ML103140661
| ML103140661 | |
| Person / Time | |
|---|---|
| Site: | Watts Bar  |
| Issue date: | 10/21/2010 |
| From: | Office of Nuclear Reactor Regulation |
| To: | Division of Operating Reactor Licensing |
| References | |
| Download: ML103140661 (22) | |
Text
1 WBN2Public Resource From:
Poole, Justin Sent:
Thursday, October 21, 2010 3:32 PM To:
Epperson, Dan Cc:
WBN2HearingFile Resource
Subject:
FW: 10-21-10_I&C RAI Response Final_NRC Copy.pdf Attachments:
10-21-10_I&C RAI Response Final_NRC Copy.pdf JustinC.Poole ProjectManager NRR/DORL/LPWB U.S.NuclearRegulatoryCommission (301)4152048 email:Justin.Poole@nrc.gov From: Boyd, Desiree L [1]
Sent: Thursday, October 21, 2010 11:57 AM To: Wiebe, Joel; Poole, Justin; Raghavan, Rags; Milano, Patrick; Campbell, Stephen Cc: Boyd, Desiree L; Hamill, Carol L
Subject:
10-21-10_I&C RAI Response Final_NRC Copy.pdf Please see attached letter that was mailed to the NRC today.
Also, Enclosure 2 attachments are not included, but I have sent a disk with the hardcopies.
~*~*~*~*~*~*~*~*~*~*~*~*~*~*~
Désireé L. Boyd WBN2LicensingSupport SunTechnicalServices dlboyd@tva.gov 4233658764
~*~*~*~*~*~*~*~*~*~*~*~*~*~*~
Hearing Identifier:
Watts_Bar_2_Operating_LA_Public Email Number:
178 Mail Envelope Properties (19D990B45D535548840D1118C451C74D6FD37239CF)
Subject:
FW: 10-21-10_I&C RAI Response Final_NRC Copy.pdf Sent Date:
10/21/2010 3:31:34 PM Received Date:
10/21/2010 3:31:35 PM From:
Poole, Justin Created By:
Justin.Poole@nrc.gov Recipients:
"WBN2HearingFile Resource" <WBN2HearingFile.Resource@nrc.gov>
Tracking Status: None "Epperson, Dan" <Dan.Epperson@nrc.gov>
Tracking Status: None Post Office:
HQCLSTR02.nrc.gov Files Size Date & Time MESSAGE 836 10/21/2010 3:31:35 PM 10-21-10_I&C RAI Response Final_NRC Copy.pdf 265198 Options Priority:
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Recipients Received:
U.S. Nuclear Regulatory Commission Page 2 October 21, 2010 cc (Enclosures):
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 October 21, 2010 GPA:ETK:CLH bcc (Enclosures):
Stephen Campbell U.S. Nuclear Regulatory Commission MS 08H4A One White Flint North 11555 Rockville Pike Rockville, Maryland 20852-2738 Hukam C. Garg U.S. Nuclear Regulatory Commission MS 09D2 One White Flint North 11555 Rockville Pike Rockville, Maryland 20852-2738 William E. Kemper U.S. Nuclear Regulatory Commission MS 09E3 One White Flint North 11555 Rockville Pike Rockville, Maryland 20852-2738 Justin C. Poole U.S. Nuclear Regulatory Commission MS 08H4A One White Flint North 11555 Rockville Pike Rockville, Maryland 20852-2738 Loren R. Plisco, 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
U.S. Nuclear Regulatory Commission Page 4 October 21, 2010 GPA:ETK:CLH bcc (Enclosures):
M. Bajestani, EQB 1B-WBN*
R. R. Baron, EQB 1B-WBN*
M. K. Brandon, ADM 1L-WBN*
W. D. Crouch, EQB 1B-WBN D. E. Grissette, ADM 1V-WBN*
E. J. Vigluicci, WT 6A-K K. W. Whittenburg, SP 2B-C*
S. A. Hilmes, EQB 1B-WBN*
EDMS, WT 3B-K (Re: T02 101005 001; T02 100811 001)
- These CCs did not receive Enclosure 2. The attached documents can be obtained by contacting the WBN Unit 2 Licensing office.
TVA Letter Dated October 21, 2010 Responses to Licensee Open Items to be Resolved for SER Approval 1.
Initial NRC Request (Item Number 47)
The PAMS System Requirements Specification (SysRS) references RG 1.97 Rev. 3 where the FSAR References Rev. 2. Please explain.
Follow-up NRC Request:
How could Westinghouse Design, and TVA approve a design to the wrong requirement?
TVA Revised Response:
The difference in revisions of RG 1.97 was not identified during the contract review process. Therefore, Westinghouse Electric Company (WEC) designed the system to the Common Q standard design, which is Revision 3. When the design work was assigned to a new engineer, the difference in revisions was not identified as an issue. When the issue was identified by the NRC, it was entered into the TVA Corrective Action Process as WBPER233598 (Attachment 1).
2.
NRC Request (Item Number 153)
FSAR section 7.2.1.1.7 added the reference to FSAR section 10.4.4.3 for exception to P-12. However, FSAR section 10.4.4.3 states bypass condition is not displayed and it is not automatically removed when conditions for bypass are no longer met. Provide the basis for this.
Additional NRC Comment on preliminary response:
TVA did not address why bypass condition is not displayed.
TVA Response:
WBN Unit 1 Engineering Design Change (EDC) E50952-A added an alternate method of Reactor Coolant System (RCS) cooldown using additional steam dump valves after entering Mode 4, by disabling the P-12 Interlock. Operators use additional condenser dump valves to aid in maintaining a cooldown rate closer to the administrative limit established by operating procedures. The WBN Unit 2 FSAR has implemented the same change to allow bypassing of the P-12 interlock.
Refer to Unit 1 UFSAR Amendment 3 Change Package 1676 S00 (Attachment 2) for the safety evaluation and basis for this change. The 50.59 for the change is included in the Change Package.
The process is controlled by the procedures used to shutdown the plant. The procedure initiates the bypass by lifting a wire at a relay and then ensures that it is restored prior to starting the plant. GO-6 Unit Shutdown From Hot Standby To Cold Shutdown, Appendix F prescribes the steps required to bypass and restore the P-12 interlock. GO-6 Appendix F, step 3.0[6] says PLACE Caution Order on 1-HS-1-103A, 1-HS-1-103B, AND 1-PIC-1-33 indicating that P-12 interlock is disabled. This provides indication to the operators that the P-12 interlock is bypassed. GO-6 Appendix F, step 4.0[1] says WHEN steam dumps no longer affect plant cooldown OR plant heatup is desired, THEN ENSURE MIG re-enables E1-1 TVA Letter Dated October 21, 2010 Responses to Licensee Open Items to be Resolved for SER Approval P-12 Interlock USING Section 6.0 of this Appendix, LANDING LEADS TO RE-ENABLE P-12 INTERLOCK. GO-6 Appendix F, step 4.0[2] states REMOVE Caution Order on 1-HS-1-103A, 1-HS-1-103B, AND 1-PIC-1-33 indicating that P-12 interlock is disabled.
3.
NRC Request (Item Number 171)
An external unidirectional communications interface was installed between the Eagle 21 test subsystem and the plant process computer. TVA should confirm that testing has demonstrated that two way communication is impossible with the described configuration.
(Open Item # 3 of Eagle 21 audit)
TVA Response:
The external Eagle 21 unidirectional communications interface will be tested prior to WBN Unit 2 fuel load.
- 4.
NRC Request (Item Number 183)
An emphasis is placed on traceability in System Requirements Specifications in the SRP, in the unmodified IEEE std 830-1993, and even more so given the modifications to the standard listed in Regulatory Guide 1.172, which breaks with typical NRC use of the word should to say Each identifiable requirement in an SRS must be traceable backwards to the system requirements and the design bases or regulatory requirements that is satisfies
[sic].
On page 1-2 of the Post Accident Monitoring Systems Software Requirements Specification in the background section, is the sentence Those sections of the above references that require modification from the generic PAMS are defined in the document referring purely to the changes from WNA-DS-01617-WBT Post Accident Monitoring System-System Requirements Specification or is it saying that there are additional changes beyond those and that the SRS defines them?
If there are additional changes, what is their origin?
TVA Response:
The generic Software Requirements Specification applies except as modified by the WBN Unit 2 System Requirements Specification. There are no additional changes.
- 5.
NRC Request (Item Number 254)
Please make the following available in Westinghouse's Rockville office.
WNA-PD-00056-WBT, Rev 1 Watts Bar Unit 2 NSSS Completion I&C Projects As the indicated source of customer specific requirements for deliverables, as indicated in the project plan, this document may serve as one end of a thread audit, and may contain information relevant to evaluating the completeness of later requirements.
E1-2 TVA Letter Dated October 21, 2010 Responses to Licensee Open Items to be Resolved for SER Approval 956080, Rev 1. Cabinet mounted electronics - Inadequate core cool monitor (ICCM-86)
Believe this to be the source of the requirements or at least algorithms and justifications for RIVLIS.
NABU-DP-00014-GEN, rev 2 Design Process for Common Q Safety Systems. As it defines the scope of other documents we are reviewing, it may clarify what documents are expected to contain what information.
TVA Response:
The documents are available for review in the WEC Rockville office per WEC letter WBT-D-2268, NRC Access to Common Q Documents at the Westinghouse Rockville Office, dated August 16, 2010 (Reference 2).
- 6.
NRC Request (Item Number 255)
Please make the following available in Westinghouse's Rockville office.
The Reusable Software Elements Documents. These contain requirements for the software.
WNA-DS-01564-GEN, Rev 1; WNA-DS-00315-GEN, Rev. 2; WNA-DS-01715-GEN, Rev 2; WNA-DS-01838-GEN, Rev. 3; WNA-DS-01839-GEN, Rev. 3; WNA-DS-01840-GEN, Rev 2; WNA-DS-01841, Rev 2; WNA-DS-01842-GEN Rev 2; WNA-DS-01845-GEN Rev.
1; WNA-DS-01846-GEN Rev. 2; WNA-DS-01847-GEN Rev. 0; WNA-DS-01848 Rev. 1; WNA-DS-01849-GEN Rev. 2; WNA-DS-01994-GEN Rev. 0; WNA-DS-00306-GEN Rev. 5; WNA-DS-02065-GEN Rev. 2; WNA-DS-01505-GEN Rev. 0 Further documentation for application-specific type circuits and custom PC elements are indicated by the SRS to be in 00000-ICE-3238, Rev 5; 00000-ICE-30140, Rev 4 and 00000-ICE-30152, Rev. 5.
TVA Response:
The documents are available for review in the WEC Rockville office per WEC letter WBT-D-2268, NRC Access to Common Q Documents at the Westinghouse Rockville Office, dated August 16, 2010 (Reference 2).
- 7.
NRC Request (Item Number 256)
Please make the following available in Westinghouse's Rockville office.
The following are documents that contain requirements used in the SRS which we incorporated by reference within that document.
Coding Standards and Guidelines for Common Q Systems, 00000-ICE-3889, Rev. 10, Westinghouse Electric Company LLC.
Application Restrictions for Generic Common Q Qualification, WNA-DS-01070-GEN, Rev. 3, Westinghouse Electric Company LLC.
System Requirements Specification for the Common Q Generic Flat Panel Display 00000-ICE-30155, Rev. 9, Westinghouse Electric Company LLC.
E1-3 TVA Letter Dated October 21, 2010 Responses to Licensee Open Items to be Resolved for SER Approval Software Requirements Specification for the Common Q Generic Flat Panel Display Software, 00000-ICE-3239, Rev. 12, Westinghouse Electric Company LLC.
Common Q Software Configuration Management Guidelines, NABU-DP-00015-GEN, Rev. 2, Westinghouse Electric Company LLC, [sic]
Standard General Requirements for Cyber Security, WNA-DS-01150-GEN, Rev. 0, Westinghouse Electric Company LLC, [sic]
TVA Response:
The documents are available for review in the WEC Rockville office per WEC letter WBT-D-2268, NRC Access to Common Q Documents at the Westinghouse Rockville Office, dated August 16, 2010 (Reference 2).
- 8.
NRC Request (Item Number 259)
Please make the following available in Westinghouse's Rockville office.
As they may demonstrate that a number of issues raised by, or that will be raised by, the NRC staff are already being resolved by the vendor, we would like to have access to V&V-769 and V&V-770 in the Exception Reports (ER) database for common Q systems.
TVA Response:
The documents are available for review in the WEC Rockville office per WEC letter WBT-D-2268, NRC Access to Common Q Documents at the Westinghouse Rockville Office, dated August 16, 2010 (Reference 2).
9.
NRC Request (Item Number 266)
Please provide a high level description of the Foxboro IA equipment used at WBN2. This description should be more detailed than a brochure on the product line (or available on the web), and less detailed than a technical manual on each field replaceable unit. It is expected that such literature already exists.
TVA Response:
FSAR Section 7.7.1.11 will be added in Amendment 101. In discussions with the NRC reviewer on October 4, 2010, it was agreed that the new FSAR section along with previously submitted documents should be sufficient to address this request. contains the draft FSAR Section 7.7.1.11.
- 10. NRC Request (Item Number 272)
In WBN2 FSAR Table 7.5-2, "Regulatory Guide 1.97 Variable List (Deviation and Justification for Deviations)," (WBNP-96) for Variable 19, "Containment Hydrogen Concentration," Deviation 2 (page 19 of 41), the variable number is listed as 15. The variable number should be listed as 19.
E1-4 TVA Letter Dated October 21, 2010 Responses to Licensee Open Items to be Resolved for SER Approval TVA Response:
The variable number will be changed to 19 in FSAR Amendment 101 as shown in draft below:
Table 7.5-2 Regulatory Guide 1.97 Variable List (Deviation and Justification for Deviations)
(Page 19 of 41)
DEVIATION 2 VARIABLE (1519)
Containment Hydrogen Concentration
- 11. NRC Request (Item Number 274 a.)
In WBN2 FSAR Table 7.5-2, "Regulatory Guide 1.97 Variable List (Deviation and Justification for Deviations)," (WBNP-96) for Variable 82, "Steam Generator Level Wide Range," Deviation 10 (page 24 of 41), in the last sentence, of the Justification, SC should be SG.
TVA Response:
The SC in the last sentence will be changed to SG in FSAR Amendment 101 as shown in draft below:
Table 7.5-2 Regulatory Guide 1.97 Variable List (Deviation and Justification for Deviations)
(Page 19 of 41)
DEVIATION 10 VARIABLE (82)
SG Level Wide Range DEVIATION FROM RG 1.97 GUIDANCE RG 1.97, Revision 2, recommends this variable as a Type D, Category 1 variable, which requires redundancy in the instrumentation. WBN recommends this variable be Category 1, Type D, but utilizing only one wide range transmitter per SG.
JUSTIFICATION SG wide range level indication is utilized as a diverse variable to auxiliary feedwater (AFW) flow for gross indication of flow to the SGs. The WBN AFW monitors are Types A1 and D2. WBN's position is that since SC SG wide range level is only used as a backup to redundant AFW flow monitors, it does not require redundancy.
E1-5 TVA Letter Dated October 21, 2010 Responses to Licensee Open Items to be Resolved for SER Approval
- 12. NRC Request (Item Number 274 b.)
Loose Parts Monitoring System: TR 3.3 refers to section 4.4.6 of the FSAR for description of the loose parts monitoring system. However, this section of the FSAR is not available.
TVA to check the reference and respond.
TVA Response:
In the next revision to the Technical Requirements Manual (TRM), the References section in the TRM Bases, Section 3.3.6 will be changed in accordance with the following draft:
1.
Watts Bar FSAR, Section 7.6.7, Loose Part Monitoring System (LPMS) System Description
- 13. NRC Request (Item Number 276)
In order for the staff to review the effects of multi control systems failure, provide the summary of the analyses documenting the effect on the plant based on the following events: (1) loss of power to all control systems powered by a single power supply; (2) failure of each instrument sensor which provides signal to two or more control systems; (3) break of any sensor impulse line which is used for sensors providing signals to two or more control systems; and (4) failure of digital system based on the common cause software failure affecting two or more control systems. For each of these events, confirm that the consequences of these events will not be outside chapter 15 analyses or beyond the capability of operators or safety systems.
TVA Response:
The NRC reviewer confirmed this question applies to non-safety systems.
The Distributed Control System (DCS) implemented using Foxboro I/A hardware, replaces most of the non-safety related control systems for WBN Unit 2. The other non-safety-related control systems are:
a) Rod Control - Failures of this system are addressed in FSAR Chapter 15.
b) Main Turbine Electro-Hydraulic Control System The following provides the requested summaries for the four events listed:
(1)
The (DCS) segmentation analysis submitted on TVA letter to NRC dated August 11, 2010, Enclosure 2 (Reference 7) demonstrates that the loss of any single power source does not result in a loss of any DCS function. The other systems within the scope of this question are configured in the same manner as Unit 1, with redundant power sources such that the failure of a single power source does not cause a loss of function.
E1-6 TVA Letter Dated October 21, 2010 Responses to Licensee Open Items to be Resolved for SER Approval (2)
Signals shared by more than one control function within the DCS are addressed in the DCS segmentation analysis submitted on TVA letter to NRC dated August 11, 2010, Enclosure 2 (Reference 7) which demonstrates that the loss of a single signal does not cause a failure of any critical control function. The impact of a loss of signal to other systems is bounded by the loss of that signal to the individual system and has the same effect as for Unit 1.
(3)
Where feasible, the Unit 2 design includes separate sense lines for redundant transmitters, thereby eliminating multiple single point failures which are present in Unit 1. A review of the transmitter sense line database was performed to identify multiple sensors on a single sense line that had control functions (transmitters and switches). Attachment 9 provides the results of the review and an analysis of the functions impacted by a sense line failure.
There are no transmitters on shared sense lines, such that a sense line failure would impact any combination of the DCS, Rod Control or Main Turbine Electro-Hydraulic Control Systems.
(4)
Limiting DCS failures were addressed in the segmentation analysis, supplemented by Fault Handling in the I/A Series System, Revision 1, submitted on TVA letter to NRC dated October 5, 2010, Attachment 42 (Reference 1). The other systems within the scope of this question are analog and therefore this question is not applicable.
- 14. NRC Request (Item Number 278)
For FSAR Section 7.6.6, provide the justification for adding valves FCV 63-8 and FCV 63-11, which require that power be removed and will be administratively controlled prior to use of RHR system for plant cooldown. Provide the P & ID and block diagram showing the operation of these valves.
TVA Response:
The Unit 1 General operating instructions GO-6 removes power prior to placing RHR in service for plant cooldown below 350°F to prevent inadvertent valve opening and over pressurization of the Safety Injection Pump (SIP) and Centrifugal Charging Pump (CCP) suction piping. The Unit 1 General operating instructions GO-1 restores power after RHR is removed from service to allow normal valve operation. The Unit 1 procedures will be used to develop the Unit 2 operating instructions to provide the administrative instructions to remove and restore power to these valves. contains control valves FCV-63-8 and -11 control and logic diagrams, (2-47w610-63-1, 2-47w610-63-2 and 2-47w610-63-5) as well as the applicable design changes (2-47w611-63-5 DRA 54670-005 and 2-47w611-63-5 DRA 54670-009) to verify that the control schemes are similar to Unit 1.
E1-7 TVA Letter Dated October 21, 2010 Responses to Licensee Open Items to be Resolved for SER Approval
- 15. NRC Request (Item Number 279)
For FSAR Section 7.6.6, provide the justification for the exception to install protective covers which operator has to remove before he can have access to control switch to operate two additional valves FCV-62-98 and FCV-62-99.
TVA Response:
The FSAR change to include the valves as exceptions to the use of protective covers was made to match Unit 1 UFSAR change Pkg. No. 1547 Safety Assessment Item 8. The exception is justified based on WBPER980417 which removed the power from the valves and had them locked open. TVA will incorporate the same changes in Unit 2 as Unit 1.
The Unit 1 changes are described in References 3 and 4.
- 16. NRC Request (Item Number 280)
For FSAR Section 7.6.6, provide the justification for the acceptability of removing FCV 63-5 from the list of valves which has operating instructions specifying the removal of power during specific modes of plant operation.
TVA Response:
Historical DCN 38661 removes the requirement that power be removed from FCV-63-5 during normal operations, and notes that the valve does not have a shunt breaker to allow MCR position indication with power removed. The Unit 2 system description has been updated to reflect the Unit 1 change to the system description, and the update of section 7.6.6 to remove the requirement of FCV-63-5 from the list of valves which has operating instructions specifying the removal of power during normal operations. This is supported by the failure modes and effects analysis for the safety injection system calculation EMP-EPMSNM043092 (which has been revised to be applicable to Unit 2), as well as the Unit 2 FSAR Table 6.3-8 both of which state that spurious closure of FCV-63-5 is not credible.
Spurious closure of FCV-63-5 is not credible because the MCR hand switch is provided with a protective cover to prevent operator error. In addition, the hand switch is wired with contacts on both sides of the motor contactor to prevent a single failure within the switch gear from spuriously closing the valve. These features eliminate the need to remove power from FCV-63-5. contains excerpts from system description N2-63-4001, Safety Injection Revision 1, DCN 3 and calculation EPMSNM043092, Failure Modes and Effects Analysis for the Safety Injection System, Revision 7, associated with this response.
- 17. NRC Request (Item Number 282)
For FSAR Section 7.6.9 which discusses the switch over from injection to recirculation, and is a ESF system, the compliance with IEEE 279 has been removed from the FSAR.
Justify this deletion.
E1-8 TVA Letter Dated October 21, 2010 Responses to Licensee Open Items to be Resolved for SER Approval TVA Response:
The re-write for Section 7.6.9 was to provide a more concise description of the instrumentation and controls. The section was too wordy, and several topics were duplicated in Section 7.3. Wording is now more closely aligned to system description.
Compliance with IEEE 279 is not intended to be eliminated, merely the reference to the standard in that particular section. The following statement is added: The automatic switchover of the RHR pumps from the injection to the recirculation Mode is part of the Engineered Safety Features Actuation System (ESFAS) discussed in chapter 7.3.
Chapter 7.3 includes a reference to IEEE Standard 279-1979. The reference in 7.6.9 was therefore considered unnecessary, and therefore removed. contains FSAR excerpts that support this response.
- 18. NRC Request (Item Number 284)
Follow-up to item 123 Please provide a readable electrical logic diagram of the Volume Control Tank Level Control System.
TVA Response: contains the Volume Control Tank Level Control System electrical logic diagrams (2-47w611-62-3, 2-47w611-62-4, vendor drawings (08F802403-FD-2603 sheets1 through 4 and 08F802403-FD-2604 sheet 1), and required Drawing Change Authorizations (DRAs) (DRA 52378-615 and DRA 52671-091).
- 19. NRC Request (Item Number 286)
SE 7.7.3, Volume Control Tank Level Control System In FSAR section 9.3.4.2.4 a change was made to the last paragraph of the Volume Control Tank description (page 9.3-31 of the Amendment 97 redline), where the "low-low level alarm" was changed to "low level alarm".
Please explain if this deletion was an editorial change to correct a typo.
TVA Response:
Low level alarm is correct - the setpoint is above the low-low interlock that opens the isolation valve, mentioned earlier in the paragraph. This is an editorial change to correct a typo.
- 20. NRC Request (Item Number 289)
Provide an ISG2 diversity analysis for the containment high range accident monitors RM-1000.
E1-9 TVA Letter Dated October 21, 2010 Responses to Licensee Open Items to be Resolved for SER Approval TVA Response:
There are four Containment High Range Radiation Monitors (HRRMs) for WBN Unit 2, a pair in upper containment and a pair in lower containment. Each pair completely meets the requirements for safety-related equipment including separation, independence, electrical isolation, seismic qualification, quality requirements, etc. Each monitor channel is a standalone instrument loop with traditional individual panel readout. They are not a part of a Highly Integrated Control Room (HICR), and there is no diversity question relating to the HRRMs and any HICR infrastructure. Therefore, the response to this RAI will address the functional uses of the HRRMs and the alternate and diverse instrumentation that could be used for those functions should a common mode software issue render both trains of HRRMs non-functional.
The Containment HRRMs have no automatic actuation function and only provide indication as required by RG 1.97, R2. These devices serve two functions at WBN.
Emergency Operating Instructions (EOI) dictate use as one of the indications of abnormal containment conditions indicative of a Loss of Coolant Accident (LOCA) after a Reactor Trip and Safety Injection. Additionally, Emergency Plan Implementing Procedures (EPIP) use these devices to assist with event classification for events which involve fuel cladding degradation.
In the EOI procedures, there are several diverse indications of containment conditions that are used to detect a LOCA. These indications are: Containment Pressure, Containment Temperature, and Containment Sump Level. These instrument channels are diverse to the HRRMs in that they do not share a software platform or any integrated information or control system features. The HRRMs are functional through individual, self contained, microprocessor based instrument loops. Containment Pressure and Sump Level indications are provided through Eagle 21 equipment which is completely diverse from the HRRMs. Containment Temperature is provided through Foxboro Spec 200 instrument channels which are completely diverse from the HRRMs. These readouts are through traditional panel meters and are not part of any HICR infrastructure.
In the EPIPs, the HRRMs are used to indicate loss of fuel clad barrier and the potential loss of a containment barrier. Potential fuel clad damage can also be determined from samples taken from the RCS and from Incore Thermocouple readings. RCS sampling does not rely on plant instrumentation systems, and the Incore Thermocouple System uses a Common Q software platform which is diverse from the HRRMs. The accessibility required to obtain post accident samples of RCS has been demonstrated to be a viable post accident action at WBN.
In the event that these four channels of HRRMs fail upscale, Annunciator Response Instructions would be followed which require evacuation of containment, sampling of RCS, checking other non-accident Radiation Monitors, notification of Radiological Control personnel to investigate, potential transition to Abnormal Operating procedures for management of potential radioactive material release, and evaluation under the EPIPs for event classification. These actions are conservative actions. In the event that these four channels of the HRRMs fail downscale, the operators would turn to diverse indications as noted above before taking any further action.
E1-10 TVA Letter Dated October 21, 2010 Responses to Licensee Open Items to be Resolved for SER Approval Therefore, there are diverse methods and equipment sets that can be used for any functions provided by the HRRMs in the event that both channels become nonfunctional.
- 21. NRC Request (Item Number 291)
The equation at the bottom of Amendment 99 page 7.7-3 is wrong. There are two ways that this equation is inconsistent with the text above it.
TVA Response:
The errors in the terms within the equation for total rod speed error [TE] will be corrected in FSAR Amendment 1 as shown in w:
10 draft belo
- 22. NRC Request (Item Number 292)
FSAR Section 7.2, Steam Generator Reference Leg: By letter dated July 27, 1994, TVA had withdrawn its commitment on Unit 1 to insulate SG reference leg. TVA had provided an analysis to justify this action which was accepted by the staff. Confirm whether SG reference leg in Unit 2 are insulated and if not then confirm that the analysis which was submitted for Unit 1 is also applicable to Unit 2.
TVA Response:
The Steam Generator (SG) level transmitter reference legs are not insulated on Unit 1 and will not be insulated on Unit 2. The analysis provided for Unit 1 is also applicable to Unit 2.
FSAR Section 7.2.1.1.2 (5) indicates that the Low-Low SG water level trip protects the reactor from loss of heat sink in the event of a loss of feedwater to one or more SGs or a major feedwater line rupture outside containment. For a feedwater line rupture inside containment, the TVA analysis credits the high containment pressure Safety Injection signal. FSAR Section 15.4.2.2 has been revised accordingly. The basis for not insulating the SG reference legs and the Unit 1 NRC licensing review are contained in References 5 and 6.
- 23. NRC Request (Item Number 312)
By letter dated September 10, 2010, TVA provided the summary evaluation of 50.59 reports which were related to FSAR Chapter 7.0. However, these evaluation only covers Amendments 0 thru 8. Provide all other evaluations which have been done since these amendments and which forms the basis for FSAR Chapter 7.0 systems.
TVA Response:
Amendment 8 is the current version of Unit 1 UFSAR, and review of the Unit 1 UFSAR change packages performed to date did not identify any additional changes to Chapter 7 requiring a 10 CFR 50.59 evaluation.
E1-11 TVA Letter Dated October 21, 2010 Responses to Licensee Open Items to be Resolved for SER Approval
- 24. NRC Request (Item Number 314)
The following 50.59 changes were listed in the March 12 RAI response letter (item 10) but were not included in the September 9 submittal of 50.59 safety evaluations. Please submit the 50.59 safety evaluations for the following changes:
x DCN 38842 (Revise OTT and OPT turbine runback setpoints) x DCN 50991 (Install Test Points) x DCN 51124 (Eliminate spurious ICS alarms associated with the SSPS [sic]
TVA Response: contains the requested safety evaluations.
- 25. NRC Request (Item Number 315)
IE Bulletin 79-27 required that emergency operating procedures to be used by control room operators to attain safe shutdown upon loss of any Class IE or non Class IE bus are adequate. WBN1 has performed the review and documented their conclusion. Confirm that WBN2 emergency procedures are adequate to achieve safe shutdown in the event of loss of any Class IE or non-Class IE bus.
TVA Response:
While the WBN Unit 2 Emergency Operating Procedures (EOPs) have not been written, they will be written the same as the Unit 1 EOPs. WBN Unit 1 personnel will perform validations to ensure that WBN Unit 2 EOPs will perform the required actions. The WBN Unit 2 EOPs will be written and validated prior to Unit 2 fuel load.
- 26. NRC Request (Item Number 316)
TVA has provided various documents in support of RM-1000 high range monitors for WBN2.
Please clarify the following:
RM-1000 v1.1 Software Verification Report 04508006 (Sequoyah)
RM-1000 v1.2 Software Verification Report 04508006 (Sequoyah)
RM-1000 System Verification Test Results (Sequoyah)
These documents were prepared for the Sequoyah plant. Is the version provided applicable to WBN2? Please confirm and explain if these documents are applicable to WBN 2 as provided or with differences?
E1-12 TVA Letter Dated October 21, 2010 Responses to Licensee Open Items to be Resolved for SER Approval E1-13 TVA Response:
The Sequoyah RM-1000 v1.1 Software Verification Report 04508006 and RM-1000 v1.2 Software Verification Report 04508006 are applicable to WBN Unit 2.
The RM-1000 System Verification Test Results report is not applicable to WBN Unit 2.
This document was for the non-safety-related software and was superseded by the 04508006 v1.1 and v1.2 reports for the safety-related software.
TVA Letter Dated October 21, 2010 Responses to Licensee Open Items to be Resolved for SER Approval Attachments
- 1.
WBPER233598 and associated engineering evaluation
- 2.
Unit 1 UFSAR Amendment 3 Change Package 1676 S00
- 3.
Draft FSAR Section 7.7.1.11, Distributed Control System
- 4.
Control valves FCV-63-8 and -11 control and logic diagrams, (2-47w610-63-1, 2-47w610-63-2 and 2-47w610-63-5) as well as the applicable design changes (2-47w611-63-5 DRA 54670-005 and 2-47w611-63-5 DRA 54670-009) to verify that the control schemes are similar to Unit 1.
- 5.
Excerpts from system description N2-63-4001, Safety Injection Rev. 1, DCN 3 and calculation EPMSNM043092, Failure Modes and Effects Analysis for the Safety Injection System, Rev. 7
- 6.
FSAR Section 7.3 excerpts
- 7.
Volume Control Tank Level Control System electrical logic diagrams (2-47w611-62-3, 2-47w611-62-4, vendor drawings (08F802403-FD-2603 sheets1 through 4 and 08F802403-FD-2604 sheet 1), and required Drawing Change Authorizations (DRAs) (DRA 52378-615 and DRA 52671-091).
- 8.
Safety Evaluations for DCN 38842 (Revise OTT and OPT turbine runback setpoints),
DCN 50991 (Install Test Points) and DCN 51124 (Eliminate spurious ICS alarms associated with the SSPS)
- 9.
Review and an analysis of the functions impacted by a sense line failure for multiple non-safety related sensors on a single sense line.
E2-1 TVA Letter Dated October 21, 2010 Responses to Licensee Open Items to be Resolved for SER Approval
References:
- 1.
TVA Letter to NRC Watts Bar Nuclear Plant (WBN) Unit 2 - Instrumentation And Controls Staff Information Requests, Dated October 5, 2010
- 2.
Westinghouse letter WBT-D-2268, NRC Access to Common Q Documents at the Westinghouse Rockville Office, Dated August 16, 2010
- 3.
System Description N3-62-4001, Chemical And Volume Control System, Rev. 31
- 4.
Drawing 1-45W760-62-3 Wiring Diagrams Chemical And Volume Control Sys Schematic Diagrams, Rev. 9, Note 4
- 5.
Watts Bar Unit 1 SER NUREG-0847, Supplement 14
- 6.
Summary Report Process Protection System Eagle 21 Upgrade, NSLB, MSS, and TTD Implementation, Watts Bar Unit 1 and 2, WCAP-13462, Revision 2, September 1994
- 7.
TVA Letter to NRC Dated August 11, 2010, Watts Bar Nuclear Plant (WBN) Unit 2, NRC August 3-4, 2010, Instrumentation and Controls Meeting, Request for Additional Information E3-1 TVA Letter Dated October 21, 2010 Responses to Licensee Open Items to be Resolved for SER Approval Commitments:
- 1. The external Eagle 21 unidirectional communications interface will be tested prior to WBN Unit 2 fuel load. [Letter Item 3]
- 3. The variable number in the WBN Unit 2 FSAR, Table 7.5-2 for Containment Hydrogen Concentration will be corrected from 15 to 19 in FSAR Amendment 101. [Letter Item 10]
- 4. In WBN2 FSAR Table 7.5-2, "Regulatory Guide 1.97 Variable List (Deviation and Justification for Deviations)," (WBNP-96) for Variable 82, "Steam Generator Level Wide Range," Deviation 10 (page 24 of 41), in the last sentence, of the Justification, SC will be corrected to SG in FSAR Amendment 101. [Letter Item 11]
- 5. In the next revision to the Technical Requirements Manual (TRM), the References section in the TRM Bases, Section 3.3.6 will be changed to read as follows:
- 1. Loose Part Monitoring System (LPMS) System Description [Letter Item 12]
- 6. The equation that determines the total error signal (TE) in the rod speed program in Section 7.7.1.1.2 of the Unit 2 FSAR will be corrected in FSAR Amendment 101 to be consistent with the text within this FSAR section. [Letter Item 21]
- 7. While the WBN Unit 2 Emergency Operating Procedures (EOPs) have not been written, they will be written the same as the Unit 1 EOPs. WBN Unit 1 personnel will perform validations to ensure that WBN Unit 2 EOPs will perform the required actions. [Letter Item 25]
- 9. The FSAR change to include the valves as exceptions to the use of protective covers was made to match Unit 1 UFSAR change Pkg. No. 1547 Safety Assessment Item 8.
The change package identified FCV-62-98 and 99 as exceptions to the use of protective covers. This change was based on WBPER980417 which removed the power from the valves and had them locked open. TVA will incorporate the same changes in Unit 2 as Unit 1. [Letter Item 0]
E4-1