic evaluation was performed to verifythe overall structural integrity ofthe condensers, as shown In pages 7 to 9 of this calculation.
Results ofthe evaluation indicate that the condenser shell stresses due to the seismic DBE loads are small. Maximum stress ratios, based on AISC allowables, are 0.1 2 for combined axial and bending and 0.1 0 for shear.
J:5bfnpmsiv11calc91802.doc
EQE N7EANAllONAL CD N0001-990113 Page~@
Attachment~
JOB NO. ~Era Et JOB CALC. NV. C 00~
8U8JECT FJ H
1 ~
SHEET NO.
BY~<~
DATE~<<- t CHH'D~ DATE M~S~7
'C WOEWSerc. C~s(
GQE'../S.
Cane $ J'rt<p c" VJ 4 $ 5c 5 in t/tr
'an&nSE;c a.Jar I/ <~ +
get'see c
L~~as(P@E ),
Pvorn FE5EAW l
0& 5 lo /~Peat 9GE. ~'
SPet"Iruert
/jtOyj~et+ L OCCLICN A. f7carE C
O.3 Z.g Y<E OWL a.c.M~r~+Jcat
= ch'+g C /3 > AsEt rE 0 A,r sh$
IDIO.<ig EBW ove-r~>l cosoksoe-t ts.1X
( Pet. 9 )
I r
J
>r,As~r eV
~ll dmcms,~s
-" go p( 3Z )(
qq.~ (g'}
8/EC/<
+n.c ~t C.SS = 7/g e/~/~
M<4n~L.
=
AsvH Ares'c
( rq ~go z~:3 Pe48)
E tr 0 7/
0
EQ" N:PNA'ttOMAL TaataaATTEAAT JOB NO. 2uW8 ma CALC. NO. Q OOW SUBJECT CD-N0001-990113 Page~E Athtchmeat~
C5aa x
SHEET NO. IP BY 7 'ATE~T<-O'I CNK'D DATE ~P MADEv589 s'Haze C4o~'d3 Sec.Vien Vgh
~a%
~ = (7/p ) (<0 ( 5 0+ 32 ) (I'.) = /72z vt 0 x (7/g x ahxaa) ( a <<<<'1
= 'f.'7a xia
>2 a
= Z.+3 xla id Qg,aaEL)/+
"' iZx(TY0,3a'.)(~)
E.aoxhi az z
3 STY=~
3 ~
7 xa (sax~>/z AKTt=L~ ~t'.EE Zn 5+iCSSS.S 207d (I+ET T)
) +Q I 7c.z p~Uc 30 Ea ~'Sv gl J.. VZ Egin)
- 2. +53(.io+
D ao"~.x=5 (>-.vw'~i~')
3 ~ d l 3(xaa 0
~ 4f QSB
'lC Sa AISC P<<TED3~'V~KC ~f. 4 )
EQE NH~ICkVAl CD-N0001-990113 Page ~LL Attachment~
~ itlLatIlCHAL JOB NQ. ~~94 JOB CALC NC,~C-CB BJBJECT r e(l c i SHEET NO.
BY
+~
DAK~I~Wl CHK'D DATE~DIE T cat4E44W ~+~<
Cc ~~ <)
P
)n P
IB P~ 4t.
w 5Eh10
)L. bib'L4 66Z +
H.
RJ a
BVI
0 ~w CD-N0001-990113 Page~/D Attachment~
JOB NO.
200010 JOB SFN IJSIVTECH SPEC CHANGE CALC.NO.
D002 SUBIECT ACDmONALSEISMICEVALUATIONSFOR THE BEN COND ENSE RS swor No. />
eY 0201TR c
CHK DATE 8 3b 1$
Table 1 Comparison of Browns Ferry and Selected Database Condensers Design Attributes Moss Landing Units 6 a7 armand Beach Units1 82 Browns Condenser Manufacturer ingersoll-Ra'nd Southwestern Foster Wheeler Fiow Type Single Pass Single Pass Single Pass Condenser Dimensions (LxWxH) 65 ft. x36 ft.
x 47 ft.
52ft. x 27 ft.
x20 ft 58 ft. X 32 ft.
x 47ft.
Condenser Surface Area 435,000 sq. ft.
21 0,000 sq. ft.
222,000 sq. ft, Condenser Shell Material Condenser Shelf Thickness Cu Bearing ASTM A-285C Cu Bearing ASTM A-285C 3/4 ASTM A-285C 7/8" Condenser Operating Weight 3,115 klps 1,767 kips 2,076 kips Tube Material Ai-Brass 80-1 0 Cu-Ni AI-6XN Tube Size 1
dla.
1" dia.
7/8" dia.
Tube Length 65 ft, 53 ft.
50 ft, l Tube Wall Thickness 20 BWG 22 BWG J'SfnPmstvgalc91802 doc
~ p,~
IFJTERTJATTOPJ A1 CD-N0001-990113 Ps c
8))
Attachment JOBND
. SCCSIS JOB BFHMSIVTECHSPECCNANDE CALO. NO, CCCS SUSJECT ADDITIOHALSEISMIC EVALUATIONSFOR TNESFHCONDEVSERB snMTNo. //
BY DATE ~F/
'f1'HK OATE ~QJ~
Table 1 (cont.)
Comparison of Browns Ferry and Selected Database Condensere Design Attributes Moss Landing Units 6 &?
Ormond Beach Units1 &2 Brogans Ferry Number of Tubes 25,590 15,220 19.480 Tube Sheet Material Muntz Muntz ASTM A-285C Tube Sheet Thickness 1-1/2" 1-1/4" No. of Tube Support Plates Tube Support Plate Material Not Given Cu Bearing ASTM A-285C ASTM A-285C Tube Support Plate Thickness 3/4" 5/8" 7/8" Tube Support Plate Spacing 48 in.
38 in.
39 in, Water Box Material 2% Ni Cast tron ASTM A-48 Class 30 Cu Bearing ASTM A-285C ASTM A-285C Expansion Joint Rubber Belt Stainless Steel Rubber Belt Hotwell Capacity 20,000 gal.
34,338 gal.
28,000 gal. (max.)
J:11bfnpmsiAcaic91B02.doc
0
CD-N0001-990113 Page ~JF" Attachment 8
JOB NO.
000'IS JOB BFNMSIVTECN FEC CHANCE CALC. NO.
COSE SUBJECT AOOITIONALSEISMIC EVALUATIONSFOR THE BFN CCNCENSERS SHEET ND. (~
BY OATS F 0 CHK DATE Bd f Ffgure 1 Comparison of Browns Ferry DBE Ground Spectrum with Selected Database Site Spectra 1.4 1
g g
0.8 D 0.6
~ 01 ~ ~
~ ~
~
~
~ 0
~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~
~ ~ 0 ~ ~ ~
~ ~ 0 ~ ~ 0
~o, 0
~
EB R'
~
E 0
~
~
0 F
I
~ ~ ~ ~
~
~
SrrTTEns Ferry BLI5XOBQ
~~ Ye!icy Srearn - USGS
Brarbank-USGS 8 Cenrro ~ EJVIl 10 Moss Leneeng lar&&E grumboare lrsy. Ave.
~ COOITYerer Long.
~ Js-Commcroo 01/6
~ Grersrele
~ N200E
~ Cansrrsl Beech - H270K
~10ALCO Ave.
0,4 4 ~
0
~
~ ~
~
~ ~
~
~
~
~
~
~
~ ~
~
~ ~
0 ~ 0 0MMOSC~ ~ ~ ~ ~ ~ ~ ~ ~
~
0.1 Froqrrerrcy, Hz.
10 100 J:~bfllPm&iV'rc&lc91802.dOC
4
IJ, PEGDJ ASST'D-N0001-990113 Page~!9 Affachmeaf~
JOB ND 2009'I B JOB BPN MSIVTECH SPEC CHANGE CALO. NO, Deca SVRIECT ADDITIDNALSBSMICEVALIIATIONSPDR TIIEBPN CONDENSEDS SXEEV NO. /9 GY DATE 2 2V CHK DATE Figure 2 Comparison of Browne FerrJJ DBE and Moss Landing Power Plant Ground Spectra BFN (1.6XDBE)
PG8 E Estffnete
BFN (OBE3 H 1.6 EE 98 09
~ 1N P
0.8
~A~
~ ~
~ ~
~
~
~ ~
~ ~
~N ~
~ Q
~ ~ 0
~ 00 ~
Na
~
~ ~
~ ~ 0
~
~
~~
~ 0 ~
Frequency (Eh)
JAbfnpmsivhcalc91802.doc
NFTEEFJATTCSJht CD-NOOO 1-990113 Page ~t/
AttSChmCDt~
JOB NO.
2009IS JOS SFMMSIVlECKSFECCHANGE CALO. NO.
CO02 SUBJECT ADDITIONALSEISL!IC EVALUATIONSFOR THE SFN CONDENSER S SHEET No.
/tA-BY DATE ~gzs~p CHX DAlE ~g'30 Figure 3
(-omparlson of Brovtns Ferry DBE and Orn1ond Beach Power Plant Ground Spectra 2.4 2C 1.B C0 P
8 1>
V7 0.8 BFH (1.6XDBE) 2 -- Horiz. N1SOE
--- - - -Moriz. N270E
BFN (DBE) 0.4 0.1 r 000 r FF 0
Wrr r
~
~
Frequency (Hz) 10 100 3'.+fnprnsiv~h91802.doc
= e
IRTTEESTATTCSTAT CD-NOOOI-990113 Page ~i~~
Athtchmcnt~
JOB NO.
200838 JOB BFN MSNTECH SPEC CHANGE CALC. NO, C422 SUILIECT AOOITIONALSEISMIC SYALUATIONSFOR 1HEBFNCONOENSEAS SHEETNO. t+
BY CATE ~22M /R CHK,~~ OATE ~y~ygg 5.0 CONC LUSJONS The comparisons ofthe condenser seismic experience data, supplemented by the additional condenser evaluation and the anchorage capacity evaluations demonstrate that the conclusions presented in the BWRQG Report (Reference l) can be applied to the BFN condensers.
That is, a significant failure of the condenser in the event of a DBE at BFN is highly unlikely and contrary to the large body of historical earthquake experience data.
Jhbfnpmsivhcalc91892.doc
ENCLOSURE 3
TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)
UNITS 2 and 3
PROPOSED TECHNICAL SPECIFICATIONS (TS)
CHANGE TS-399 INCREASED MAIN STEAM ISOLATION VALVE (MSIV) LEAKAGE
RESPONSE
TO REQUEST FOR ADDITIONAL INFORMATION (RAI)
DATED NOVEMBER 23 ~
1 999
RESPONSE
TO NRC STAFF QUESTIONS ON DOSE METHODOLGY Item 1
TVA's letter of September 28,
- 1999, contains the statement that the change request is based on the utilization of the Boiling Water Reactor Owners'roup (BWROG) methodology described in NEDC-31858P, Revision 2, BWROG Report for Increasing MSIV Leakage Rate Limits and Elimination of Leakage Control Systems.
Page E-10 of the submittal contains a similar statement.
- However, the staff has reason to question these statements:
a
~
b.
Page 3 of the TVA submittal contains the statement "This analysis uses the holdup and plateout factors described in NEDC-31858P...."
Was the methodology of NEDC-31858P Revision 2 used as stated on page 1 of the letter and on page E-10 of the submittal, or were selected parameters used as stated on page 3?
The analysis summary provided by TVA indicates that TVA has ratioed the previous MSIV leakage results obtained for a leakage of 11.5 scfh to 100 scfh, and that TVA also ratioed these results for the increase in power rating (1.05x),
and the 1.02x instrument penalty.
The staff compared the values for 11.5 scfh tabulated in the summary with the values provided by GE to TVA in 1992 (ND-Q2031-920075R1) and observed identical results.
The results provided in 1992 were based on the NEDC-31858P revision 1.
GE made several changes in the analysis methodology for revision 2.
Revision 1
was never approved by the NRC as a topical report.
Please confirm that the TVA analysis was performed using the methodology of NEDC-31858 Revision 2, including all incorporated assumptions, parameters, and methods.
If the revision 2 methodology as described in the approved BWROG
A topical report was not used, please correct the submittal and provide sufficient information for the NRC staff to make a finding of the acceptability of the methodology TVA did use.
TVA Res onse to Item 1
There were no changes in dose methodology between NEDC-31858P Revision 1 and Revision 2.
Note that the NEDC Appendix C
(Dose Methodology) is dated September 1991 in both Revision 1 and 2.
Therefore, the reference in the September 28, 1999, submittal to Revision 2 of NEDC-31858 is appropriate and the March 3,
- 1999, NRC Safety Evaluation Report (SER) is likewise applicable.
NEDC-31858 was subsequently issued in final form as NEDC-31858P-A in August 1999.
As noted in the response to RAI Item 2 below, TVA has reperformed the MSIV leakage dose calculations.
These were performed in accordance with the NEDC-31858 methodology as reviewed in the NRC SER.
NRC Item 2 TVA's ratioing of the dose results obtained with a leak rate of 11.5 scfh to reflect the proposed higher leak rate appears to assume that the BWROG deposition model is linear with regard to flow rate.
Table 8-3 of NEDC-31858P Revision 2 indicates that the doses increased by approximately a
factor of three when the leakage was increased a factor of two.
Please justify the assumption of linear proportionality.
TVA Res onse to Item 2 After further review, we agree that using a linear extrapolation to scale the MSIV leakage contribution to dose is not conservative.
Therefore, TVA has performed the specific MSIV dose calculations rather than using extrapolation factors for the MSIV leakage dose contribution.
These were completed in accordance with the NEDC methodology as reviewed in the NRC SER.
This new analysis resulted in a reduction of the requested MSIV allowable leakage rate.
Therefore, TVA is providing an amended change request as part of this response as contained in Enclosures 4 and 5.
E3-2
'4
NRC Item 3 The information TVA provided is not clear with regard to the leakage rate actually assumed in the radiological calculations.
Page E1-14 of the submittal indicates that the radiological calculations were based on a total net MSIV leakage of 400 scfh.
The analysis input tabulation for the TVA analysis indicates that the MSIV leakage is "...200 scfh/valve (400 scfh maximum which equates to 100 scfh/valve average),
this translates to a time dependent total flow for 4 valves of:..<list>."
The specified list shows time dependent leakage rates ranging from 172.54 cfh to 24.73 cfh.
Please confirm that the calculations assumed a total of 400 scfh for the entire release period.
TVA Res onse The base calculation for MSIV leakage contribution to dose was provided by GE to TVA in 1992 (Calculation ND-Q2031-920075 R1).
For the accuracy of the remaining dose calculations, an MSIV source removal term is needed to properly account for the remainder of the net mass release from the primary containment.
In other words, the increased MSIV leakage reduces the source concentration of radioisotopes in primary containment that are modeled in other leakage pathways.
The time dependent flow listing referred to in the RAI is this MSIV leakage source reduction term as converted from standard cubic feet per hour (scfh) to cubic feet per hour (cfh).
The conversion factor is based on time dependent temperature and pressures.
As noted in Item 2, TVA has reperformed the dose calculations as discussed in Enclosure 4.
NRC Item 4 Page C-28 of NEDC-31858P Revision 2, discusses the fraction of MSIV leakage that will flow to the HP turbine.
Based on the proposed alternative leakage path and assuming loss of offsite power and single failure, what is the fraction of MSIV leakage to the HP turbine is assumed in your analyses.
If the fraction is greater than 0.01 (page C-30 of NEDC-31858P Revision 2), please confirm that doses from this release path were addressed in the TVA analyses.
E3-3
TVA Res onse to Item 4
The BFN alternate leakage treatment (ALT) flow path is shown in Figure 3-1 of Attachment 4 of the September 28,
- 1999, TS-399 submittal.
The ALT path is from the outboard side of the MSIVs through Flow Control Valve (FCV)-1-58 to the condenser and satisfies the sizing requirements of NEDC-31858P-A paragraph 6.1.1(2) which states that the ALT flow path should, based on the radiological dose methodology, be at least 1 square inch internal cross sectional area.
FCV-1-58 has Emergency Diesel Generator power available
- and, hence, does not rely on the availability of offsii'e power.
The orificed bypass path around FCV-1-58 shown in Figure 3-1 addresses Section 5.2 of the NRC safety evaluation dated March 3,
- 1999, which states that a secondary path to the condenser, having an orifice, should exist.
This secondary path is considered a contingency alignment in the event of the unlikely failure of FCV-1-58 and is not sized to meet the 1-inch path provision discussed, in the NEDC specified for the credited ALT path.
- Moreover, NEDC-31858 does not prescribe that a secondary ALT which path is fully redundant to the credited ALT path in terms of sizing be available in the event of a single failure.
Therefore, the MSIV increased leakage dose calculations assume the primary ALT path is available which meets the 0.01 ratio criteria referenced on page C-30 of NEDC-31858P-A and the fraction of leakage going to High Pressure (HP) turbine is specifically accounted for in the MSIV dose calculations.
In this situation, the dose contribution from this HP turbine pathway is very small (truncated to zero in the 1992 (ND-Q2031-920075R1))
calculations.
As discussed in Item 2 above, TVA has reperformed the dose calculations for the MSIV leakage dose contribution.
These were completed in accordance with the NEDC methodology which includes the HP turbine path.
NRC Item 5 TVA has revised the X/Q values for its control room in its analyses of the increased MSIV leakage.
For the top of stack release, the X/Q values are on the order of 1.0E-16 seconds(sec)/m'or the Unit 1 intake and 1.0E-10 sec/m'or the Unit 3 intake.
The staff understands that these X/Q values were derived using the ARCON96 methodology.
The discussion on page 30 of the documentation for ARCON96 (NUREG/CR-6331 Rev
- 1) addresses the case of elevated stacks E3-4
The top of the stack dose contribution to the control room is small in comparison to the dose resulting from increased MSIV leakage, and the use of these Regulatory Guide based X/Q's rather than the ARCON96 X/Qs has a minor impact on net doses.
TVA continues to use ARCON96 methodology to estimate ground level releases.
The resulting net allowable MSIV leakage rate is provided in the amended TS change request in Enclosures 4 and 5.
NRC Item 6 J
The release rate from the condenser to the environment as modeled in NEDC-31858P Revision 2 methodology apparently assumes the mechanical vacuum pump (MVP) will be tripped.
Due to a previous modification at BFNP, the MVP no longer trips on a MSLRM signal.
When TVA analyzed the consequences of removing the automatic functions initiated by the
The 1850 cfm flow rate of the MVP is significantly greater than the 400 scfh flow rate of the MSIV leakage, suggesting that holdup in the condenser may be limited.
Please describe if and how this impact is considered in your analyses.
TVA Res onse to Item 6
The Main Steam Line Radiation Monitor (MSLRM) MVP trip on high radiation has not been removed.
Refer to BFN Final Safety Analysis Report (FSAR) Section 7.12.1.3 for a discussion of this function.
The MVPs pumps are only used during startup at very low reactor powers to establish an initial vacuum prior to placing the steam jet air ejectors (SJAEs) in service.
The SJAEs,are the preferred method of maintaining condenser vacuum since they provide dilution steam for control of combustible offgas products and SJAE flow is treated by the normal offgas system.
Operating Instructions require the MVPs not be used above 5% reactor power and the MVPs will auto-trip and isolate on increasing condenser vacuum (when operating vacuum is established by the SJAEs).
The MVPs have no auto-start capability.
Based on the above, the assumption that the MVPs are not in service is appropriate for the MSIV dose calculations for TS-399 since these calculations are based on accidents occurring at full reactor power.
E3-6
NRC Item 7 The analysis input tabulation provided includes a value for turbine building free volume and turbine building exhaust rate.
These parameters appear to imply that credit is being taken for holdup in the turbine building.
Page C-70 of NEDC-31858P Revision 2 states that no credit has been taken for holdup in the turbine building.
In its letter dated June 12,
- 1998, TVA stated in response to question 56 that holdup in the turbine building was not credited in assessing the consequences of MSIV leakage.
Please confirm that the analyses do not credit holdup in the turbine building.
TVA Res onse to Item 7 The'SIV leakage dose analyses do not credit fission product holdup by the turbine building.
The turbine building volumes listed in the analysis inputs tabulation are results from a previous calculation revision and, as noted, do not enter into the dose calculations.
NRC Item 8 The analysis input tabulation provides a flow rate of 24,750 for the SGTS with all three trains running.
- However, previous submittals to the NRC and 514.6.3.6 of the August 1999 FSAR indicate the SGTS flow to be 22,000 cfm.
Please explain the difference in flow rates.
TVA Res onse to Item 8 BFN has tested the SGTS flow to be between 22,000 cfm and 22, 500 cfm.
The flow used in the dose analysis is 22, 500 plus a typical 10% variance for a total of 24750 cfm.
This is a conservative assumption since the higher SGTS flow will reduce building and SGTS train hold-up times which would provide for additional decay of fission products.
This assumption is also conservative since more isotopes are calculated to be released earlier in the accident sequence when the X/Q values are less favorable.
NRC Item 9
The information TVA provided indicates that TVA considered a
potential release via the hardened wetwell vent.
TVA' analysis assumes 10 cfh with a decay period of eight hours, which does not appear to be consistent.
If the 10 cfm is expected damper leakage (i.e.,
bypass leakage),
why is the pathway not considered earlier in the event.
Intentional flow initiated at eight hours would likely have a higher
flow rate given the size of the pathway (14").
Such intentional flow is beyond the design basis.
Please provide an explanation of these assumptions.
TVA Res onse to Item 9 The Hardened Wetwell Vent (HWWV) release path is modeled as a potential leakage path in the dose analysis.
The calculation models leakage from the primary containment into HWWV only, not the intentional operation of the HWWV which is reserved for the mitigation of beyond design basis events.
The 10 cfh value is based on the Primary Containment Leak Rate Test Program leak rate criteria for the HWWV valves as listed in Table 5.2-2 of the
- FSAR, The HWWV line is over 500 feet long and is a 14-inch pipe.
Assuming slug flow in the 14-inch line (13.25 inches inner diameter) at 10 cfh yields a travel time to exit the HWWV piping of:
T (hours)
T (hours)
T (hours)
T (hours)
T (hours) pipe volume/leakage flow rate (pipe length x pipe cross section) /leakage flow rate (500 feet x x x radius') / leakage flow rate (500 feet x 3.14 x ((13.25 inches/
(2 x 12 inches/foot))') / 10 feet'/hour 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> In the dose calculation, 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> is used as the commencement point for HWWV contribution, which is clearly conservative.
E3-8
ENCLOSURE 4
TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)
UNITS 2 and 3
PROPOSED TECHNICAL SPECIFICATIONS (TS)
CHANGE TS-399 INCREASED MAIN STEAM ISOLATION VALVE (MSIV) LEAKAGE
RESPONSE
TO REQUEST FOR ADDITIONAL INFORMATION (RAI)
DATED NOVEMBER 23, 1999 PROPOSED TECHNICAL SPECIFICATION (TS)
CHANGE TS-399 DESCRIPTION AND EVALUATION OF PROPOSED CHANGE DESCRIPTION OF PROPOSED TS CHANGE In a letter dated September 28,
- 1999, TVA requested a change to the Units 2 and 3
TS Surveillance Requirement (SR) 3.6.1.3.10 to increase the allowed MSIV leakage from 11.5 standard cubic feet per hour (scfh) per valve to 200 scfh for individual MSIVs with a 400 scfh combined maximum pathway leakage for all four MSIV lines.
In i he November 23,
- 1999, request for additional information, NRC questioned the use of extrapolation factors to calculate the dose associated with an increased MSIV leakage criteria.
After further review TVA agreed that using a linear extrapolation to scale the MSIV leakage did not provide conservative dose resulting from the increased MSIV leakage.
Subsequently, TVA performed calculations to determine the MSIV leakrate dose concentration.
The recalculation resulted in a combined maximum pathway leakage of 168 scfh.
Therefore, TVA has revised the requested change to increase the allowed MSIV leakrate from 11.5 scfh per valve to 100 scfh for individual MSIVs with a 150 scfh combined maximum pathway leakage for all four MSIV lines.
The TS Bases are likewise being revised to match the proposed change.
A marked-up copy showing the exact TS and Bases changes is provided in Enclosure 5.
REASON FOR THE PROPOSED CHANGE As discussed in the September 28,
- 1999, TS change
- request, refurbishment of a MSIV to meet the 11.5 scfh criteria is a man-hour intensive effort which accumulates approximately 4.5 man-rem during a complete rebuild.
With a 100 scfh limit for individual MSIVs and 150 scfh combined maximum
- pathway, no Unit 2 MSIVC would have required rework in the last four operating cycles, and only one valve during the two most recent Unit 3 operating cycles.
The change would lower personnel radiation exposure and improve the performance integrity of the MSIVs by reducing the number of maintenance activities associated with restoring the leakage to an overly strict lower limit.
Approval of this proposed TS change would also be an economic benefit to TVA in terms of direct costs and a
reduction in outage activities.
SAFETY ANALYSIS Radiolo ical Dose Assessment In the November 23,
- 1999, Request For Additional Information, the staff questioned the appropriateness of extrapolating the results of the earlier dose calculations to determine the affects of a larger MSIV leakage criteria.
After subsequent
- review, TVA determined that using a linear extrapolation to scale the MSIV leakage did not provide conservative dose results from the increased MSIV leakage.
To address the issue, TVA reperformed the dose calculations to determine the MSIV leakage dose rather than use a linear extrapolation method.
The contribution from MSIV leakage was calculated using the methodology described by NEDC-31858P Revision 2, Increasing MSIV Leakage Rate Limits and Elimination of Leakage Control Systems.
For this TS change, the offsite dose calculations and control room dose calculations have been revised using a
total net MSIV leakage of 168 scfh for all four MSIV lines.
The table below provides the dose in man-rem from 168 scfh MSIV leakage.
E4-2
ghpgbg@4!@~lb%><
- '~g'a~g~~jxQ':.j~~ii
- 29. 48
- 0. 6827 0.1576 5.840 0.1665 0.1006 85.97 0.4815 0.4839 The revised calculation show that the dose contribution from the increased MSIV leakrate is far below the 10 CFR 20.1201(a)(1)(ii) limits.
- Also, 10 CFR 100 and GDC-19 dose limits are maintained.
For conservatism, TVA has chosen to reduce the allowable total MSIV leakage to 150 scfh in the amended TS change request.
E4-3
-.e 0
ENCLOSURE 5
TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)
UNITS 2 and 3
PROPOSED TECHNICAL SPECIFICATIONS (TS)
CHANGE TS-399 INCREASED MAIN STEAM ISOLATION VALVE (MSIV) LEAKAGE
RESPONSE
TO REQUEST FOR ADDITIONAL INFORMATION (RAI)
DATED NOVEMBER 23, 1999 MARKED-UP TS PAGES I.
AFFECTED PAGE LIST Unit 2 3.3-16 B 3.6-35 Unit 3 3.6-16 B 3.6-35 II.
MARKED-UP PAGE S SEE ATTACHED
" "0
SURVEILLANCEREQUIREMENTS continued SURVEILLANCE PCIVs 3.6.1.3 FREQUENCY SR 3.6.1.3.5 Verifythe isolation time of each power operated, automatic PCIV, except for MSIVs, is within limits.
In accordance with the Inservice Testing Program SR 3.6.1.3.6 Verifythe isolation time of each MSIV is z 3 seconds and (5 seconds.
In accordance with the Inservice Testing Program SR 3.6.1.3.7 Verify each automatic PCIV actuates to the isolation position on an actual or simulated isolation signal.
24 months SR 3.6.1.3.8 Verify each reactor instrumentation line EFCV actuates to the isolation position on a simulated instrument line break signal.
24 months SR 3.6.1.3.9 Remove and test the explosive squib from each shear isolation valve of the TIP System.
24 months on a STAGGERED TEST BASIS SR 3.6.1.3.10 100 scfh and that the combined maximum pathway leakage rate for all four main steam lines is
( 150 scfh Verify leakage rate through each MSIVis s 44-.S-sech-when tested at a 25 psig.
In accordance with the Primary Containment Leakage Rate Testing Program SR 3.6.1.3.11 Verify combined leakage through water tested lines that penetrate primary containment are within the limits specified in the Primary Containment Leakage Rate Testing Program.
In accordance with the Primary Containment Leakage Rate Testing Program BFN UNIT2 3.6-16
PCIVs B3.6.1.3
'BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.6.1.3.9 The TIP shear isolation valves are actuated by explosive charges.
An in place functional test is not possible with this design.
The explosive squib is removed and tested to provide assurance that the valves will actuate when required.
The replacement charge for the explosive squib shall be from the same manufactured batch as the one fired or from another batch that has been certified by having one of the batch successfully fired. The Frequency of 24 months on a STAGGERED TEST BASIS is considered adequate given the administrative controls on replacement charges and the frequent checks of circuit continuity (SR 3.6.1.3.4).
SR 3.6.1.3.10 100 The combined maximum pathway leakage rate for all four main steam lines must be
< 150 scfh when tested at
> 25 psig. If the leakage rate through an individual MSIV exceeds 100 scfh, the leakage rate shall be restored below the alarm limit value as specified in the Containment Leakage Rate Testing Program referenced in TS 5.5.12.
The analyses in References 1 and 5 are based on leakage that is less than the s ecified leakage rate.
Leakage through each MSIVmust be ~~ cfh when tested at a P< (25 psig). This ensures that MSIV leakage is properly accounted for in determining the overall primary containment leakage rate. The Frequency is specified in the Primary Containment Leakage Rate Testing Program.
SR 3.6.1.3.11 Surveillance of water tested lines ensures that sufficient inventory will be available to provide a sealing function for at least 30 days at a pressure of 1.1 Pa.
Sufficient inventory ensures there is no path for leakage of primary containment atmosphere to the environment following a DBA. Leakage from containment isolation valves that terminate below the suppression pool water level may be excluded from the total leakage provided a sufficient fluid inventory is available as described in 10 CFR 50, Appendix J, Option B.
BFN UNIT2 B 3.6-35
~ SURVEILLANCEREQUIREMENTS continued SURVEILLANCE PCIVs B3.6.1.3 FREQUENCY SR 3.6.1.3.5 Verifythe isolation time of each power operated, automatic PCIV, except for MSIVs, is within limits.
In accordance with the Inservice Testing Program SR 3.6.1.3.6 Verifythe isolation time of each MSIV is a 3 seconds and s 5 seconds.
In accordance with the Inservice Testing Program SR 3.6.1.3.7 Verify each automatic PCIV actuates to the isolation position on an actual or simulated isolation signal.
24 months SR 3.6.1.3.8 Verifyeach reactor instrumentation line EFCV actuates to the isolation position on a simulated instrument line break signal.
24 months SR 3.6.1.3.9 Remove and test the explosive squib from 24 months on a each shear isolation valve of the TIP System.
STAGGERED TEST BASIS SR 3.6.1.3.10 Verify leakage rate through each MSIV is s 4-1-.S-sefh-when tested at a 25 psig.
100 scfh and that the combined maximum pathway leakage rate for all four main steam lines is
( 150 scfh In accordance with the Primary Containment Leakage Rate Testing Program SR 3.6.1.3.11 Verifycombined leakage through water tested lines that penetrate primary containment are within the limits specified in the Primary Containment Leakage Rate Testing Program.
In accordance with the Primary Containment Leakage Rate Testing Program BFN UNIT 3 3.6-16
PCIVs B3.6.1.3 P
BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.6.1.3.9 The TIP shear isolation valves are actuated by explosive charges.
An in place functional test is not possible with this design.
The explosive squib is removed and tested to provide assurance that the valves willactuate when required.
The replacement charge for the explosive squib shall be from the same manufactured batch as the one fired or from another batch that has been certified by having one of the batch successfully fired. The Frequency of 24 months on a STAGGERED TEST BASIS is considered adequate given the administrative controls on replacement charges and the frequent checks of circuit continuity (SR 3.6.1.3.4).
SR 3.6.1.3.10 100 The combined maximum pathway leakage rate for all four main steam lines must be
< 150 scfh when tested at 25 psig. If the leakage rate through an individual MSIV exceeds 100 scfh, the leakage rate shall be restored below the alarm limit value as specified in the Containment Leakage Rate Testing Program referenced in TS 5.5.12.
The analyses in References 1 and 5 are based on leakage that is less than the s ecified leakage rate.
Leakage through each MSIVmust be s 4-'t4 cfh when tested at a Pi (25 psig).
his ensures that MSIV leakage is properly accounted for in determining the overall primary containment leakage rate. The Frequency is specified in the Primary Containment Leakage Rate Testing Program.
SR 3.6.1.3.11 Surveillance of water tested lines ensures that sufficient inventory willbe available to provide a sealing function for at least 30 days at a pressure of 1.1 Pa.
Sufficient inventory ensures there is no path for leakage of primary containment atmosphere to the environment following a DBA. Leakage from containment isolation valves that terminate below the suppression pool water level may be excluded from the total leakage provided a sufficient fluid inventory is available as described in 10 CFR 50, Appendix J, Option B.
BFN UNIT 3 B 3.6-35
ENCLOSURE 6
TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)
UNITS 2 and 3
PROPOSED TECHNICAL SPECIFICATIONS (TS)
CHANGE TS-399 INCREASED MAIN STEAM ISOLATION VALVE (MSIV) LEAKAGE
RESPONSE
TO REQUEST FOR ADDITIONAL INFORMATION (RAI)
DATED NOVEMBER 23, 1999 COMMITMENT LISTING Section XI surveillance testing will consist of disassembly and inspection on a rotating basis (one check valve each refueling outage) in accordance with Position 2 of GL 89-04.
The piping and components within the boundaries of the MSIV ALT path are considered to be within the scope of the BFN Section XI IST and ISI programs,
- and, accordingly, will be inspected and tested in accordance with the IST/ISI programs.
Additional detail is provided (in the response) for certain aspects of the program pertaining to the RAI questions.