ML20237E814
| ML20237E814 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood  |
| Issue date: | 08/27/1998 |
| From: | Krich R COMMONWEALTH EDISON CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| GL-96-01, GL-96-06, GL-96-1, GL-96-6, NUDOCS 9809010196 | |
| Download: ML20237E814 (33) | |
Text
,
Comnwnwcalth FAison Company 1400 Opus Place ihmners Grove, IL e0515-5701 August 27,1998 United States Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 1
Subject:
Byron Station, Units 1 and 2 Facility Operating License Nos. NPF-37 and NPF-66 NRC Docket Nos. STN 50-455 andJTN 50-456 Braidwood Station, Units 1 and 2 Facility Operating License Nos. NPF-72 and NPF-77 NRC Docket Nos. STN 50-456 and STN 50-457 Response to Request for Additional Information Generic Letter 96-06, " Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions"
References:
1.
Letter from J. B. Hickman (NRC) to O. D. Kingsley (Comed), " Request for Additional Information Related to the Generic Letter (GL) 96-06 Response for Byron Station, Units I and 2 and Braidwood Station, Units 1 I
and 2," dated May 1,1998.
2.
Letter from J. Hosmer (Comed) to US Nuclear Regulatory Commission 1
(NRC), "The Commonwealth Edison (Comed) Company Response to NRC Generic Letter 96-06," dated May 2,1997.
p j
l By letter dated May 1,1998 (Reference 1), the NRC requested that the Commonwealth Edison Company (Comed) provide additional information needed to complete the NRC's review of our response to NRC Generic Letter (GL) 96-06, " Assurance of Equipment
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Operability and Containment Integrity During Design-Basis Accident Conditions," as applicable to the Byron and Braidwood Stations. There was no specified due date for submitting this additional information.
The responses to the request for additional information are provided in the attachment to this letter. Additionally, Comed is using this response to update the NRC on the current status of containment penetration activities described in Reference 2. The Byron and Braidwood Stations will supplement this i sponse by February 26,1999.
9909010196 9808 I[7 PDR ADOCK 05000454 P
PDR l
A Unicom Company
]
l USNRC August 27,1998 l
l Please direct any comment or questions regarding this matter to Ms. Marcia Lesniak at 630-663-6484.
Respectively, s
f R.M. Knch
. Vice President - Regulatory Services
\\
Attachment Regional Administrator -NRC Region ill cc:
NRC Senior Resident inspector - Braidwood Station NRC Senior Resident Inspector - Byron Station i
Attachment Response to Request for Additional Information Regarding NRC Generic Letter 96-06, " Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Condition," for the Byron and Braidwood Stations.
- 1. In its submittal of May 2,1997, the licensee indicated that 11 pipe segments in each unit would require the installation of spring loaded relief valves or check valves or modifications to containment isolation valves. Please provide a specific description of the design changes for each of these 11 pipe segments in each unit.
In the Comed submittal of May 2,1997 the following containment penetrations were identified as requiring design changes to mitigate thermally induced overpressure conditions.
P-5, P-6, P-8, and P Containment Chilled Water System E
P Reactor Coolant Drain Tank Pump Discharge P Co.nponent Cooling Water Return from Reactor Coolant Pump (Main Steam Line Break (MSLB) Recovery Only)
P Containment Fire Protection Supply P Reactor CoAnt System I oop Fill lleader P Primary Water Supply to Reactor Coolant Pump Seal #3 and Pressurizer Relief Tank P Containment Floor Drain Sump Pump Discharge P Safety injection Accumulator Fill Line Penetrations P-5, P-8, P-37 and P-55 Based upon Supplement I to GL 96-06 and the results of various ongoing initiatives, Comed has elected to defer modification activities for penetrations P-5, P-8, P-37 and P-55 while an analytical solution employing ASME Section 111 Appendix F is being pursued. The analytical evaluations of these four penetrations are scheduled to be completed by the end of 1998.
Penetration P-24 This penetration has been evaluated as having adequate overpressure protection based on the configuration of the installed piping system.
The modifications planned were to ensure that valve _CC685 (the containment isolation valve outside containment for P-24) would be capable of opening against the maximum possible pressure of 2485 psig. It has since been determined that valve _CC685 is capable of opening against the worst-case pressure condition. Therefore, modification of the piping associated with containmen: penetration P-24 is neither desired nor warranted.
I
Attachment Response to Request for Additional Information Regarding NRC Generic Letter 96-06, " Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Condition," for the Byron and Braidwood Stations.
A description of the design changes for the remaining six containment penetrations where modifications are presently planned are listed below. Please note there are no differences in the design changes for the four Byron and Braidwood units (Unit 1 component numbers are used; Unit 2 component numbers are similar):
Penetrations P-6 and P-10 Relief valves are being installed inside containment prior to the inner containment isolation check valves 1WOOO7A (penetration P-6) and IWOOO7B (penetration P-10). This configuration will allow pressure relief through these check valves into containment and still maintain the valves' containment isolation function. The relief valves to be installed are 1" x 1 %" size. Drawings are located in the Updated Final Safety Analysis Report (UFSAR)
Figure 9.2-16 sheet 2.
For Braidwood, these modifications are currently scheduled to be installed during refueling outages A1R07 (Fall 1998) and A2R07 (late Spring 1999). For Byron, these modifications are currently scheduled to be installed in B1R09 (Spring 1999) and B2R08 (Fall 1999).
Penetration P-11 A relief valve is being installed inside containment between containment isolation valves 1RE1003 and 1RE9170. This configuration will allow for thermal pressure relief of t! e
!solated piping section through the relief valve. The relief valve to be installed is %" x 1" size.
For Braidwood, these modifications are currently scheduled to be installed during A2R07 and AlR08 (Spring 2000). Please note the Unit I scheduled installation date has been revised from the Reference 2 response. For Byron, these modifications are currently scheduled to be installed during BIR09 and B2R08.
Penetration P-34 A relief valve is being installed inside containment prior to the inner containment isolation check valve 1FP345. This configuration will allow pressure relief through the check valve into containment and still maintain the valve's containment isolation function. The relief valve to be installed is %" x 1" size. Drawings are located in the Byron /Braidwood Stations Fire Protection Report, Appendix 5.0 Figure M52 sheet 1.
For Braidwood, these modifications are currently scheduled to be installed during A1R07 and A2R07. For Byron, these modifications are currently scheduled to be installed during B1R09 and B2R08.
2
Attachment Response to Request for Additional Information Regarding NRC Generic Letter 96-06," Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Condition," for the Byron and Braidwood Stations.
Penetration P-44 A %" by-pass line with a spring-loaded check valve will be installed inside containment around diaphragm valve 1RY8030 which is downstream of containment isolation check valve 1RY8046. This configuration will allow pressure relief through this check valve into containment (to the Pressurizer Relief Tank) and still maintain the check valves containment isolation function. Drawings are located in the UFSAR Figure 5.1-1 sheet 8.
For Braidwood, these modifications are scheduled to be installed during AlR07 and A2R07.
For Byron, these modifications are scheduled to be installed during B1R09 and B2R08.
Penetration P-47 A relief valve is being installed inside containment between containment isolation valves 1RF026 and 1RF027. This configuration will allow for thermal pressure relief of the isolated containment piping section through the relief valve. The relief valve to be installed is a %" x 1" size. Drawings are located in the UFSAR Figure 11.2-7 sheets 1 and 2.
For Braidwood these modifications are scheduled to be installed during A2R07 and AlR08.
Please note the Unit I scheduled installation date has been revised from the Reference 2 response. For Byron, these modifications are scheduled to be installed during B1R09 and B2R08.
3
Attachment Response to Request for Additional Information Regarding NRC Generic Letter 96-06, " Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Condition," for the Byron and Braidwood Stations.
- 2. In its submittal of May 2,1997, the licensee indicated that overpressurization of one piping segment in each unit was intended to be mitigated by procedural changes and that overpressurization of two piping segments in each unit was intended to be mitigated by draining. Please verify the specific method of mitigation to be implemented and describe the specific procedural changes involved.
In the Comed submittal of May 2,1997, the following containment penetrations were identified as having thermally induced overpressurization was mitigated by procedural changes or draining.
P-30 Containment Demineralized Water Supply P-32 Fuel Pool Cooling Return to Refueling Cavity e
P-57 Fuel Pool Cooling Suction from Refueling Cavity Details on the mitigation methodologies are provided below. Unit I component numbers are listed. Unit 2 component numbers are similar.
Penetration P-30 For Braidwood Station, valve 1 WM192A will be left open during normal operation. This e
valve is located inside containment and provides a path from the Demineralized Water (WM) system (which is isolated to the containment during normal operation) to the containment atmosphere. This configuration will allow pressure relief through containment isolation check valve 1 WM191 into containment and still maintain the check valve's containment isolation function. The Braidwood Station operating mechanical lineup procedure for the WM system has been revised to require valve 1WM192A to be left in the open position
]
during normal power operation. This methodology has been employed at Braidwood Station l
since the return-to-service from AIR 06 (Spring 1997) and A2R06 (Fall 1997).
Please note that Byron Station has chosen to procedurally drain the WM penetrations. The Unit 1 penetration was drained during B1R08 (February 1998), and the Unit 2 penetration was drained during B2R07 (May 1998) using Byron special process procedures. Byron Station operating procedures are currently used to perform the draining evolution. Draining activities for these penetrations have been implemented since the return-to-service from B1R08 and B2R07.
Note that although the Braidwood and Byron methodologies are different, the end result in both cases is acceptable because penetration P-30 is precluded from thermal overpressurization.
4 L____.__..__._..
J
Att:chment Response to Request for Additional Information Regarding NRC Generic Letter 96-06, " Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Condition," for the Byron and Braidwood Stations.
Penetrations P-32 and P-57 These penetrations are used only during refueling activities when the refueling cavity in containment is filled. Thermal overpressure mitigation for these penetrations is accomplished by draining the penetrations after refueling activities are complete and the containment refueling cavity has been drained.
At Braidwood Station, procedural controls have been implemented to ensure draining of penetrations P-32 and P-57 after use during outage periods prior to returning the plant to power operations. Draining activities for these penetrations have been implemented at Braidwood Station since the return-to-service from AIR 06 and A2R06.
At Byron Station, Unit 1 penetrations were drained during the BIR08 containment integrated leak rate test and the containment isolation valves have been administratively taken out-of-service. Procedural controls will be in place for BIR09 draining activities. A special process procedure was performed during B2R07 to drain the Unit 2 penetrations and the Unit 2 containment isolation valves were taken out-of-service. Permanent procedures for Unit 2 are being developed and are scheduled for completion by the end of 1998. Draining activities for these penetrations have been implemented since the return-to-service from B1R08 and B2R07.
I i
5
l i
Attrehment l
Response to Request for Additional Information Regarding NRC Generic Letter 96-06, " Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Condition," for the Byron and Braidwood Stations.
l I
l t
j
- 3. In the submittal of May 2,1997, the licensee indicated that two piping segments in each l
unit were being analyzed in detail to determine if there is an actual overpressure condition.
l Please verify if an overpressure condition exists and,if so, provide the method of mitigation of the overpressure. If the licensee is crediting the structural adequacy of these segments to withstand the overpressure condition, the following information for these piping segments should be provided:
l Provide the applicable design criteria for the piping and the valves. Include tne required load combinations.
Provide a drawing of the piping run between the isolation valves. Include the lengths and thickness' of the piping segments and the type and thickness of the insulation.
Provide the maximum-calculated temperature and pressure for the pipe run. Describe, in detail, the method used to calculate these pressure and temperature values. This should include a discussion of the heat transfer model used in the analysis and the basis for the heat transfer coefficients used in the analysis.
In the Comed submittal of May 2,1997, two containment penetrations were identified as warranting a more detailed analysis to determine if an overpressure concern actually exists:
P Process Sampling System (PS) (Containment isolation valves - normally closed during plant operation), and P Chemical and Volume Control System (CV) Normal Charging Flow Path (containment isolation valves - closes on a containment isolation signal).
These penetrations are being analyzed in detail (beyond the original screening process that conservatively enveloped many conditions and circumstances) to determine if there is an actual overpressure condition and the magnitude of such a condition, ifit exists. Based upon Supplement I to GL 96-06 and the results of various on-going industry initiatives, the analytical review of these two penetrations may also consider the use of ASME Section III Appendix F criteria and/or the use of qualified insulation to reduce the expected peak temperature. Comed will provide this information to the NRC when the analysis is complete. This analysis is expected to be completed by early 1999. Byron and Braidwood Stations expect a supplement to l
this response to be submitted by February 26,1999. At that time the applicable construction l
drawings for Braidwood Unit 2 and Byron Units 1 and 2 will also be provided.
6
l Attrchment Response to Request for Additional Information Regarding NRC Generic Letter 96-06, " Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Condition," for the Byron and Braidwood Stations.
The following provides the available information for each penetration. Braidwood Unit 1 information is provided. This information is considered to be typical of the four Braidwood and Byron units. (Note that the other three units have similar configurations):
P Process Sample This penetration contains four process lines:
IPS183A-3/8", IPS184A-3/8", IPS185A-3/8", and IPS186A-3/8" (Note that the Unit I and Unit 2 lines have similar configurations with different line numbers)
Design Criteria:
e These lines are classified as ASME Section 111 Class 2 between (and including) their respective containment isolation valves.
The design pressures and temperatures are as listed below:
1PS183A-3/8" 2485 psig 680 F e
e 1PS184A-3/8" 2485 psig 680 F e
1 PSI 85A-3/8" 700 psig 300 F (SI Accumulator sample line)
IPS186A-3/8" 2485 psig 650 F e
Piping System Drawings and Piping Materials used:
e The following drawings provide the dimensional layout of the piping segments associated with P-70. The materials of construction are also listed on these drawings. All lines are uninsulated.
I.ine Number Applicable As-Built Drawings 1 PS183 A-3/8" PG-2616A-11 PG-2616C-32 1 PS184A-3/8" PG-2616A-14 PG-2616C-22 1 PSI 85A-3/8" PG-2616A-17 PG-2616C-14 1 PS186A-3/8" PG-2616A-20 PG-2616C-34 7
Attochment Response to Request for Additional Information Regarding NRC Generic Letter 96-06, " Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Condition," for the Byron and Braidwood Stations.
Analysis Methodologies and Deta:Is:
Since the analysis activities associated with containment penetration P-70 are not yet complete, the load combinations, maximum temperature, maximum resulting pressure and details of the analytical methods employed are presently not available. Comed will provide this information to the NRC when the analysis is complete. Completion of this analysis is anticipated by early 1999. Byron and Braidwood Stations expect a supplement to this response to be submitted by February 26,1999.
P Chemical and Volume Control (CV) Normal Charcine Flow Path This penetration contains one process line: 1CV09D-3" I
l Additionally the envelope of piping potentially subject to thermal overpressurization includes l
line ICV 09E-3" and valves ICV 8381, ICV 8324A and ICV 8324B.
Design Criteria:
These lines are classified as ASME Section 111 Class 2 between (and beyond) the containment isolation valves. Refer to UFSAR Figure 9.3-4 sheets 5 and 8.
The design temperature and pressure for these lines are listed below:
1CV09D-3" 2845 psig 200 F e
1CV09E-3" 2845 psig 200 F e
Piping System Drawings and Piping Materials Used:
e The following drawings provide the dimensional layout of the piping segments associated with P-71.
l Line Number Applicable As-Built Drawings j
1CV09D-3" 1 A-CV-2
{
l CV09E-3" 1C-CV-41 l
IC-CV-38 i
i l
8 1
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_ ____ __________-_____ ________ a
Attrehment Response to Request for AdditionalInformation Regarding NRC Generic Letter 96-06, " Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Condition," for the Byron and Braidwood Stations.
These lines are uninsulated and are constructed of the following materials:
Pine s
Material:
Sizes 8 in. and smaller....... Seamless austenitic steel, ASME SA-375, Gr. TP304 and/or SA-312, Gr. TP30.
Wall thickness:
Sizes 8 in. and smaller...... Schedule 160 per ANSI B36.10.
Fittines Sizes 2 in. and smaller....... Socket-weld,6000 lb. std. ASME SA-182, Gr. F304.
Sizes 2 % in. through 8 in..... Butt-weld, ASME SA-403, Gr. WP304 or WP316.
Valves Body and Bonnet Material:
Sizes 8 in. and smaller....... Forged or cast alloy steel. ASME SA-182, Gr. F304 or F316 or SA-351, Gr. CF8 or CF8M.
- Analysis Methodologies and Details:
Since the analysis activities associated with containment penetration P-71 are not yet complete, the load combinations, maximum temperature, maximum resulting pressure and details of the analytical methods employed are presently not available. Comed will provide this information to the NRC when the analysis is complete. Completion of this analysis is anticipated by early 1999. Byron and Braidwood Stations will supplement this response by February 26,1999.
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