ML070890318
| ML070890318 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 05/17/2007 |
| From: | Peter Bamford NRC/NRR/ADRO/DORL/LPLI-2 |
| To: | Crane C AmerGen Energy Co |
| Miller G, NRR/DORL, 415-2481 | |
| References | |
| TAC MD1807 | |
| Download: ML070890318 (10) | |
Text
May 17, 2007 Mr. Christopher M. Crane President and Chief Executive Officer AmerGen Energy Company, LLC 4300 Winfield Road Warrenville, IL 60555
SUBJECT:
THREE MILE ISLAND NUCLEAR STATION, UNIT NO. 1 - REQUEST FOR ADDITIONAL INFORMATION REGARDING PROPOSED STEAM GENERATOR TUBE INTEGRITY TECHNICAL SPECIFICATION CHANGES (TAC NO. MD1807)
Dear Mr. Crane:
By letter dated May 15, 2006 (Agencywide Documents Access and Management System (ADAMS) Accession Number ML061420294), AmerGen Energy Company, LLC, submitted an amendment request for the Three Mile Island Nuclear Station, Unit No. 1. The proposed amendment would revise the steam generator tube integrity technical specifications to be consistent with the Nuclear Regulatory Commission (NRC)-approved Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler, TSTF-449, Steam Generator Tube Integrity, Revision 4 (ADAMS Accession Number ML051090200).
The NRC staff has been reviewing the submittal and has determined that additional information is needed to complete its review. The specific questions are found in the enclosed request for additional information. The questions were sent via facsimile on February 20, 2007, to Mr. David Distel of your staff. The draft questions were sent to ensure that the questions were understandable, the regulatory basis for the questions was clear, and to determine if the information was previously docketed. These questions were discussed with your staff in a teleconference on March 23, 2007, with a follow-up teleconference being held on May 7, 2007.
Based on these teleconferences and internal staff review, a more concise set of questions is being transmitted herein. Several of the original questions have either been removed or characterized as observations and others have been clarified. However, the principal issues requiring further information remain unchanged from the February 20, 2007 transmittal of the draft questions. A response date of 30 days from the date of this letter was originally agreed upon in the March 23, 2007 teleconference. However, since the staff feels you have had ample time to review and begin to formulate responses to these questions and because, on balance, the burden of response has been reduced, a response within 14 days of the date of this letter is requested.
Please note that if you do not respond to this letter within the prescribed response time or provide an acceptable alternate date, in writing, we may reject your application for amendment
C. Crane under the provisions of Title 10 of the Code of Federal Regulations, Section 2.108. If you have any questions, I can be reached at (301) 415-2833.
Sincerely,
/ra/
Peter Bamford, Project Manager Plant Licensing Branch I-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-289
Enclosure:
Request for Additional Information cc w/encl: See next page
C. Crane under the provisions of Title 10 of the Code of Federal Regulations, Section 2.108. If you have any questions, I can be reached at (301) 415-2833.
Sincerely,
/ra/
Peter Bamford, Project Manager Plant Licensing Branch I-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-289
Enclosure:
Request for Additional Information cc w/encl: See next page DISTRIBUTION:
PUBLIC RidsNrrDORLLpl1-2 RidsOgcRp L. Miller, NRR LPL1-2 R/F RidsNrrLACSola RidsAcrsAcnwMailCenter K. Karwoski, NRR RidsNrrPMGMiller RidsNrrDorlDpr RidsRgn1MailCenter RidsNrrDciCgsb RidsNrrPMPBamford Accession Number: ML070890318
- via email OFFICE LPL1-2/PM LPL1-2/PM LPL1-2/LA CGSB/BC LPL1-2/BC NAME PBamford GEMiller*
RSola Ahiser (LM for)
HChernoff DATE 5/14/07 05/14/07 5/14/07 5/15/07 w/ edits 5/17/07 Official Record Copy
Three Mile Island Nuclear Station, Unit 1 cc:
Site Vice President - Three Mile Island Nuclear Station, Unit 1 AmerGen Energy Company, LLC P. O. Box 480 Middletown, PA 17057 Senior Vice President - Nuclear Services AmerGen Energy Company, LLC 4300 Winfield Road Warrenville, IL 60555 Vice President - Operations, Mid-Atlantic AmerGen Energy Company, LLC 200 Exelon Way, KSA 3-N Kennett Square, PA 19348 Vice President - Licensing and Regulatory Affairs AmerGen Energy Company, LLC 4300 Winfield Road Warrenville, IL 60555 Regional Administrator Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Chairman Board of County Commissioners of Dauphin County Dauphin County Courthouse Harrisburg, PA 17120 Chairman Board of Supervisors of Londonderry Township R.D. #1, Geyers Church Road Middletown, PA 17057 Senior Resident Inspector (TMI-1)
U.S. Nuclear Regulatory Commission P.O. Box 219 Middletown, PA 17057 Director - Licensing and Regulatory Affairs AmerGen Energy Company, LLC 200 Exelon Way, KSA 3-E Kennett Square, PA 19348 Director Bureau of Radiation Protection Pennsylvania Department of Environmental Protection Rachel Carson State Office Building P.O. Box 8469 Harrisburg, PA 17105-8469 Plant Manager - Three Mile Island Nuclear Station, Unit 1 AmerGen Energy Company, LLC P. O. Box 480 Middletown, PA 17057 Regulatory Assurance Manager - Three Mile Island Nuclear Station, Unit 1 AmerGen Energy Company, LLC P.O. Box 480 Middletown, PA 17057 Ronald Bellamy, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Michael A. Schoppman Framatome ANP Suite 705 1911 North Ft. Myer Drive Rosslyn, VA 22209 Vice President, General Counsel and Secretary AmerGen Energy Company, LLC 2301 Market Street, S23-1 Philadelphia, PA 19101
Three Mile Island Nuclear Station, Unit 1 cc:
Dr. Judith Johnsrud National Energy Committee Sierra Club 433 Orlando Avenue State College, PA 16803 Eric Epstein TMI Alert 4100 Hillsdale Road Harrisburg, PA 17112 Correspondence Control Desk AmerGen Energy Company, LLC P.O. Box 160 Kennett Square, PA 19348 Manager Licensing - Three Mile Island Nuclear Station, Unit 1 Exelon Generation Company, LLC 200 Exelon Way, KSA 3-E Kennett Square, PA 19348 Assistant General Counsel AmerGen Energy Company, LLC 200 Exelon Way Kennett Square, PA 19348
Enclosure REQUEST FOR ADDITIONAL INFORMATION (RAI)
THREE MILE ISLAND UNIT 1 STEAM GENERATOR TUBE INTEGRITY TECHNICAL SPECIFICATION AMENDMENT TAC NO. MD1807 DOCKET NO. 50-289 By letter dated May 15, 2006 (ML061420294 [Agencywide Document Access and Management System accession number]), AmerGen Energy Company, LLC (the licensee) submitted a license amendment request (LAR) regarding Three Mile Island Unit 1 (TMI-1) steam generator (SG) tube integrity technical specifications (TS). The proposed amendment would revise the SG tube integrity TSs to be consistent with the U.S. Nuclear Regulatory Commission (NRC)-approved Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler, TSTF-449, Steam Generator Tube Integrity, Revision 4 (ML051090200).
The licensee provided additional information regarding their SG tube integrity TSs in letters dated October 6, 2006 (ML062830331) and December 12, 2006 (ML063480459). Based on the review of the information provided by the licensee, the staff has the following additional questions.
1.
The cover page of the October 6, 2006, document indicates that the SG tube integrity TSs are going to be made consistent with Revision 1 of TSTF-449 and Page 1 of of the same document indicates that changes to the SG tube integrity TSs will be consistent with Revision 3 of TSTF-449. Please clarify that the modifications being made are consistent with TSTF-449, Revision 4.
2.
The Actions and Note in TS Section 3.1.1.2 should be rearranged to indicate that TS Section 3.1.1.2.a.(3) should only be entered if TS Section 3.1.1.2.a.(2) was not met. As a result, discuss your plans to remove the last sentence of the Note and relocate it as part of TS Section 3.1.1.2.a.(3). This sentence should also be modified to remove reference to TS Section 3.1.1.2.a.(1) since entry into TS Section 3.1.1.2.a.(3) is not permitted if TS Section 3.1.1.2.a.(1) is not satisfied. For example, If the requirements of Section 3.1.1.2.a.(2) are not met for one or more tubes then perform the following.
3.
It is the staffs understanding from discussions during a conference call on May 7, 2007, that you assume that the primary-to-secondary leakage during a steam line break (SLB) is more limiting than for the other design basis accidents (DBAs) that assume primary-to-secondary leakage exists. As a result of this assumption, you only determine the amount of leakage during a SLB and compare this to the assumptions in your accident analyses. Please provide the technical basis for why it is only necessary to assess the leakage during a SLB (i.e., demonstrate that by satisfying the leakage limit associated with a SLB you will meet the leakage limit for the other DBAs).
4.
In TS Table 4.1.2, the Test for primary to secondary leakage is listed as Evaluate.
The meaning of this term is not clear. Isnt the Test for primary to secondary leakage, continuous monitoring of the effluent (steam and feedwater systems) for radioactive isotopes or performing radiochemical analyses of grab samples of the steam and feedwater systems? Similarly, isnt the Test for reactor coolant system leakage a water inventory balance rather than Evaluate? Please discuss your plans to modify your proposal to more accurately reflect the Test for monitoring primary-to-secondary leakage.
In addition, discuss your plans to modify the Note in Table 4.1-2 to make it more consistent with TSTF-449. The TSTF wording is Note: Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishment of steady state operation.
5.
Even though TMI-1 will not have approved repair methods, please discuss your plans to describe the existing kinetic expansions and sleeves where the following note appears in the TSTF-449, [Steam generator tube repairs are only performed using approved repair methods as described in the Steam Generator Program.]
6.
Please discuss your plans to modify TS Section 6.9.6.f to be consistent with TSTF-449 by reading, Total number and percentage of tubes plugged or repaired to date.
7.
Given that TMI-1 has tubes that are sleeved, please discuss your plans to modify TS Section 6.19.c to clearly indicate the repair criteria for the non-sleeved region of the sleeved tubes and the repair criteria for the sleeved region of the tube. In addition, discuss your plans to clearly indicate that the alternative to the 40-percent depth based criteria can only be applied to non-sleeved tubes.
8.
Please discuss your plans to include the definition of inside diameter intergranular attack from TS Section 4.19.4.a.9 in TS Section 6.19.c.1 (current proposed numbering).
In addition, please discuss your plans to remove repaired or from the fourth sentence of this TS since there will be no approved repair methods at TMI-1.
9.
The reason for referencing the accident-induced leakage criteria in proposed TS Section 6.19.c.1 and 6.19.c.2 is not apparent. If it is consistent with your design and licensing basis it would seem that it is addressed by the first sentence under TS Section 6.19.b.2.
In addition, please discuss your plans to clearly indicate that the only exception to the 1 gpm per SG limit is for leakage attributed to the kinetic expansions. For example, Leakage from all sources excluding the leakage attributed to the degradation described in TS Section 6.19.c.2 is also not to exceed 1 gpm per SG.
10.
Please discuss your plans to modify TS Section 6.19.d.4 to make it more consistent with your current TSs. For example, AmerGen Engineering Report, ECR No. TM 01-00328 during all subsequent SG inspections.
11.
Please discuss your plans to modify TS Section 6.19.f to indicate that repairs were performed by kinetic expansion and sleeving but no new tube repairs can be made without prior NRC approval.
12.
In your proposed TS Section 6.19.d, you define the length of the tube as from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet.
Based on this definition, the parent tube behind the upper sleeve joint is part of the pressure boundary and is required to be inspected. Please confirm that this inspection will be performed with the implementation of TSTF-449. As currently proposed the 40-percent repair criteria would apply to the parent tube behind the upper sleeve joint since proposed TS Section 6.19.c.2 does not address the repair criteria for the parent tube behind the sleeve upper joint. However, a more appropriate repair criteria for flaws in the parent tubing at the upper sleeve joint may be 20-percent through-wall since this appears to be consistent with the original sleeve qualification (please note that this is based on the licensees submittal and not a review of the original qualification report).
As a result, please discuss your plans to clarify the acceptance criteria for any flaws in the parent tube behind (adjacent to) the sleeve's upper joint.
In addition, it is not clear why the phrase "in accordance with ECR No. TM 02-01121, Rev. 2" is needed in the specification. Please clarify the purpose for the last phrase in TS Section 6.19.c.2. If ECR No. TM 02-01121, Revision 2 does not clarify the repair criteria for sleeves, discuss your plans to delete it.
- 13.
In your proposed TS (and TSTF-449) a SG tube is defined as the entire length of the tube, including the tube wall [and any repairs made to it], between the tube-to-tubesheet weld at the tube inlet and the tube-to-tubesheet weld at the tube outlet. Given this definition, the proposed repair criteria in TS Section 6.19.c could be misinterpreted.
Please discuss your plans to modify your TS to more clearly define the repair criteria for the sleeved portion of a tube. For example, this TS may be modified using the following; 1.
The non-sleeved region of a tube found by inservice inspection to contain flaws with a depth equal to or exceeding 40-percent of the nominal tube wall thickness shall be plugged or repaired except when alternate tube repair criteria permitted by technical specifications are satisfied.
2.
Tubes shall be plugged if the sleeved region of a tube is found by in-service inspection to contain (actual repair criteria is dependent upon resolution of questions 7 and 12).
3.
The following alternate repair criteria may be applied as an alternative to the 40-percent depth based criteria in TS 6.19.c.1:
a.
Volumetric Inside Diameter (ID) Inter-Granular Attack (IGA) indications...
b.
Upper tubesheet kinetic expansion indications...
- 14.
Given that TS Section 6.19.d provides the inspection requirements for the tube (which by definition includes the sleeve) and the inspections must be performed to ensure SG tube integrity, it is not clear that the proposed requirements in TS Section 6.19.d.6 are needed. As a result, please discuss your plans for removing these proposed requirements. If the proposed requirements are maintained, please discuss your plans to add TS Section 6.19.d.6 to the first paragraph of TS Section 6.19.d.
- 15.
As currently written, proposed TS Section 6.19.f is unclear. Please discuss your plans to modify this proposed requirement. For example, "There are currently no approved repair methods; however, tubes repaired with sleeves and by kinetically expanding the tube in the upper tubesheet prior to 2006 may remain in service subject to the requirements of TS Sections 3.1.1.2, 4.19, and 6.19.
- 16.
As currently proposed, any flaws in the parent tube between the sleeve lower end and the parent tubes kinetic expansion are required to be plugged on detection. Since several flaws were removed from service as a result of the sleeving campaigns (several imperfections and one inside diameter indication greater than 40-percent through-wall),
it would appear that upon adoption of the new proposed TSs there will be several tubes that exceed the repair criteria. This will require a tube integrity assessment, possibly a forced shutdown, and will result in these tubes being plugged. Please clarify that this was your intent and provide the results of the tube integrity assessment.
- 17.
Proposed TS Section 6.9.6.i implies that repairs are authorized at TMI-1. Given that tubes have been repaired in the past but that no new repairs will be permitted, please discuss your plans to modify TS Section 6.9.6.i to indicate that you will provide the number of tubes remaining in service using repair methods previously implemented.
The questions listed above represent information the NRC staff needs to make a regulatory decision on the merits of the proposed LAR. The TS Bases are a licensee controlled document. Changes to the Bases are evaluated in accordance with the TMI-1 TS Bases Control Program described in TS 6.18. This program is subject to inspection by NRC staff.
The following observations on the TS Bases sections that were submitted along with the LAR are provided for your consideration. Modification to these Bases is strongly suggested to ensure that TSs are not inadvertently modified by a discussion in the Bases.
A.
The reference to TS Section 6.19.c.2 in the Bases for TS Section 3.1.6 is not clear, and the sentence (containing this reference) does not appear complete (when compared to TSTF-449). With respect to the corresponding sentence in TSTF-449, it indicates that primary-to-secondary leakage has the potential to increase as a result of accident induced conditions. This point is removed from your Bases markup (with the exception of the leakage attributed to a SLB).
B.
It was stated on Page 4-78 of Enclosure 2 (Bases for TS Section 4.19) that In these analyses, the steam discharge to the atmosphere is based on the total primary to secondary leakage from all SGs of 1 gallon per minute or is assumed to increase to the leakage rates described in TS 6.19.c.2 as a result of accident-induced conditions. With respect to the corresponding sentence in TSTF-449, it indicates that primary-to-secondary leakage has the potential to increase as a result of accident-induced conditions. This point is removed from your Bases markup (with the exception of the leakage attributed to a SLB).
C.
The following Bases sections should be revised so as to be consistent with the wording in the proposed TS sections:
i.
The last sentence in the first paragraph of the Bases for TS Section 3.1.6 is not clear since one could interpret it to mean that the leakage of 144 gallons per day (gpd) is permitted from each SG. For example, The TS requirement to limit the sum of the primary to secondary leakage from both SGs to less than or equal to 144 gallons per day is significantly less than the conditions assumed in the safety analysis. The first sentence of the second paragraph of the same Bases should be modified for a similar reason. For example, The limit on the sum of the primary to secondary leakage from both SGs of 144 gallons per day bounds the TSTF-449, Rev. 4 limit of 150 gallons per day per SG, which is based on the operational leakage performance criterion in Nuclear Energy Institute 97-06, Steam Generator Program Guidelines (Ref.1).
ii.
The first sentence of the first paragraph of the Bases for TS Section 4.1 is not clear since it could be interpreted to mean the leakage of 144 gpd is permitted from each SG. For example, The primary to secondary leakage surveillance in TS Table 4.1-2, Item 12, verifies that the sum of the primary to secondary leakage from both SGs is less than or equal to 144 gallons per day. The last sentence of the first paragraph of the same Bases should be similarily clarified. For example, The operational leakage rate limit applies to the sum of the leakage through both SGs.
iii.
In the first full paragraph on Page 4-80, you should clarify that the leakage limit is the sum of the leakage from both SGs.
D.
In the Bases for TS Section 3.4, you indicated that the following sentence will be removed, According to Specification 3.1.1.2a, both [once-through steam generators]
shall be operable whenever the reactor coolant average temperature is above 250 degrees F. For example, the sentence should be modified to read, According to Specification 3.1.1.2.a, both SGs shall have tube integrity whenever the reactor coolant average temperature is above 200 degrees F.
E.
If Table 4.1-2 is modified per Question 4 by removing reference to the initial leak rate, a corresponding change should be made to the second paragraph of the insert to TS Page 4-2b (Bases for TS Section 4.1).
F.
The note currently on Page 4-83a of your existing TSs should be added to your associated proposed TS Bases.
G.
TSTF-449 indicates, The SG heat removal function is addressed by..., specific TS sections in the Bases for the SG tube integrity section. A similar sentence at the end of the first paragraph in your Bases for TS Section 4.19 should be added.
H.
On Page 4-78 of Enclosure 2, you state, a SG tube is defined as the entire length of the tube, including the tube wall and any repairs made to it,.... A discussion regarding tubes that have been sleeved and that additional sleeves will not be installed without prior NRC approval should be added to provide clarification relating to the current SG status.
I.
Since serviceability is no longer defined in your TSs, you should modify the second paragraph under Surveillance Requirement 4.19.2 in your Bases to remove reference to it.