ML11105A217
| ML11105A217 | |
| Person / Time | |
|---|---|
| Site: | Watts Bar  |
| Issue date: | 04/07/2011 |
| From: | - No Known Affiliation |
| To: | Division of Operating Reactor Licensing |
| References | |
| Download: ML11105A217 (70) | |
Text
WBN2Public Resource From: Boyd, Desiree L [dlboyd@tva.gov]
Sent: Thursday, April 07, 2011 7:10 AM To: Epperson, Dan; Poole, Justin; Raghavan, Rags; Milano, Patrick; Campbell, Stephen Cc: Crouch, William D; Hamill, Carol L; Boyd, Desiree L
Subject:
TVA letter to NRC_04-06-11_SSER 22 OPEN ITEMS Attachments: 04-06-11_SSER 22 OPEN ITEMS_Final.pdf Please see attached TVA letter that was sent to the NRC today.
Thank You,
~*~*~*~*~*~*~*~*~*~*~*~*~*~*~
Désireé L. Boyd WBN 2 Licensing Support Sun Technical Services dlboyd@tva.gov 4233658764
~*~*~*~*~*~*~*~*~*~*~*~*~*~*~
1
Hearing Identifier: Watts_Bar_2_Operating_LA_Public Email Number: 338 Mail Envelope Properties (7AB41F650F76BD44B5BCAB7C0CCABFAF1D758C84)
Subject:
TVA letter to NRC_04-06-11_SSER 22 OPEN ITEMS Sent Date: 4/7/2011 7:10:01 AM Received Date: 4/7/2011 7:10:56 AM From: Boyd, Desiree L Created By: dlboyd@tva.gov Recipients:
"Crouch, William D" <wdcrouch@tva.gov>
Tracking Status: None "Hamill, Carol L" <clhamill@tva.gov>
Tracking Status: None "Boyd, Desiree L" <dlboyd@tva.gov>
Tracking Status: None "Epperson, Dan" <Dan.Epperson@nrc.gov>
Tracking Status: None "Poole, Justin" <Justin.Poole@nrc.gov>
Tracking Status: None "Raghavan, Rags" <Rags.Raghavan@nrc.gov>
Tracking Status: None "Milano, Patrick" <Patrick.Milano@nrc.gov>
Tracking Status: None "Campbell, Stephen" <Stephen.Campbell@nrc.gov>
Tracking Status: None Post Office: TVANUCXVS2.main.tva.gov Files Size Date & Time MESSAGE 336 4/7/2011 7:10:56 AM 04-06-11_SSER 22 OPEN ITEMS_Final.pdf 3589844 Options Priority: Standard Return Notification: No Reply Requested: No Sensitivity: Normal Expiration Date:
Recipients Received:
U.S. Nuclear Regulatory Commission Page 2 April 6, 2011
Enclosures:
- 1. Response to Action Items From Appendix HH of NUREG-0847, Supplement 22
- 2. Action Items From Appendix HH of NUREG-0847, Supplement 22 to Be Answered Later
- 3. Action Items From Appendix HH of NUREG-0847, Supplement 22 For NRC Inspection/Review
- 4. List of Regulatory Commitments Attachments:
- 1. Response to TVA for Core Loading Information and End of Cycle Assembly Burnup Data for Watts Bar Unit 2 Cycle 1
- 2. A Review of Electronic Components in a Radiation Environment of 5x104 RADS
U. S. Nuclear Regulatory Commission Region II Marquis One Tower 245 Peachtree Center Ave., NE Suite 1200 Atlanta, Georgia 30303-1257 NRC Resident Inspector Unit 2 Watts Bar Nuclear Plant 1260 Nuclear Plant Road Spring City, Tennessee 37381
U.S. Nuclear Regulatory Commission Page 3 April 6, 2011 bcc (Enclosures):
Stephen Campbell U.S. Nuclear Regulatory Commission MS 08H4A One White Flint North 11555 Rockville Pike Rockville, Maryland 20852-2738 Charles Casto, Deputy Regional Administrator for Construction U. S. Nuclear Regulatory Commission Region II Marquis One Tower 245 Peachtree Center Ave., NE Suite 1200 Atlanta, Georgia 30303-1257
ENCLOSURE 1 Response to Action Items From Appendix HH of NUREG-0847, Supplement 22 Tennessee Valley Authority - Watts Bar Nuclear Plant - Unit 2, Docket No. 50-391 The item numbers used below correspond to item numbers in SSER 22, Appendix HH.
- 3. Confirm TVA submitted update to FSAR section 8.3.1.4.1. (NRC safety evaluation dated August 31, 2009, ADAMS Accession No. ML092151155)
Response: In Amendment 95 of the Unit 2 FSAR, the third sentence of the third paragraph of 8.3.1.4.1 reads "Any conduit exceeding 40% cable fill will be evaluated and justified by engineering."
Amendment 95 of the Unit 2 FSAR was submitted to the NRC via TVA letter dated November 24, 2009, Watts Bar Nuclear Plant (WBN) - Unit 2 - Final Safety Analysis Report (FSAR), Amendment 95, ADAMS Accession number ML093370275).
- 5. Verify timely submittal of pre-startup core map and perform technical review. (TVA letter dated September 7, 2007, ADAMS Accession No. ML072570676)
Response: Attachment 1 provides the requested core map.
- 6. Verify implementation of TSTF-449. (TVA letter dated September 7, 2007, ADAMS Accession No. ML072570676)
Response: Amendment 65 to the Unit 1 TS revised the existing steam generator tube surveillance program and was modeled after TSTF-449, Rev. 4. The NRC approved Amendment 65 via letter dated November 3, 2006, Watts Bar Nuclear Plant, Unit 1 - Issuance of Amendment Regarding Steam Generator Tube Integrity (TS-05-10) (TAC No. MC9271). Revision 82 made the associated changes to the Unit 1 TS Bases.
Developmental Revision A to the Unit 2 TS and TS Bases made the equivalent changes to the Unit 2 TS / TS Bases. Affected TS sections include the following: LEAKAGE definition in 1.1, LCO 3.4.13 (RCS Operational LEAKAGE), LCO 3.4.17 (SG Tube Integrity), 5.7.2.12 (Steam Generator (SG)
Program), and 5.9.9 (Steam Generator Tube Inspection Report).
Developmental Revision A of the Unit 2 TS was submitted to the NRC via letter dated March 4, 2009, Watts Bar Nuclear Plant (WBN) Unit 2 -
Operating License Application Update, (ADAMS Accession number ML090700378).
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ENCLOSURE 1 Response to Action Items From Appendix HH of NUREG-0847, Supplement 22 Tennessee Valley Authority - Watts Bar Nuclear Plant - Unit 2, Docket No. 50-391
- 8. Verify rod control system operability during power ascension. TVA should provide a pre-startup map to the NRC staff indicating the rodded fuel assemblies and a projected end of cycle burnup of each rodded assembly for the initial fuel cycle 6-months prior to fuel load.
(NRC safety evaluation dated May 3, 2010, ADAMS Accession No. ML101200035)
Response: Attachment 1 provides the requested pre-startup map indicating the rodded fuel assemblies and the projected end of cycle burnup of each rodded assembly for the initial fuel cycle.
- 14. TVA stated that the Unit 2 PTLR is included in the Unit 2 System Description for the Reactor Coolant System (WBN2-68-4001), which will be revised to reflect required revisions to the PTLR by September 17, 2010. (Section 5.3.1)
Response: Revision 1 (effective August 12, 2010) to the Unit 2 System Description for the Reactor Coolant System (WBN2-68-4001) was revised to reflect the required revisions to the Pressure and Test Limits Report (PTLR).
- 15. TVA should confirm to the NRC staff the completion of Primary Stress Corrosion Cracking (PWSCC) mitigation activities on the Alloy 600 dissimilar metal butt welds (DMBWs) in the primary loop piping. (Section 3.6.3)
Response: Unit 2 has completed the Mechanical Stress Improvement Process (MSIP).
Amendment 103 to the Unit 2 FSAR added five new paragraphs to the end of Section 5.5.3.3.1 (Material Corrosion/Erosion Evaluation) to describe this process.
Amendment 103 was submitted via TVA to NRC letter dated March 15, 2011, Watts Bar Nuclear Plant (WBN) - Unit 2 - Final Safety Analysis Report (FSAR), Amendment 103.
- 18. Based on the extensive layup period of equipment within WBN Unit 2, the NRC staff must review, prior to fuel load, the assumptions used by TVA to re-establish a baseline for the qualified life of equipment. The purpose of the staffs review is to ensure that TVA has addressed the effects of environmental conditions on equipment during the layup period.
(Section 3.11.2.2)
Response: This item was addressed in the response to RAI 3.11 - EQ - 1. in TVA to NRC letter dated December 17, 2010, Watts Bar Nuclear Plant (WBN) Unit 2 -
Safety Evaluation Report Supplement 22 (SSER22) - Response to Requests for Additional Information (ADAMS Accession No. ML103540560).
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ENCLOSURE 1 Response to Action Items From Appendix HH of NUREG-0847, Supplement 22 Tennessee Valley Authority - Watts Bar Nuclear Plant - Unit 2, Docket No. 50-391
- 20. Resolve whether or not routine maintenance activities should result in increasing the EQ of the 6.9 kV motors to Category I status in accordance with 10 CFR 50.49.
(Section 3.11.2.2.1).
Response: The refurbishment of the 6.9 kV motors for Unit 2 involved routine maintenance activities. These maintenance activities did not modify or repair the motor insulation system originally supplied by Westinghouse. However, review of the original qualification report indicates that the testing performed meets the requirements for a Category I qualification. Motors which only require routine maintenance will have their binders revised and will be re-classified as Category I.
In one case (Containment Spray Pump Motor), the maintenance activities determined the need to rewind the motor. The rewound motor insulation system is qualified in accordance with the EPRI motor rewind program which meets Category I criteria.
- 22. TVA must clarify its use of the term equivalent (e.g., identical, similar) regarding the replacement terminal blocks to the NRC staff. If the blocks are similar, then a similarity analysis should be completed and presented to the NRC for review. (Section 3.11.2.2.1)
Response: This item was addressed in the response to RAI 3.11 - EQ - 3.b. in TVA to NRC letter dated December 17, 2010, Watts Bar Nuclear Plant (WBN) Unit 2
- Safety Evaluation Report Supplement 22 (SSER22) - Response to Requests for Additional Information (ADAMS Accession No. ML103540560).
The response stated, For EQ applications, the replacement terminal blocks will be new GE CR151B terminal blocks certified to test reports that document qualification to NUREG-0588, Category I criteria.
TVA discussed this issue with the NRC during the ACRS meeting on February 24, 2011. The NRC staff accepted TVAs explanation of the term equivalent as provided above. Therefore, TVA considers this item to be closed.
- 23. Resolve whether or not TVAs reasoning for not upgrading the MSIV solenoid valves to Category I is a sound reason to the contrary, as specified in 10 CFR 50.49(l).
(Section 3.11.2.2.1)
Response: TVA will qualify the MSIV solenoids to the Category I criteria.
- 24. The NRC staff requires supporting documentation from TVA to justify its establishment of a mild environment threshold for total integrated dose of less than 1x103 rads for electronic components such as semiconductors or electronic components containing organic material.
(Section 3.11.2.2.1)
Response: Calculation A Review of Electronic Components in a Radiation Environment of 5x104 RADS is provided as Attachment 2.
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ENCLOSURE 1 Response to Action Items From Appendix HH of NUREG-0847, Supplement 22 Tennessee Valley Authority - Watts Bar Nuclear Plant - Unit 2, Docket No. 50-391
- 26. For the scenario with an accident in one unit and concurrent shutdown of the second unit without offsite power, TVA stated that Unit 2 pre-operational testing will validate the diesel response to sequencing of loads on the Unit 2 emergency diesel generators (EDGs). The NRC staff will evaluate the status of this issue and will update the status of the EDG load response in a future SSER. (Section 8.1)
Response: There are four diesel generators (DGs) which supply onsite power to both Units 1 and 2 at Watts Bar Nuclear Plant. Each DG is dedicated to supply power to shutdown boards as follows:
x DG 1A-A feeds power into Unit 1, 6.9 kV shutdown board 1A-A x DG 2A-A feeds power into Unit 2, 6.9 kV shutdown board 2A-A x DG 1B-B feeds power into Unit 1, 6.9 kV shutdown board 1B-B x DG 2B-B feeds power into Unit 2, 6.9 kV shutdown board 2B-B Redundant trains of ESF loads for each unit are powered from each shutdown board. If offsite power is lost (LOOP), one train in each unit is capable of powering the loads required to mitigate the consequences of an accident or safely shut down the unit.
The following loading tables provide the blackout loading plus the common accident loads (load rejection, with an accident on the opposite unit and a loss of offsite power) for the safe shutdown of the non-accident unit. As discussed previously, these loadings are bounded by the accident loading.
Maximum Steady-State Running Load, 0 hrs to 2 hrs*
Short- Minimum Time Margin 1A-A 1B-B 2A-A 2B-B Rating (%)
Kw 3,540.71 3,492.81 3,593.87 3,702.44 4,840 23.5 Time (sec) 1,810 1,810 1,810 1,810 KVA 3,952.32 4,182.61 3,979.04 4,123.56 6,050 30.8 Time (sec) 1,810 1,810 1,810 1,810 E1-4
ENCLOSURE 1 Response to Action Items From Appendix HH of NUREG-0847, Supplement 22 Tennessee Valley Authority - Watts Bar Nuclear Plant - Unit 2, Docket No. 50-391 Maximum Steady-State Running Load, 2 hrs to End**
Minimum Continuous Margin 1A-A 1B-B 2A-A 2B-B Rating (%)
Kw 3,540.71 3,492.81 3,593.87 3,702.44 4,400 15.8 Time (sec) 7,200 7,200 7,200 7,200 KVA 3,952.32 4,182.61 3,979.04 4,123.56 5,500 23.9 Time (sec) 7,200 7,200 7,200 7,200 Maximum Starting + Running (Transient) Loading, 0 to 180 sec Cold Minimum Engine Margin 1A-A 1B-B 2A-A 2B-B Capability (%)
Kw 3,508.32 3,320.81 3,396.14 3,806.81 4,785 20.4 Time (sec) 90 90 90 90 Maximum Starting + Running (Transient) Loading, 180 sec to End Hot Minimum Engine Margin 1A-A 1B-B 2A-A 2B-B Capability (%)
Kw 3,806.30 3,948.00 3,994.52 3,997.23 5,073 21.2 Time (sec) 360 360 360 360 Maximum Step Load Increase (Excitation), 0 sec to End Generator Minimum Step Load Margin 1A-A 1B-B 2A-A 2B-B Capability (%)
Kw 3,645.01 3,728.75 3,722.30 3,725.01 8,000 53.4 Time (sec) 20 20 20 20
- Automatic load sequencing only
- Includes operator actions (load additions)
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ENCLOSURE 1 Response to Action Items From Appendix HH of NUREG-0847, Supplement 22 Tennessee Valley Authority - Watts Bar Nuclear Plant - Unit 2, Docket No. 50-391
- 27. TVA should provide a summary of margin studies based on scenarios described in Section 8.1 for CSSTs A, B, C, and D. (Section 8.2.2)
Response: TVA to NRC letter dated December 6, 2010, Watts Bar Nuclear Plant (WBN)
Unit 2 - Safety Evaluation Report Supplement 22 (SSER22) - Response to Requests for Additional Information, (ADAMS accession number ML103420569) included the response to RAI 8.2.2 - 1. That response stated, The loading for a dual unit trip (item a) is slightly less than the loading with one unit in accident and a spurious accident signal in the other unit.
Therefore, a separate load flow was not performed.
A separate load flow was performed for a dual unit shutdown resulting from an abnormal operational occurrence with and without offsite power. The resulting loading on CSSTs is provided in the following table:
STEADY STATE LOADING RATING MW MVAR MVA MVA CSST C - X 10.75 4.68 11.72 24/32/40 CSST C - Y 11.02 4.96 12.08 24/32/40 CSST C - P 21.80 10.69 24.28 33/44/55 (The above loading on CSST C is with both ESF trains of both units powered from this transformer; CSST D is out of service)
CSST D - X 10.75 4.69 11.73 24/32/40 CSST D - Y 11.02 4.96 12.08 24/32/40 CSST D - P 21.80 10.70 24.28 33/44/55 (The above loading on CSST D is with both ESF trains of both units powered from this transformer; CSST C is out of service)
CSST A - X 21.86 9.28 23.75 36/48/60*
CSST A - Y 29.89 17.72 34.75 36/48/60*
CSST A - P 52.04 33.35 61.81 57/76/95*
(The above loading on CSST A is with one ESF train of each unit transferred to this transformer. CSST D is out of service; CSSTs C, A, and B are available.)
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ENCLOSURE 1 Response to Action Items From Appendix HH of NUREG-0847, Supplement 22 Tennessee Valley Authority - Watts Bar Nuclear Plant - Unit 2, Docket No. 50-391 CSST B - X 21.86 9.28 23.75 36/48/60*
CSST B - Y 28.14 16.66 32.70 36/48/60*
CSST B - P 50.29 31.82 59.51 57/76/95*
(The above loading on CSST B is with one ESF train of each unit transferred to this transformer. CSST C is out of service; CSSTs D, A, and B are available.)
The second FA rating for CSSTs A and B is FUTURE.
The worst case margin for CSSTs C and D is 70% (X, Y winding) and 55% for primary winding. The worst case margin for CSSTs A and B is 27% (X, Y winding) and 18% for primary winding.
This additional analysis will be included in the next revision of AC Auxiliary Power System Analysis Calculation EDQ00099920070002.
- 28. TVA should provide to the NRC staff a detailed discussion showing that the load tap changer is able to maintain the 6.9 kV bus voltage control band given the normal and post-contingency transmission operating voltage band, bounding voltage drop on the grid, and plant conditions. (Section 8.2.2)
Response: For CSSTs C and D, the load tap changer (LTC) is set to regulate 6.9kV shutdown board voltage at 7,071V (102.5%). For CSSTs A and B, the LTC is set to regulate the voltage at the 6.9kV start buses (which can power the 6.9kV shutdown boards through the 6.9kV unit boards) at 7,071V (102.5%).
The upper and lower setpoints of the dead bands are 7,132V (103.4%) and 7,010V (101.6%), respectively. The dead band considered is +/-82.2V equivalent to the operating tolerances identified for these setpoints. The LTCs have the following parameters:
CSST C and D: Taps +/-10%, Tap Step 1.25%, Total No of Taps 17, Initial Time Delay 2 seconds, Operating Time 1 second. Taps are provided on each secondary winding.
CSST A and B: Taps +/-16.8%, Tap Step 1.05%, Total No of Taps 33, Initial Time Delay 1 second, Operating Time 2 seconds.
Taps are provided on the primary winding.
The analysis evaluates the 6.9-kV shutdown board minimum voltage requirements considering a maximum (bounding) grid voltage drop of 9 kV and a minimum grid voltage of 153kV and all plant conditions. Although the calculated shutdown board voltage falls below the degraded voltage relay dropout setpoint due to block start of ESF motors, it recovers above the degraded voltage relay reset setpoint in 5 seconds. The minimum time for the degraded voltage relays to isolate the offsite power from the 6.9kV Shutdown Boards is 8.5 seconds.
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ENCLOSURE 1 Response to Action Items From Appendix HH of NUREG-0847, Supplement 22 Tennessee Valley Authority - Watts Bar Nuclear Plant - Unit 2, Docket No. 50-391 Attachment 3 provides the Electrical Transient Analysis Program (ETAP) voltage recovery plots following a DBE on one unit while the other unit is in simultaneous orderly shutdown. These plots pictorially depict the LTC function at different times following a DBE.
During normal operation and post-accident with bounding grid voltage (153kV), the voltage on the 6.9kV shutdown boards is maintained within the LTC control band. As shown in the ETAP plots, the voltage on the shutdown boards falls below the degraded voltage relay setpoint due to block start of ESF motors but recovers to a value above the degraded voltage relay reset value before the degraded voltage relay timer times out so as not to isolate the shutdown boards from the offsite power. The source is therefore in compliance with GDC 17 and is able to supply offsite power to 1E loads with an accident in one unit, safe shutdown of the opposite unit, and the worst case single failure.
- 31. TVA should evaluate the re-sequencing of loads, with time delays involved, in the scenario of a LOCA followed by a delayed LOOP, and ensure that all loads will be sequenced within the time assumed in the accident analysis. (Section 8.3.1.11)
Response: LOCA followed by LOOP TVA to NRC letter dated December 6, 2010, Watts Bar Nuclear Plant (WBN)
Unit 2 - Safety Evaluation Report Supplement 22 (SSER22) - Response to Requests for Additional Information, (ADAMS accession number ML103420569) included the response to RAI 8.3.1.11. That response stated, A LOCA followed by a delayed LOOP is not a Design Basis Event for WBN.
The design basis for WBN assumes a simultaneous LOOP - LOCA. The Hydraulic Analysis does not support a LOCA with a delayed LOOP event; however, the logic is designed to ensure that loads are re-sequenced during a LOCA with a delayed LOOP, to prevent a block start on a diesel generator.
This logic does not impact the sequencing for the design bases event, simultaneous LOOP - LOCA.
When the LOOP occurs, the diesel will start, based on detection by the Loss of Voltage relay. Loads which sequence on due to a blackout signal (Charging Pump, Auxiliary Feedwater, Essential Raw Cooling Water Pump, Closed Cooling, etc.) will begin sequencing on.
When a subsequent LOCA signal occurs, the diesel will remain running and connected to the Shutdown Board. Loads which are required for accident mitigation and which have previously sequenced on to the Shutdown Board, due to the LOOP, will remain running. Loads which are not required for accident mitigation will be tripped. Remaining loads required for accident mitigation, which have not been sequenced on at the time of the LOCA, will have their timers reset to 0 and will sequence on at the appropriate time for the LOCA signal.
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ENCLOSURE 1 Response to Action Items From Appendix HH of NUREG-0847, Supplement 22 Tennessee Valley Authority - Watts Bar Nuclear Plant - Unit 2, Docket No. 50-391 LOCA - Delayed LOOP When the LOCA occurs, the loads which are not running in normal operation will block start. At the same time, the diesels will start on the LOCA signal, but will not tie to the Shutdown Board.
When a subsequent LOOP occurs, all sequenced loads will be stripped from the board from a Loss of Voltage (approximately 86%) signal. Once the loss of voltage relay has reached its set point and the diesel is available, the diesel breaker will close and the sequence timers will begin to time. The first large motor (Centrifugal Charging Pump) connects at 5 seconds and is followed by the remaining accident required loads. This provides assurance that the voltage has decayed on the boards and no residual out of phase reconnection occurs.
- 33. TVA stated in Attachment 9 of its letter dated July 31, 2010, that certain design change notices (DCNs) are required or anticipated for completion of WBN Unit 2, and that these DCNs were unverified assumptions used in its analysis of the 125 V dc vital battery system.
Verification of completion of these DCNs to the NRC staff is necessary prior to issuance of the operating license. (Section 8.3.2.3)
Response: The applicable DCNs are as follow:
x DCN 53421 for the removal/abandonment of Reciprocating Charging Pump 2-MTR-62-101, supplied from 480V SHDN BD 2B1-B, Compt. 3B, has been issued.
x DCN 54636 for the cable modifications for Unit 2 AFWP Turbine Trip and Throttle Valve and Turbine Controls has been issued.
NRC will be notified when the physical work has been completed for these two DCNs.
- 34. TVA stated that the method of compliance with Phase I guidelines would be substantially similar to the current Unit 1 program and that a new Section 3.12 will be added to the Unit 2 FSAR that will be materially equivalent to Section 3.12 of the current Unit 1 FSAR.
(Section 9.1.4)
Response: Amendment 103 to the Unit 2 FSAR added new Section 3.12 (Control of Heavy Loads). This new section is materially equivalent to Section 3.12 of the Unit 1 UFSAR.
Amendment 103 was submitted via TVA to NRC letter dated March 15, 2011, Watts Bar Nuclear Plant (WBN) - Unit 2 - Final Safety Analysis Report (FSAR), Amendment 103.
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ENCLOSURE 1 Response to Action Items From Appendix HH of NUREG-0847, Supplement 22 Tennessee Valley Authority - Watts Bar Nuclear Plant - Unit 2, Docket No. 50-391
- 36. TVA should provide information to the NRC staff to enable verification that the SGBS meets the requirements and guidance specified in the SER or provide justification that the SGBS meets other standards that demonstrate conformance to GDC 1 and GDC 14.
(Section 10.4.8)
Response: Section 2.1.1, Safety Functions, of the SGB System Description Documents N3-15-4002 (Unit 1) and WBN2-15-4002 (Unit 2), state the following:
The SGB piping downstream of the containment isolation valves and located in the main stream valve vault room shall be TVA Class G. This piping is seismically supported to maintain the pressure boundary.
The SGB piping located in the turbine building shall be TVA Class H.
The Unit 1 and Unit 2 SGB flow diagrams, 1, 2-47W801-2, also recognize the same TVA Class G and Class H class breaks located downstream of the safety-related SGB containment isolation valves.
The SGB flow diagrams and System Description document that TVA Class G and Class H classifications located downstream of the safety-related containment isolation valves are consistent with the data that was deleted in FSAR Section 10.4.8.1, Steam Generator Blowdown System - Design Basis, Item 6 Component and Code listings described above. It is also noted that NRC Quality Group D classification is equivalent to TVA Class G and H classifications as stated in the NUREG 0847 Section 3.2.2, System Quality Group Classification. Therefore, the design requirements in NRC GDC-1, Quality Standards and Records, and NRC GDC-14, Reactor Coolant Pressure Boundary are not challenged.
Amendment 104 to the Unit 2 FSAR will revise Table 3.2-2 to note that TVA Class G and H piping within the SGB System exists downstream of the safety-related containment isolation valves.
- 44. TVA should provide additional information to clarify how the initial and irradiated RTNDT was determined. (Section 5.3.1)
Response: This response clarifies how the initial and irradiated RTNDT values were determined for the Watts Bar Unit 2 reactor pressure vessel beltline materials.
Unit 2 FSAR Section 5.2.4.1 established that the vessel was designed to 1971 Addenda of the ASME Code, an edition that predates the requirements to determine the unirradiated RTNDT. (Those requirements were established in the Summer 1972 Addenda to the Code,Section III, Subarticle NB-2300, whereas the Watts Bar Unit 2 vessel was designed to an earlier version of the Code.) Because the tests performed to assess the adequacy of the fracture toughness predated the Summer 1972 Addenda to the Code, it was necessary to use the methods described in NRC Branch Technical Position (BTP) Materials Engineering Branch (MTEB) 5-2, Fracture Toughness Requirements for Older Plants. For the Watts Bar Unit 2 vessel, the vessel shell materials were tested by the vessel fabricator using both drop-weight E1-10
ENCLOSURE 1 Response to Action Items From Appendix HH of NUREG-0847, Supplement 22 Tennessee Valley Authority - Watts Bar Nuclear Plant - Unit 2, Docket No. 50-391 and Charpy impact test specimens. The drop-weight specimens were tested to determine the unirradiated nil-ductility transition temperature (NDTT) in accordance with ASTM E 208. In the ASME Code,Section III, Subarticle NB-2300, the NDTT is used with axial (weak) orientation Charpy test data to determine the initial (unirradiated) RTNDT. For Watts Bar Unit 2, the orientation of the Charpy impact test specimens was in the tangential (strong) orientation rather than in the axial (weak) orientation currently required in NB-2300 to determine the initial RTNDT. BTP MTEB 5-2 provides methods to determine the initial RTNDT using the drop-weight and Charpy impact test results generated for the Watts Bar Unit 2 vessel shell forgings and welds. In summary, both drop-weight and Charpy impact specimens in the tangential (strong) orientation were tested and the results were evaluated to determine the initial RTNDT following the methods in NRC BTP MTEB 5-2.
In addition to those tests performed by the vessel fabricator, unirradiated tests were performed on the Watts Bar Unit 2 reactor vessel surveillance program materials. Tests consisted of Charpy impact specimens from the intermediate shell forging and the core region metal that were oriented in both the tangential (strong) and axial (weak) orientations. When the surveillance program Charpy impact specimens are used with the drop-weight NDTT values obtained by the vessel fabricator, the initial RTNDT values obtained using NRC BTP MTEB 5-2 are found to be conservative.
The irradiated RTNDT, termed the Adjusted Reference Temperature (ART), is used to establish the Pressure-Temperature (P-T) limit curves for the vessel as documented in the Pressure and Temperature Limits Report (PTLR). The PTLR for Watts Bar Unit 2 is discussed in Unit 2 FSAR Section 5.2.4.3. The initial P-T limit curves are based on predictions of the effects of irradiation using the methods in NRC Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials. As post-irradiation test results become available from the evaluation of test specimens from the Watts Bar Unit 2 reactor vessel surveillance program, ASTM E 185-82, Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels, uses those test results to assess the accuracy and conservatism of the predictions based on the methods of NRC Regulatory Guide 1.99, Revision 2. The reactor vessel irradiation surveillance program for Watts Bar Unit 2 is discussed in Unit 2 FSAR Section 5.4.3.6. The effect of irradiation is measured using the Charpy impact specimens. Note that there are no drop-weight test specimens irradiated as part of the Watts Bar Unit 2 surveillance program. The drop-weight specimens are used only for tests on the unirradiated material to determine the drop-weight NDTT.
In summary, both drop-weight and Charpy impact specimens (strong orientation) were tested and the results were evaluated to determine the initial (unirradiated) RTNDT following the methods in NRC BTP MTEB 5-2.
Additional tests performed as part of the reactor vessel surveillance program using Charpy impact specimens (weak orientation for the intermediate shell forging), and those data obtained following the ASME Code,Section III, Subarticle NB-2300 demonstrated the initial RTNDT following the methods in E1-11
ENCLOSURE 1 Response to Action Items From Appendix HH of NUREG-0847, Supplement 22 Tennessee Valley Authority - Watts Bar Nuclear Plant - Unit 2, Docket No. 50-391 NRC BTP MTEB 5-2 to be conservative. The irradiated RTNDT, termed the ART, will be determined using the methods in NRC Regulatory Guide 1.99.
As post-irradiation test results become available from the reactor vessel surveillance program materials (the intermediate shell forging and the core region weld metal), those data will be used to assess the accuracy and conservatism of the predictions.
- 45. TVA stated in its response to RAI 5.3.2-2, dated July 31, 2010, that the PTLR would be revised to incorporate the COMS arming temperature. (Section 5.3.2)
Response: Revision 1 (effective August 12, 2010) to the Unit 2 System Description for the Reactor Coolant System (WBN2-68-4001) was revised to reflect the required revisions to the PTLR. Appendix B, Section 3.2 (Arming Temperature) states, COMS shall be armed when any RCS cold leg temperature is <225°F.
- 46. The LTOP lift settings were not included in the PTLR, but were provided in TVAs response to RAI 5.3.2-2 in its letter dated July 31, 2010. TVA stated in its RAI response that the PTLR would be revised to incorporate the LTOP lift settings into the PTLR. (Section 5.3.2)
Response: Revision 1 (effective August 12, 2010) to the Unit 2 System Description for the Reactor Coolant System (WBN2-68-4001) was revised to reflect the required revisions to the PTLR. Appendix B, TABLE 3.1-1 (Watts Bar Unit 2 PORV Setpoints vs Temperature) contains the lift settings.
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ENCLOSURE 2 Action Items From Appendix HH of NUREG-0847, Supplement 22 to Be Answered Later Tennessee Valley Authority - Watts Bar Nuclear Plant - Unit 2, Docket No. 50-391 The item numbers used below correspond to item numbers in SSER 22, Appendix HH.
- 10. Confirm that TVA has an adequate number of licensed and non-licensed operators in the training pipeline to support the preoperational test program, fuel loading, and dual unit operation. (Section 13.1.3)
- 13. TVA is expected to submit an IST program and specific relief requests for WBN Unit 2 nine months before the projected date of OL issuance. (Section 3.9.6)
- 25. Prior to the issuance of an operating license, TVA is required to provide satisfactory documentation that it has obtained the maximum secondary liability insurance coverage pursuant to 10 CFR 140.11(a)(4), and not less than the amount required by 10 CFR 50.54(w) with respect to property insurance, and the NRC staff has reviewed and approved the documentation. (Section 22.3)
- 29. TVA should provide the transmission system specifics (grid stability analyses) to the NRC staff. In order to verify compliance with GDC 17, the results of the grid stability analyses must indicate that loss of the largest electric supply to the grid, loss of the largest load from the grid, loss of the most critical transmission line, or loss of both units themselves, will not cause grid instability. (Section 8.2.2)
- 30. TVA should confirm that all other safety-related equipment (in addition to the Class 1E motors) will have adequate starting and running voltage at the most limiting safety related components (such as motor operated valves, contactors, solenoid valves or relays) at the degraded voltage relay setpoint dropout setting. TVA should also confirm that the final Technical Specifications are properly derived from these analytical values for the degraded voltage settings. (Section 8.3.1.2)
- 32. TVA should provide to the NRC staff the details of the administrative limits of EDG voltage and speed range, and the basis for its conclusion that the impact is negligible, and describe how it accounts for the administrative limits in the Technical Specification surveillance requirements for EDG voltage and frequency. (Section 8.3.1.14)
- 35. TVA should provide information to the NRC staff that the CCS will produce feedwater purity in accordance with BTP MTEB 5-3 or, alternatively, provide justification for producing feedwater purity to another acceptable standard. (Section 10.4.6)
- 37. The NRC staff will review the combined WBN Unit 1 and 2 Appendix C prior to issuance of the Unit 2 OL to confirm (1) that the proposed Unit 2 changes were incorporated into Appendix C, and (2) that changes made to Appendix C for Unit 1 since Revision 92 and the changes made to the NP-REP since Revision 92 do not affect the bases of the staffs findings in this SER supplement. (Section 13.3.2)
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ENCLOSURE 2 Action Items From Appendix HH of NUREG-0847, Supplement 22 to Be Answered Later Tennessee Valley Authority - Watts Bar Nuclear Plant - Unit 2, Docket No. 50-391
- 43.Section V of Appendix E to 10 CFR Part 50 requires TVA to submit its detailed implementing procedures for its emergency plan no less than 180 days before the scheduled issuance of an operating license. Completion of this requirement will be confirmed by the NRC staff prior to the issuance of an operating license. (Section 13.3.2.18)
- 47. The NRC staff noted that TVAs changes to Section 6.2.6 in FSAR Amendment 97, regarding the implementation of Option B of Appendix J, were incomplete, because several statements remained regarding performing water-sealed valve leakage tests as specified in 10 CFR
[Part] 50, Appendix J. With the adoption of Option B, the specified testing requirements are no longer applicable; Option A to Appendix J retains these requirements. The NRC discussed this discrepancy with TVA in a telephone conference on September 28, 2010. TVA stated that it would remove the inaccurate reference to Appendix J for specific water testing requirements in a future FSAR amendment. (Section 6.2.6)
E2-2
ENCLOSURE 3 Action Items From Appendix HH of NUREG-0847, Supplement 22 For NRC Inspection / Review Tennessee Valley Authority - Watts Bar Nuclear Plant - Unit 2, Docket No. 50-391 The item numbers used below correspond to item numbers in SSER 22, Appendix HH.
- 1. Review evaluations and corrective actions associated with a power assisted cable pull. (NRC safety evaluation dated August 31, 2009, ADAMS Accession No. ML092151155)
- 2. Conduct appropriate inspection activities to verify cable lengths used in calculations and analysis match as-installed configuration. (NRC safety evaluation dated August 31, 2009, ADAMS Accession No. ML092151155)
- 4. Conduct appropriate inspection activities to verify that TVAs maximum SWBP criteria for signal level and coaxial cables do not exceed the cable manufacturers maximum SWBP criteria. (NRC safety evaluation dated August 31, 2009, ADAMS Accession No.
- 7. Verify commitment completion and review electrical design calculations. (TVA letter dated October 9, 1990, ADAMS Accession No. ML073551056)
- 9. Confirm that education and experience of management and principal supervisory positions down through the shift supervisory level conform to Regulatory Guide 1.8. (Section 13.1.3)
- 11. The plant administrative procedures should clearly state that, when the Assistant Shift Engineer assumes his duties as Fire Brigade Leader, his control room duties are temporarily assumed by the Shift Supervisor (Shift Engineer), or by another SRO, if one is available. The plant administrative procedures should clearly describe this transfer of control room duties.
(Section 13.1.3)
- 12. TVAs implementation of NGDC PP-20 and EDCR Appendix J is subject to future NRC audit and inspection. (Section 25.9)
- 16. Based on the uniqueness of EQ, the NRC staff must perform a detailed inspection and evaluation prior to fuel load to determine how the WBN Unit 2 EQ program complies with the requirements of 10 CFR 50.49. (Section 3.11.2)
(Section 3.11.2.1)
- 19. The NRC staff should complete its review of TVAs EQ Program procedures for WBN Unit 2 prior to fuel load. (Section 3.11.2.2.1)
- 21. The NRC staff should confirm that the Electrical Penetration Assemblies (EPAs) are installed in the tested configuration, and that the feedthrough module is manufactured by the same company and is consistent with the EQ test report for the EPA. (Section 3.11.2.2.1)
E3-1
ENCLOSURE 3 Action Items From Appendix HH of NUREG-0847, Supplement 22 For NRC Inspection / Review Tennessee Valley Authority - Watts Bar Nuclear Plant - Unit 2, Docket No. 50-391
- 38. The NRC staff will confirm the availability and operability of the ERDS for Unit 2 prior to issuance of the Unit 2 OL. (Section 13.3.2.6)
- 39. The NRC staff will confirm the adequacy of the communications capability to support dual unit operations prior to issuance of the Unit 2 OL. (Section 13.3.2.6)
- 40. The NRC staff will confirm the adequacy of the emergency facilities and equipment to support dual unit operations prior to issuance of the Unit 2 OL. (Section 13.3.2.8)
- 41. TVA committed to (1) update plant data displays as necessary to include Unit 2, and (2) to update dose assessment models to provide capabilities for assessing releases from both WBN units. The NRC staff will confirm the adequacy of these items prior to issuance of the Unit 2 OL. (Section 13.3.2.9)
- 42. The NRC staff will confirm the adequacy of the accident assessment capabilities to support dual unit operations prior to issuance of the Unit 2 OL. (Section 13.3.2.9)
- 48. The NRC staff should verify that its conclusions in the review of FSAR Section 15.4.1 do not affect the conclusions of the staff regarding the acceptability of Section 6.5.3. (Section 6.5.3)
- 49. The NRC staff was unable to determine how TVA linked the training qualification requirements of ANSI N45.2-1971 to TVA Procedure TI-119. Therefore, the implementation of training and qualification for inspectors will be the subject of future NRC staff inspections.
(NRC letter dated July 2, 2010, ADAMS Accession No. ML101720050)
- 50. TVA stated that about 5 percent of the anchor bolts for safety-related pipe supports do not have quality control documentation, because the pull tests have not yet been performed.
Since the documentation is still under development, the NRC staff will conduct inspections to follow-up on the adequate implementation of this construction refurbishment program requirement. (NRC letter dated July 2, 2010, ADAMS Accession No. ML101720050)
- 51. The implementation of TVA Procedure TI-119 will be the subject of NRC follow-up inspection to determine if the construction refurbishment program requirements are being adequately implemented. (NRC letter dated July 2, 2010, ADAMS Accession No. ML101720050)
E3-2
ENCLOSURE 4 List of Regulatory Commitments Tennessee Valley Authority - Watts Bar Nuclear Plant - Unit 2, Docket No. 50-391
- 2. NRC will be notified when the physical work has been completed for DCNs 53421 and 54636. (response to action item 33)
- 3. Amendment 104 to the Unit 2 FSAR will revise Table 3.2-2 to note that TVA Class G and H piping within the SGB System exists downstream of the safety-related containment isolation valves. (response to action item 36)
E4-1
Tennessee Valley Authority - Watts Bar Nuclear Plant - Unit 2, Docket No. 50-391 Attachment 1 Response to TVA for Core Loading Information and End of Cycle Assembly Burnup Data for Watts Bar Unit 2 Cycle 1
WESTINGHOUSE ELECTRIC COMPANY LLC Attachment to Calculation Note Number Revision Page CN-WB01-021 0 A-1 Response to TVA for Core Loading Information and End of Cycle Assembly Burnup Data for Watts Bar Unit 2 Cycle 1 Authored:
T. A. Jones (ND)*
Nuclear Design A Verified:
R. N. Milanova (ND)*
Nuclear Design A Approved:
D. E. Sipes*
Manager, Nuclear Design A
Attachment:
4 pages
- Electronically approved records are authenticated in the Electronic Document Management System.
2011 Westinghouse Electric Company LLC All Rights Reserved CE Word Version 14-1
WESTINGHOUSE ELECTRIC COMPANY LLC Attachment to Calculation Note Number Revision Page CN-WB01-021 0 A-2 Reference(s): 1) CN-WB01-010, Revision 1, Revision 1 ** Watts Bar Unit 2 Cycle 1 (WBT01) ANC 8 Model and Core Loading Plan (CLP) Generation - Short Form Revision Figure 1 provides the core loading pattern for WBT01, while Figure 2 provides the control and shutdown rod locations. Figure 3 provides assembly average burnups for all core locations, at the maximum analyzed Cycle 1 burnup (17675 MWD/MTU).
CE Word Version 14-1
WESTINGHOUSE ELECTRIC COMPANY LLC Attachment to Calculation Note Number Revision Page CN-WB01-021 0 A-3 CE Word Version 14-1
WESTINGHOUSE ELECTRIC COMPANY LLC Attachment to Calculation Note Number Revision Page CN-WB01-021 0 A-4 CE Word Version 14-1
WESTINGHOUSE ELECTRIC COMPANY LLC Attachment to Calculation Note Number Revision Page CN-WB01-021 0 A-5 FIGURE 3 WATTS BAR UNIT 2, CYCLE 1 ASSEMBLY AVERAGE BURNUP AT MAXIMUM ANALYZED CYCLE BURNUP (17675 MWD/MTU)
- / _ l#/ l/ / ___/## Version Job No. Date/Time Case No. Ref. Time Ref. BU Title
- / __ l/ / /##### ======= ======= ================= ========= ============= =========== =========
- /_/#l_/_/l_/l___/## 9.3.0 2147 03/22/11 11:55:01 16 10968.1 17675 WBT01 ANC
CA-Burnup Assembly Burnup 180 deg 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 1 9777 11835 13123 12316 13123 11835 9777 2 10072 14620 16933 16986 18591 17543 18591 16986 16933 14620 10072 3 10072 16055 17589 18595 19616 19446 19993 19446 19616 18595 17589 16055 10072 4 14620 17589 18934 20143 20026 20695 20230 20695 20026 20143 18934 17589 14620 5 9777 16933 18595 20143 20165 20880 20409 20966 20409 20880 20165 20143 18595 16933 9777 6 11835 16986 19616 20026 20880 20400 20632 20374 20632 20400 20880 20026 19616 16986 11835 7 13123 18591 19446 20695 20409 20632 20315 20567 20315 20632 20409 20695 19446 18591 13123 8 12316 17543 19993 20230 20966 20374 20567 20287 20567 20374 20966 20230 19993 17543 12316 9 13123 18591 19446 20695 20409 20632 20315 20567 20315 20632 20409 20695 19446 18591 13123 10 11835 16986 19616 20026 20880 20400 20632 20374 20632 20400 20880 20026 19616 16986 11835 11 9777 16933 18595 20143 20165 20880 20409 20966 20409 20880 20165 20143 18595 16933 9777 12 14620 17589 18934 20143 20026 20695 20230 20695 20026 20143 18934 17589 14620 13 10072 16055 17589 18595 19616 19446 19993 19446 19616 18595 17589 16055 10072 14 10072 14620 16933 16986 18591 17543 18591 16986 16933 14620 10072 15 9777 11835 13123 12316 13123 11835 9777 Max Min Maximum Loc Minimum Loc
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20966 5-8 9777 1-5 End-Edit CE Word Version 14-1
Tennessee Valley Authority - Watts Bar Nuclear Plant - Unit 2, Docket No. 50-391 Attachment 2 A Review of Electronic Components in a Radiation Environment of 5x104 RADS
Tennessee Valley Authority - Watts Bar Nuclear Plant - Unit 2, Docket No. 50-391 Attachment 3 Electrical Transient Analysis Program (ETAP) Voltage Recovery Plots
SOURCE: CSST D, CSST C Similar Unit 1 SI Phase B, Unit 2 Normal Operation Voltage Recovery Plot- 6.9kV Shutdown Board 1A-A LTC TAPS (SECONDARY WINDING)
Initial Time Delay - 2 Seconds OPERATING TIME - 1 Second for each step Number of Steps= 17 TAP per Step 1.25%
MIn/Max -10%/ +10%
Degraded Voltage Relay Resets @ 6672 Volts Degraded Voltage Relay Reset Time
< or = to 5 Degraded Voltage Relay seconds Reset Time Minimum 8.5 Seconds
SOURCE: CSST D, CSST C Similar Unit 1 SI Phase B, Unit 2 Normal Operation Voltage Recovery Plot- 6.9kV Shutdown Board 1B-B LTC TAPS (SECONDARY WINDING)
Initial Time Delay - 2 Seconds OPERATING TIME - 1 Second for each step Number of Steps= 17 TAP per Step 1.25%
MIn/Max -10%/ +10%
Degraded Voltage Relay Resets @ 6672 Volts Degraded Degraded Voltage Voltage Relay Relay Reset Time Reset Time Minimum 8.5 Seconds
< or = to 5 seconds
SOURCE: CSST D, CSST C Similar Unit 2 SI Phase B, Unit 1 Normal Operation Voltage Recovery Plot- 6.9kV Shutdown Board 2A-A LTC TAPS (SECONDARY WINDING)
Initial Time Delay - 2 Seconds OPERATING TIME - 1 Second for each step Number of Steps= 17 TAP per Step 1.25%
MIn/Max -10%/ +10%
Degraded Voltage Relay Resets @ 6672 Volts Degraded Voltage Relay Degraded Voltage Reset Time Relay Reset Time
< or = to 5 Minimum 8.5 seconds Seconds
SOURCE: CSST D, CSST C Similar Unit 2 SI Phase B, Unit 1 Normal Operation Voltage Recovery Plot- 6.9kV Shutdown Board 2B-B LTC TAPS (SECONDARY WINDING)
Initial Time Delay - 2 Seconds OPERATING TIME - 1 Second for each step Number of Steps= 17 TAP per Step 1.25%
MIn/Max -10%/ +10%
Degraded Voltage Relay Resets @ 6672 Volts Degraded Degraded Voltage Voltage Relay Relay Reset Time Reset Time Minimum 8.5 Seconds
< or = to 5 seconds
SOURCE: CSST A OR B Unit 1 SI Phase B, Unit 2 Normal Operation Voltage Recovery Plot- 6.9kV Shutdown Board 1A-A LTC TAPS (PRIMARY WINDING)
Initial Time Delay - 1 Seconds OPERATING TIME - 2 Second for each step Number of Steps= 33 TAP per Step 1.05%
MIn/Max -16.8%/ +16.8%
Degraded Voltage Relay Resets @ 6672 Volts Degraded Voltage Relay Reset Time Degraded Voltage Minimum 8.5 Relay Reset Time Seconds
< or = to 3.5 seconds
SOURCE: CSST A OR B Unit 1 SI Phase B, Unit 2 Normal Operation Voltage Recovery Plot- 6.9kV Shutdown Board 1B-B LTC TAPS (PRIMARY WINDING)
Initial Time Delay - 1 Seconds OPERATING TIME - 2 Second for each step Number of Steps= 33 TAP per Step 1.05%
MIn/Max -16.8%/ +16.8%
Degraded Voltage Relay Resets @ 6672 Volts Degraded Voltage Degraded Voltage Relay Reset Time Relay Reset Time
< or = to 3.5 Minimum 8.5 seconds Seconds
SOURCE: CSST A OR B Unit 2 SI Phase B, Unit 1 Normal Operation Voltage Recovery Plot- 6.9kV Shutdown Board 2A-A LTC TAPS (PRIMARY WINDING)
Initial Time Delay - 1 Seconds OPERATING TIME - 2 Second for each step Number of Steps= 33 TAP per Step 1.05%
Degraded Voltage MIn/Max -16.8%/ +16.8%
Relay Resets @ 6672 Volts Degraded Voltage Degraded Voltage Relay Reset Time Relay Reset Time
< or = to 3.5 seconds Minimum 8.5 Seconds
SOURCE: CSST A OR B Unit 2 SI Phase B, Unit 1 Normal Operation Voltage Recovery Plot- 6.9kV Shutdown Board 2B-B LTC TAPS (PRIMARY WINDING)
Initial Time Delay - 1 Seconds OPERATING TIME - 2 Second for each step Number of Steps= 33 TAP per Step 1.05%
MIn/Max -16.8%/ +16.8%
Degraded Voltage Relay Resets @ 6672 Volts Degraded Voltage Degraded Voltage Relay Reset Time Relay Reset Time
< or = to 3.5 seconds Minimum 8.5 Seconds