BSEP 06-0074, Relief Request RR-38, Pressure Testing of Drain, Vent, and Fill Lines within the Reactor Coolant Pressure Boundary: Difference between revisions

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{{#Wiki_filter:I- £JProgress Energy 10 CFR 50.55a(a)(3)(ii)
{{#Wiki_filter:I-  
£ JProgress Energy 10 CFR 50.55a(a)(3)(ii)
JUL 18 2006 SERIAL: BSEP 06-0074 U. S. Nuclear Regulatory Commission ATIN: Document Control Desk Washington, DC 20555-0001
JUL 18 2006 SERIAL: BSEP 06-0074 U. S. Nuclear Regulatory Commission ATIN: Document Control Desk Washington, DC 20555-0001


==Subject:==
==Subject:==
Brunswick Steam Electric Plant, Unit Nos. 1 and 2 Docket Nos. 50-325 and 50-324/License Nos. DPR-71 and DPR-62 Relief Request RR-38, Pressure Testing of Drain, Vent, Test, and Fill Lines within the Reactor Coolant Pressure Boundary Ladies and Gentlemen:
Brunswick Steam Electric Plant, Unit Nos. 1 and 2 Docket Nos. 50-325 and 50-324/License Nos. DPR-71 and DPR-62 Relief Request RR-38, Pressure Testing of Drain, Vent, Test, and Fill Lines within the Reactor Coolant Pressure Boundary Ladies and Gentlemen:
In accordance with 10 CFR 50.55a(a)(3)(ii), Carolina Power & Light Company, now doing business as Progress Energy Carolinas, Inc., requests NRC approval of a relief request for the third 10-year interval Inservice Inspection Program for the Brunswick Steam Electric Plant (BSEP), Unit Nos. 1 and 2. The relief request proposes an alternative examination to perform the Class 1 system leakage test with the first reactor coolant pressure boundary drain, vent, test, and fill line isolation valves in the closed position.The details of Relief Request RR-38 are enclosed.
In accordance with 10 CFR 50.55a(a)(3)(ii), Carolina Power & Light Company, now doing business as Progress Energy Carolinas, Inc., requests NRC approval of a relief request for the third 10-year interval Inservice Inspection Program for the Brunswick Steam Electric Plant (BSEP), Unit Nos. 1 and 2. The relief request proposes an alternative examination to perform the Class 1 system leakage test with the first reactor coolant pressure boundary drain, vent, test, and fill line isolation valves in the closed position.
Approval of Relief Request RR-38 is requested by February 1, 2007, to support preparation activities for the Unit 2 refueling outage currently scheduled to begin March 10, 2007.No regulatory commitments are contained in this letter. Please refer any questions regarding this submittal to Mr. Leonard R. Beller, Supervisor  
The details of Relief Request RR-38 are enclosed. Approval of Relief Request RR-38 is requested by February 1, 2007, to support preparation activities for the Unit 2 refueling outage currently scheduled to begin March 10, 2007.
-Licensing/Regulatory Programs, at (910) 457-2073.Sincerely, q01 1C. CY.Randy C. Ivey Manager -Support Services Brunswick Steam Electric Plant Progress Energy Carolinas, Inc.Brunswick Nuclear Plant -7 P.O. Box 10429 7 4 Southport, NC 28461  
No regulatory commitments are contained in this letter. Please refer any questions regarding this submittal to Mr. Leonard R. Beller, Supervisor - Licensing/Regulatory Programs, at (910) 457-2073.
-I Document Control Desk BSEP 06-0074 / Page 2 WRM/wrm  
Sincerely, q01 1C. CY.
Randy C. Ivey Manager - Support Services Brunswick Steam Electric Plant Progress Energy Carolinas, Inc.
Brunswick Nuclear Plant  
-7 P.O. Box 10429 7 4 Southport, NC 28461
 
I Document Control Desk BSEP 06-0074 / Page 2 WRM/wrm


==Enclosure:==
==Enclosure:==
10 CFR 50.55a Relief Request Number RR-38 cc (with enclosure):
U. S. Nuclear Regulatory Commission, Region II ATTN: Dr. William D. Travers, Regional Administrator Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, GA 30303-8931 U. S. Nuclear Regulatory Commission ATTN: Mr. Eugene M. DiPaolo, NRC Senior Resident Inspector 8470 River Road Southport, NC 28461-8869 U. S. Nuclear Regulatory Commission (Electronic Copy Only)
ATTN: Ms. Brenda L. Mozafari (Mail Stop OWFN 8G9) 11555 Rockville Pike Rockville, MD 20852-2738 Ms. Jo A. Sanford Chair - North Carolina Utilities Commission P.O. Box 29510 Raleigh, NC 27626-05 10 Mr. Jack Given, Bureau Chief North Carolina Department of Labor Boiler Safety Bureau 1101 Mail Service Center Raleigh, NC 27699-1101
i 7-BSEP 06-0074 Enclosure Page I of 23 10 CFR 50.55a Relief Request Number RR-38 Proposed Alternative In Accordance with 10 CFR 50.55a(a)(3)(ii)
- Hardship or Unusual Difficulty Without Compensating Increase in Level of Quality and Safety -
: 1. ASME Code Components Affected Code Class: Class I Category: B-P System: Reactor Coolant Pressure Boundary (RCPB)
Affected Components: See Attachment 1 for a listing of the first isolation valves
===2. Applicable Code Edition and Addenda===
The Code of Record for the third 10-year inservice inspection interval at the Brunswick Steam Electric Plant (BSEP), Units 1 and 2, is the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section XI, 1989 Edition, with no addenda.
The third 10-year inservice inspection interval began May 11, 1998, and will conclude on May 10, 2008.
During the third 10-year inservice inspection interval, the alternative requirements of ASME Code Case N-498-4 are being implemented.
===3. Applicable Code Requirement===
The ASME Code, Section XI, Table IWB-2500-1, Examination Category B-P, Note 2, requires the pressure retaining boundary during the system hydrostatic test include all Class 1 components within the system boundary.
The alternative requirements of ASME Code Case N-498-4 requires that the boundary subject to test pressurization during the system leakage test extend to all Class 1 pressure retaining components within the system boundary.
===4. Reason for Request===
The drain, vent, test, and fill lines within the RCPB are typical one-inch nominal pipe size or less. These connections include two manual isolation valves whose purpose is to satisfy the
I U
BSEP 06-0074 Enclosure Page 2 of 23 design requirement for double isolation of the RCPB. During normal operation, these manual isolation valves are maintained in the closed or locked-closed position. Thus, components downstream of the first isolation valve are not subjected to reactor coolant system pressure unless leakage through the inboard valves occurs.
As stated above, Code requirements would require test pressurization to extend to all Class 1 pressure retaining components within the system boundary. To comply with this requirement, CP&L would be required to the open the first isolation valve. Having the first isolation valve open during the pressure test would defeat the design requirement for double isolation of the RCPB. As such, this non-standard configuration would increase the risk for inventory loss. Because of the potential for inventory loss, this configuration also creates safety concerns for the personnel performing the visual examination.
In addition, opening the first manual isolation valve will create a hardship in regards to personnel exposure and contamination. Opening these valves will require personnel to enter radiation fields to position the valves for the test, restore the valves following the test, and to perform the required independent valve position verification. Since these valves are typically located in close proximity to the main RCPB piping, CP&L estimates the dose associated with this effort, for each pressure test, as approximately 1.058 Rem for Unit 1 and 1.308 Rem for Unit 2. Because of the location of these valves, the risk for personnel contamination increases.
Based on CP&L's evaluation of this Code requirement, opening the first isolation valve to allow pressurization of the downstream components will not increase the level of quality or safety at the plant. As such, CP&L has concluded that placing the plant and personnel at risk is unwarranted. For this reason, CP&L is requesting relief from the requirement of the ASME Code, Section X1, Table IWB-2500-1, Examination Category B-P, Note 2, and ASME Code Case N-498-4(a)(5).
===5. Proposed Alternative and Basis for Use===
Proposed Alternative The VT-2 visual examination of the components downstream of the isolation valves listed in will extend to and include the second closed valve at the boundary extremity.
This visual examination will be performed with the isolation valves in their normal, closed operating position.
Basis for Use Because of the potential safety concerns and hardships, CP&L has concluded that the proposed alternative provides an acceptable level of safety and quality based on the following reasons:


10 CFR 50.55a Relief Request Number RR-38 cc (with enclosure):
BSEP 06-0074 Enclosure Page 3 of 23
U. S. Nuclear Regulatory Commission, Region II ATTN: Dr. William D. Travers, Regional Administrator Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, GA 30303-8931 U. S. Nuclear Regulatory Commission ATTN: Mr. Eugene M. DiPaolo, NRC Senior Resident Inspector 8470 River Road Southport, NC 28461-8869 U. S. Nuclear Regulatory Commission (Electronic Copy Only)ATTN: Ms. Brenda L. Mozafari (Mail Stop OWFN 8G9)11555 Rockville Pike Rockville, MD 20852-2738 Ms. Jo A. Sanford Chair -North Carolina Utilities Commission P.O. Box 29510 Raleigh, NC 27626-05 10 Mr. Jack Given, Bureau Chief North Carolina Department of Labor Boiler Safety Bureau 1101 Mail Service Center Raleigh, NC 27699-1101 i -, 7-BSEP 06-0074 Enclosure Page I of 23 10 CFR 50.55a Relief Request Number RR-38 Proposed Alternative In Accordance with 10 CFR 50.55a(a)(3)(ii)
: 1. The piping, fittings, and valves within these lines were designed and constructed to the highest standards. The components were designed for pressures and temperatures greater than they experience during normal operation. They were constructed to standards commensurate to the requirements of the ASME Code, Section III for Class 1 components. Because of these high standards, there is reasonable assurance that leakage integrity will be maintained during normal operation.
-Hardship or Unusual Difficulty Without Compensating Increase in Level of Quality and Safety -1. ASME Code Components Affected Code Class: Class I Category:
: 2. The proposed alternative is a proven method for assuring leakage integrity. This alternative is the same requirement that is used during the Code required system leakage test that is performed every refueling outage.
B-P System: Reactor Coolant Pressure Boundary (RCPB)Affected Components:
: 3. Only the isolable portion of these connections will not be pressurized during the test.
See Attachment 1 for a listing of the first isolation valves 2. Applicable Code Edition and Addenda The Code of Record for the third 10-year inservice inspection interval at the Brunswick Steam Electric Plant (BSEP), Units 1 and 2, is the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section XI, 1989 Edition, with no addenda.The third 10-year inservice inspection interval began May 11, 1998, and will conclude on May 10, 2008.During the third 10-year inservice inspection interval, the alternative requirements of ASME Code Case N-498-4 are being implemented.
Since these lines are in the same configuration during normal operation, approving this alternative poses no new safety concerns. As outlined in the alternative requirement, the VT-2 visual examination will extend to and include the second closed valve at the boundary extremity.
: 3. Applicable Code Requirement The ASME Code, Section XI, Table IWB-2500-1, Examination Category B-P, Note 2, requires the pressure retaining boundary during the system hydrostatic test include all Class 1 components within the system boundary.The alternative requirements of ASME Code Case N-498-4 requires that the boundary subject to test pressurization during the system leakage test extend to all Class 1 pressure retaining components within the system boundary.4. Reason for Request The drain, vent, test, and fill lines within the RCPB are typical one-inch nominal pipe size or less. These connections include two manual isolation valves whose purpose is to satisfy the I -U BSEP 06-0074 Enclosure Page 2 of 23 design requirement for double isolation of the RCPB. During normal operation, these manual isolation valves are maintained in the closed or locked-closed position.
Thus, components downstream of the first isolation valve are not subjected to reactor coolant system pressure unless leakage through the inboard valves occurs.As stated above, Code requirements would require test pressurization to extend to all Class 1 pressure retaining components within the system boundary.
To comply with this requirement, CP&L would be required to the open the first isolation valve. Having the first isolation valve open during the pressure test would defeat the design requirement for double isolation of the RCPB. As such, this non-standard configuration would increase the risk for inventory loss. Because of the potential for inventory loss, this configuration also creates safety concerns for the personnel performing the visual examination.
In addition, opening the first manual isolation valve will create a hardship in regards to personnel exposure and contamination.
Opening these valves will require personnel to enter radiation fields to position the valves for the test, restore the valves following the test, and to perform the required independent valve position verification.
Since these valves are typically located in close proximity to the main RCPB piping, CP&L estimates the dose associated with this effort, for each pressure test, as approximately 1.058 Rem for Unit 1 and 1.308 Rem for Unit 2. Because of the location of these valves, the risk for personnel contamination increases.
Based on CP&L's evaluation of this Code requirement, opening the first isolation valve to allow pressurization of the downstream components will not increase the level of quality or safety at the plant. As such, CP&L has concluded that placing the plant and personnel at risk is unwarranted.
For this reason, CP&L is requesting relief from the requirement of the ASME Code, Section X1, Table IWB-2500-1, Examination Category B-P, Note 2, and ASME Code Case N-498-4(a)(5).
: 5. Proposed Alternative and Basis for Use Proposed Alternative The VT-2 visual examination of the components downstream of the isolation valves listed in Attachment 1 will extend to and include the second closed valve at the boundary extremity.
This visual examination will be performed with the isolation valves in their normal, closed operating position.Basis for Use Because of the potential safety concerns and hardships, CP&L has concluded that the proposed alternative provides an acceptable level of safety and quality based on the following reasons:
BSEP 06-0074 Enclosure Page 3 of 23 1. The piping, fittings, and valves within these lines were designed and constructed to the highest standards.
The components were designed for pressures and temperatures greater than they experience during normal operation.
They were constructed to standards commensurate to the requirements of the ASME Code, Section III for Class 1 components.
Because of these high standards, there is reasonable assurance that leakage integrity will be maintained during normal operation.
: 2. The proposed alternative is a proven method for assuring leakage integrity.
This alternative is the same requirement that is used during the Code required system leakage test that is performed every refueling outage.3. Only the isolable portion of these connections will not be pressurized during the test.Since these lines are in the same configuration during normal operation, approving this alternative poses no new safety concerns.
As outlined in the alternative requirement, the VT-2 visual examination will extend to and include the second closed valve at the boundary extremity.
: 4. Not pressurizing component connections, piping, and valves that are one-inch nominal pipe size and smaller during a pressure test is an acceptable practice per the ASME Code, Section XI. After repairs by welding on the pressure-retaining boundary, a system hydrostatic test is exempt for these small diameter components per paragraph IWA-4700(a).
: 4. Not pressurizing component connections, piping, and valves that are one-inch nominal pipe size and smaller during a pressure test is an acceptable practice per the ASME Code, Section XI. After repairs by welding on the pressure-retaining boundary, a system hydrostatic test is exempt for these small diameter components per paragraph IWA-4700(a).
In summary, CP&L has concluded that extending the system pressure to the components downstream of the first normally closed isolation valve is a hardship and poses safety concerns for the plant and personnel.
In summary, CP&L has concluded that extending the system pressure to the components downstream of the first normally closed isolation valve is a hardship and poses safety concerns for the plant and personnel. The components affected by this relief request were designed and constructed to the highest standards available. The test configuration of these components is the same as they experience during normal operations and Code required system leakage test. In addition, extending the test pressure to these components once per ten years is unjustifiable considering these same components would be exempt from pressure testing if repaired or replaced. For these reasons, approving the use of the proposed alternative will provide an acceptable level of safety and quality.
The components affected by this relief request were designed and constructed to the highest standards available.
 
The test configuration of these components is the same as they experience during normal operations and Code required system leakage test. In addition, extending the test pressure to these components once per ten years is unjustifiable considering these same components would be exempt from pressure testing if repaired or replaced.
===6. Duration of Proposed Alternative===
For these reasons, approving the use of the proposed alternative will provide an acceptable level of safety and quality.6. Duration of Proposed Alternative Use of the alternative is proposed for the remainder of the current 10-year inservice inspection interval.7. Precedents This proposed alternative is similar, but not identical, to a relief request submitted by the Hatch Nuclear Plant in a letter dated March 30, 2005 (i.e., ADAMS Accession Number ML050940201), as approved by NRC letter dated November 9, 2005 (i.e., ADAMS Accession Number ML052970008).
Use of the alternative is proposed for the remainder of the current 10-year inservice inspection interval.
BSEP 06-0074 Enclosure Page 4 of 23 8. References
 
: 1. Title 10 of the Code of Federal Regulations, Part 50, Section 55a, Codes and Standards (i.e., 10 CFR 50.55a).2. ASME Code, Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, 1989 Edition (no Addenda).3. ASME Code Case N-498-4, Alternative Requirements for 10-Year System Hydrostatic Testing of Class 1, 2, and 3 System.
===7. Precedents===
i BSEP 06-0074 Enclosure Page 5 of 23 Attachment 1 First General Isolation Normal Location Valve Description Position Drawing B21-FOO1 RX Inboard High Point Vent Valve Locked Figure 14 Closed B21-V18 B21-FO11A Inboard Body Drain Valve Closed Figure 2 B21-V20 B21-FO1OA Inboard Body Drain Valve Closed Figure 2 B21-V23 B21-F011B Inboard Body Drain Valve Closed Figure 3 B21-V25 B21-F01OB Inboard Body Drain Valve Closed Figure 3 B21-V160 Test valve for Excess Flow Check Valve Closed Figure 9 B21-IV-2456(Pen.
This proposed alternative is similar, but not identical, to a relief request submitted by the Hatch Nuclear Plant in a {{letter dated|date=March 30, 2005|text=letter dated March 30, 2005}} (i.e., ADAMS Accession Number ML050940201), as approved by NRC {{letter dated|date=November 9, 2005|text=letter dated November 9, 2005}} (i.e., ADAMS Accession Number ML052970008).
X-49A-A)B21-V161 Test valve for Excess Flow Check Valve Locked Figure 10 B21-IV-2455(Pen.
 
X-49B-A) Closed B21-V167 B21-F016 Test Inboard Isolation Valve Locked Figure 5 Closed B32-F025A B32-F023A Vent Root Valve Closed Figure 7 B32-F025B B32-F023B Vent Root Valve Closed Figure 8 B32-F027A B32-F023A Inboard Drain Valve Closed Figure 7 B32-F027B B32-F023B Inboard Drain Valve Closed Figure 8 B32-F029 Reactor Pressure Vessel Drain Inboard Valve Closed Figure 15 B32-F034A B32-FO31A Inboard Vent Valve Closed Figure 7 B32-F034B B32-FO31B Inboard Vent Valve Closed Figure 8 B32-F036A B32-F031A Inboard Body Drain Valve Closed Figure 7 B32-F036B B32-FO31B Inboard Body Drain Valve Closed Figure 8 B32-F046A B32-F043A Inboard Root Valve Closed Figure 7 B32-F046B B32-F043B Inboard Root Valve Closed Figure 8 B32-F048A B32-F043A Inboard Body Drain Valve Closed Figure 7  
BSEP 06-0074 Enclosure Page 4 of 23
.1 L BSEP 06-0074 Enclosure Page 6 of 23 Attachment 1 First General Isolation Normal Location Valve Description Position Drawing B32-F048B B32-F043B Inboard Body Drain Valve Closed Figure 8 B32-V36 B32-F032A Vent Root Valve Closed Figure 7 B32-V38 B32-F032B Vent Root Valve Closed Figure 8 B32-FO51A Reactor Recirculation Loop A Inboard Drain Closed Figure 7 (HW)B32-FO51B Reactor Recirculation Loop B Inboard Drain Closed Figure 8 (HW)C41-V8 C41-F007 Inboard Test Isolation Valve Locked Figure 16 Closed 1-Ell-V112 El.-F060A Inboard Body Drain Valve Closed Figure 11 Ell-V117 El1-F050A Inboard Body Drain Valve Closed Figure 11 Ell-V130 Ell-F050B Inboard Body Drain Valve Closed Figure 12 1-Ell-V132 E1-F060B Inboard Body Drain Valve Closed Figure 12 Ell-V82 Ell-FO15A Inboard Body Drain Valve Locked Figure 11 Closed Ell-V169 Ell-FO15B Inboard Body Drain Valve Locked Figure 12 Closed Ell-V5000 E11-F009 Inboard LLRT Test Connection Locked Figure 13 Closed E21-V27 E21-F006A Downstream Inboard Body Drain Closed Figure 1 Valve E21-V39 E21-FO06B Downstream Inboard Body Drain Closed Figure 1 Valve 2-E21-V41 Core Spray Div II Inboard Vent Valve Closed Figure 1 E21-V67 E21-FO07A Inboard Body Drain Valve Closed Figure 1 E21-V69 E21-FO07B Inboard Body Drain Valve Closed Figure 1 BSEP 06-0074 Enclosure Page 7 of 23 Attachment 1 First General Isolation Normal Location Valve Description Position Drawing E41-V174 E41-F002 Inboard ISI Test Valve Locked Figure 4 Closed E51-V1O1 E51-F007 Inboard ISI Test Valve Locked Figure 6 Closed G31-F002 RWCU Inlet Line Test Valve Locked Figure 15 Closed BSEP 06-0074 Enclosure Page 8 of 23 Figure 1 -Core Spray (CS) A and B Loops DRYWELL PENETRATION X-16B EL.62'-9" AZ. 28d'*V74*V73 DRYWELL PENETRATION X-16A EL.67'-9" AZ. 790* Unit I only** Unit 2 only BSEP 06-0074 Enclosure Page 9 of 23 Figure 2 -Feedwater A Loop BSEP 06-0074 Enclosure Page 10 of 23 Figure 3 -Feedwater B Loop BSEP 06-0074 Enclosure Page 11 of 23 Figure 4 -High Pressure Coolant Injection (IPCI) Steam Line EL. 36'-0" AZ. 2310 BSEP 06-0074 Enclosure Page 12 of 23 Figure 5 -Primary Steam A Line]FO13B PRIMARY STEAM LINE "A" DRYWELL PEN. X-7A EL. 22'-4", AZ. 50 DRYWELL PEN. X-8 EL. 20'-2 ',, AZ. 00 PRIMARY STEAM LINE "B" PRIMARY STEAM LINE "C" PRIMARY STEAM LINE "D" BSEP 06-0074 Enclosure Page 13 of 23 Figure 6 -Reactor Core Isolation Cooling (RCIC) Steam Line DRYWELL PENETRATION V86 F043D X-72E EL. 36'-O" AZ .2440 \ DRYWELL PENETRATION X-72F EL. 36'-0" AZ. 2440 PRIMARY CONTAINMENT DRYWELL PENETRATION ON X-61E EL. 36'-0" AZ. 900 DRYWELL PENETRATION X-61F EL. 36'-0" AZ. 900 DRYWELL PENETRATION X-10 EL. 231-6" AZ. 1850  
: 8. References
.B BSEP 06-0074 Enclosure Page 14 of 23 Figure 7 -Reactor Recirculation A Loop PENETRATION X-56E BSEP 06-0074 Enclosure Page 15 of 23 Figure 8 -Reactor Recirculation B Loop*Unit Two Only Figure 9 -Nuclear Steam Supply Instrumentation REACTOR VESSEL BSEP 06-0074 Enclosure Page 16 of 23 62 4.EL. 86'-0" E7 AZ. 2250 EL. 62'-6" AZ. 2250-X69D-B21.776 X-69E-X-69F ID'D 2D'U S EL. 36-0" AZ. 2700 P M ZBýH 3l .E EL. 38'-0 n -AZ. 2700 At BSEP 06-0074 Enclosure Page 17 of 23 Figure 10 -Nuclear Steam Supply Instrumentation REACTOR VESSEL i BSEP 06-0074 Enclosure Page 18 of 23 Figure 11 -Residual Heat Removal (RHR) A Loop r--------  
: 1. Title 10 of the Code of Federal Regulations, Part 50, Section 55a, Codes and Standards (i.e., 10 CFR 50.55a).
---------- N2A r r-- ---------------------  
: 2. ASME Code, Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, 1989 Edition (no Addenda).
-------N2B----------------
: 3. ASME Code Case N-498-4, Alternative Requirements for 10-Year System Hydrostatic Testing of Class 1, 2, and 3 System.
jN2C---- ----- ----------------
 
N2D-'---- -------------------
i BSEP 06-0074 Enclosure Page 5 of 23 First General Isolation Normal Location Valve Description Position Drawing B21-FOO1 RX Inboard High Point Vent Valve Locked Figure 14 Closed B21-V18 B21-FO11A Inboard Body Drain Valve Closed Figure 2 B21-V20 B21-FO1OA Inboard Body Drain Valve Closed Figure 2 B21-V23 B21-F011B Inboard Body Drain Valve Closed Figure 3 B21-V25 B21-F01OB Inboard Body Drain Valve Closed Figure 3 B21-V160 Test valve for Excess Flow Check Valve Closed Figure 9 B21-IV-2456(Pen. X-49A-A)
N2E-NIB PRIMARY --CONTAINMENT u .c oLL_ _ -LLI REACTOR-- -PRESSURE VESSEL EL \I I r"--IREATO o _ I RECII01 V114 V32 I PUMPI Vil V11 COO1A 1`0050A I VI V112 vI I-""' -t -i 4- --,- F " FO31A IY It- _-j:xl- --X-13A EL. 23'-6" AZ. 1600*Unit One Only
B21-V161 Test valve for Excess Flow Check Valve Locked Figure 10 B21-IV-2455(Pen. X-49B-A)
..-I BSEP 06-0074 Enclosure Page 19 of 23 Figure 12 -Residual Heat Removal (RIIR) B Loop GRTG. EL. 38'-0" REACTOR RECIRC.PUMP C001B PRIMARY CONTAINMENT
Closed B21-V167 B21-F016 Test Inboard Isolation Valve Locked Figure 5 Closed B32-F025A B32-F023A Vent Root Valve Closed Figure 7 B32-F025B B32-F023B Vent Root Valve Closed Figure 8 B32-F027A B32-F023A Inboard Drain Valve Closed Figure 7 B32-F027B B32-F023B Inboard Drain Valve Closed Figure 8 B32-F029 Reactor Pressure Vessel Drain Inboard Valve Closed Figure 15 B32-F034A B32-FO31A Inboard Vent Valve Closed Figure 7 B32-F034B B32-FO31B Inboard Vent Valve Closed Figure 8 B32-F036A B32-F031A Inboard Body Drain Valve Closed Figure 7 B32-F036B B32-FO31B Inboard Body Drain Valve Closed Figure 8 B32-F046A B32-F043A Inboard Root Valve Closed Figure 7 B32-F046B B32-F043B Inboard Root Valve Closed Figure 8 B32-F048A B32-F043A Inboard Body Drain Valve Closed Figure 7
"--------V128--V33 4 F,._F050B DRY WELL PEN. (X-13B)EL. 30'-0"----------
 
AZ 270 0  F00B 28F---------------
.1 L
1-4--4 --- --;F031B 87-24-900[-l~l....._
BSEP 06-0074 Enclosure Page 6 of 23 First General Isolation Normal Location Valve Description Position Drawing B32-F048B B32-F043B Inboard Body Drain Valve Closed Figure 8 B32-V36 B32-F032A Vent Root Valve Closed Figure 7 B32-V38 B32-F032B Vent Root Valve Closed Figure 8 B32-FO51A Reactor Recirculation Loop A Inboard Drain Closed Figure 7 (HW)
I EL. 23'-6" AZ 2000 GRTG. EL. 17.-0"*UNIT I ONLY
B32-FO51B Reactor Recirculation Loop B Inboard Drain Closed Figure 8 (HW)
*BSEP 06-0074 Enclosure Page 20 of 23 Figure 13 -Residual Heat Removal (RHR) Shutdown Cooling V103 BSEP 06-0074 Enclosure Page 21 of 23 Figure 14 -RPV Closure Head Vent Lines 706-3/4-603
C41-V8 C41-F007 Inboard Test Isolation Valve Locked Figure 16 Closed 1-Ell-V112 El.-F060A Inboard Body Drain Valve Closed Figure 11 Ell-V117 El1-F050A Inboard Body Drain Valve Closed Figure 11 Ell-V130 Ell-F050B Inboard Body Drain Valve Closed Figure 12 1-Ell-V132 E1-F060B Inboard Body Drain Valve Closed Figure 12 Ell-V82 Ell-FO15A Inboard Body Drain Valve Locked Figure 11 Closed Ell-V169 Ell-FO15B Inboard Body Drain Valve Locked Figure 12 Closed Ell-V5000 E11-F009 Inboard LLRT Test Connection Locked Figure 13 Closed E21-V27 E21-F006A Downstream Inboard Body Drain Closed Figure 1 Valve E21-V39 E21-FO06B Downstream Inboard Body Drain Closed Figure 1 Valve 2-E21-V41 Core Spray Div II Inboard Vent Valve Closed Figure 1 E21-V67 E21-FO07A Inboard Body Drain Valve Closed Figure 1 E21-V69 E21-FO07B Inboard Body Drain Valve Closed Figure 1
.l r, I" ,-.BSEP 06-0074 Enclosure Page 22 of 23 Figure 15 -Reactor Water Cleanup (RWCU)DRYWELL PENETRATION X-14 EL. 62'-9 AZ. 2450 LF004 LF027[F 1001 1-6-603"-K 4u--PRIMARY CONTAINMENT G31 Ell I RHR F10-20-6034 F067"A" LOOP RECIRCULATION SUCTION LINE REACTOR VESSEL (0203-o 050 16-2-603 G31 B N15 26-2--600 F030 F030 32d BSEP 06-0074 Enclosure Page 23 of 23 Figure 16 -Standby Liquid Control (SLC)DRYWELL PEN. X-42 EL. 71'-0" -AZ. 2100 11/2"}}
 
BSEP 06-0074 Enclosure Page 7 of 23 First General Isolation Normal Location Valve Description Position Drawing E41-V174 E41-F002 Inboard ISI Test Valve Locked Figure 4 Closed E51-V1O1 E51-F007 Inboard ISI Test Valve Locked Figure 6 Closed G31-F002 RWCU Inlet Line Test Valve Locked Figure 15 Closed
 
BSEP 06-0074 Enclosure Page 8 of 23 Figure 1 - Core Spray (CS) A and B Loops DRYWELL PENETRATION X-16B EL.62'-9" AZ. 28d'
*V74
*V73 DRYWELL PENETRATION X-16A EL.67'-9" AZ. 790
* Unit I only
** Unit 2 only
 
BSEP 06-0074 Enclosure Page 9 of 23 Figure 2 - Feedwater A Loop
 
BSEP 06-0074 Enclosure Page 10 of 23 Figure 3 - Feedwater B Loop
 
BSEP 06-0074 Enclosure Page 11 of 23 Figure 4 - High Pressure Coolant Injection (IPCI) Steam Line EL. 36'-0" AZ. 2310
 
BSEP 06-0074 Enclosure Page 12 of 23 Figure 5 - Primary Steam A Line
]FO13B PRIMARY STEAM LINE "A" DRYWELL PEN. X-7A EL. 22'-4", AZ. 50 DRYWELL PEN. X-8 EL. 20'-2 ',, AZ. 00 PRIMARY STEAM LINE "B" PRIMARY STEAM LINE "C" PRIMARY STEAM LINE "D"
 
BSEP 06-0074 Enclosure Page 13 of 23 Figure 6 - Reactor Core Isolation Cooling (RCIC) Steam Line DRYWELL PENETRATION V86 F043D X-72E EL. 36'-O" AZ.2440  
\\
DRYWELL PENETRATION X-72F EL. 36'-0" AZ. 2440 PRIMARY CONTAINMENT DRYWELL PENETRATION ON X-61E EL. 36'-0" AZ. 900 DRYWELL PENETRATION X-61F EL. 36'-0" AZ. 900 DRYWELL PENETRATION X-10 EL. 231-6" AZ. 1850
 
B BSEP 06-0074 Enclosure Page 14 of 23 Figure 7 - Reactor Recirculation A Loop PENETRATION X-56E
 
BSEP 06-0074 Enclosure Page 15 of 23 Figure 8 - Reactor Recirculation B Loop
*Unit Two Only
 
Figure 9 - Nuclear Steam Supply Instrumentation REACTOR VESSEL BSEP 06-0074 Enclosure Page 16 of 23 62 4
EL. 86'-0" E7 AZ. 2250 EL. 62'-6" AZ. 2250 X69D B21.776 X-69E
-X-69F ID
'D 2D
'U S
EL. 36-0" AZ. 2700 P
M ZB
ýH 3l  
.E EL. 38'-0 n
AZ. 2700 At
 
BSEP 06-0074 Enclosure Page 17 of 23 Figure 10 - Nuclear Steam Supply Instrumentation REACTOR VESSEL
 
i BSEP 06-0074 Enclosure Page 18 of 23 Figure 11 - Residual Heat Removal (RHR) A Loop r--------
N2A r
r--
N2B jN2C N2D N2E
- NIB PRIMARY CONTAINMENT u.c oLL_
LLI REACTOR
-PRESSURE VESSEL EL
\\
I I
r"--
IREATO o _
I RECII01 V114 V32 I PUMPI Vil V11 COO1A 1`0050A I
VI V112 vI I
-t
- i 4-  
-,- F "
FO31A IYIt- _-j:xl-X-13A EL. 23'-6" AZ. 1600
*Unit One Only
 
..-I BSEP 06-0074 Enclosure Page 19 of 23 Figure 12 - Residual Heat Removal (RIIR) B Loop GRTG. EL. 38'-0" REACTOR RECIRC.
PUMP C001B PRIMARY CONTAINMENT
"--------V128--V33 4
F,._F050B DRY WELL PEN. (X-13B)
EL. 30'-0"
---------- AZ 2700 F00B 28F 1-4--4 F031B 87-24-900
[-l~l....._
I EL. 23'-6" AZ 2000 GRTG. EL. 17.-0"
*UNIT I ONLY
 
BSEP 06-0074 Enclosure Page 20 of 23 Figure 13 - Residual Heat Removal (RHR) Shutdown Cooling V103
 
BSEP 06-0074 Enclosure Page 21 of 23 Figure 14 - RPV Closure Head Vent Lines 706-3/4-603
 
.l r, I" BSEP 06-0074 Enclosure Page 22 of 23 Figure 15 - Reactor Water Cleanup (RWCU)
DRYWELL PENETRATION X-14 EL. 62'-9 AZ. 2450 LF004 LF027
[F 1001 1-6-603"-
K 4u--PRIMARY CONTAINMENT G31 Ell I RHR F10-20-6034 F067 "A" LOOP RECIRCULATION SUCTION LINE REACTOR VESSEL
(
0203
-o 050 16-2-603 G31 B
N15 26-2--600 F030 F030 32d
 
BSEP 06-0074 Enclosure Page 23 of 23 Figure 16 - Standby Liquid Control (SLC)
DRYWELL PEN. X-42 EL. 71'-0" - AZ. 2100 11/2"}}

Latest revision as of 07:39, 15 January 2025

Relief Request RR-38, Pressure Testing of Drain, Vent, and Fill Lines within the Reactor Coolant Pressure Boundary
ML062140037
Person / Time
Site: Brunswick  Duke Energy icon.png
Issue date: 07/18/2006
From: Ivey R
Progress Energy Carolinas
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
BSEP 06-0074, RR-38
Download: ML062140037 (25)
v • d • e


Text

I-

£ JProgress Energy 10 CFR 50.55a(a)(3)(ii)

JUL 18 2006 SERIAL: BSEP 06-0074 U. S. Nuclear Regulatory Commission ATIN: Document Control Desk Washington, DC 20555-0001

Subject:

Brunswick Steam Electric Plant, Unit Nos. 1 and 2 Docket Nos. 50-325 and 50-324/License Nos. DPR-71 and DPR-62 Relief Request RR-38, Pressure Testing of Drain, Vent, Test, and Fill Lines within the Reactor Coolant Pressure Boundary Ladies and Gentlemen:

In accordance with 10 CFR 50.55a(a)(3)(ii), Carolina Power & Light Company, now doing business as Progress Energy Carolinas, Inc., requests NRC approval of a relief request for the third 10-year interval Inservice Inspection Program for the Brunswick Steam Electric Plant (BSEP), Unit Nos. 1 and 2. The relief request proposes an alternative examination to perform the Class 1 system leakage test with the first reactor coolant pressure boundary drain, vent, test, and fill line isolation valves in the closed position.

The details of Relief Request RR-38 are enclosed. Approval of Relief Request RR-38 is requested by February 1, 2007, to support preparation activities for the Unit 2 refueling outage currently scheduled to begin March 10, 2007.

No regulatory commitments are contained in this letter. Please refer any questions regarding this submittal to Mr. Leonard R. Beller, Supervisor - Licensing/Regulatory Programs, at (910) 457-2073.

Sincerely, q01 1C. CY.

Randy C. Ivey Manager - Support Services Brunswick Steam Electric Plant Progress Energy Carolinas, Inc.

Brunswick Nuclear Plant

-7 P.O. Box 10429 7 4 Southport, NC 28461

I Document Control Desk BSEP 06-0074 / Page 2 WRM/wrm

Enclosure:

10 CFR 50.55a Relief Request Number RR-38 cc (with enclosure):

U. S. Nuclear Regulatory Commission, Region II ATTN: Dr. William D. Travers, Regional Administrator Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, GA 30303-8931 U. S. Nuclear Regulatory Commission ATTN: Mr. Eugene M. DiPaolo, NRC Senior Resident Inspector 8470 River Road Southport, NC 28461-8869 U. S. Nuclear Regulatory Commission (Electronic Copy Only)

ATTN: Ms. Brenda L. Mozafari (Mail Stop OWFN 8G9) 11555 Rockville Pike Rockville, MD 20852-2738 Ms. Jo A. Sanford Chair - North Carolina Utilities Commission P.O. Box 29510 Raleigh, NC 27626-05 10 Mr. Jack Given, Bureau Chief North Carolina Department of Labor Boiler Safety Bureau 1101 Mail Service Center Raleigh, NC 27699-1101

i 7-BSEP 06-0074 Enclosure Page I of 23 10 CFR 50.55a Relief Request Number RR-38 Proposed Alternative In Accordance with 10 CFR 50.55a(a)(3)(ii)

- Hardship or Unusual Difficulty Without Compensating Increase in Level of Quality and Safety -

1. ASME Code Components Affected Code Class: Class I Category: B-P System: Reactor Coolant Pressure Boundary (RCPB)

Affected Components: See Attachment 1 for a listing of the first isolation valves

2. Applicable Code Edition and Addenda

The Code of Record for the third 10-year inservice inspection interval at the Brunswick Steam Electric Plant (BSEP), Units 1 and 2, is the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, 1989 Edition, with no addenda.

The third 10-year inservice inspection interval began May 11, 1998, and will conclude on May 10, 2008.

During the third 10-year inservice inspection interval, the alternative requirements of ASME Code Case N-498-4 are being implemented.

3. Applicable Code Requirement

The ASME Code,Section XI, Table IWB-2500-1, Examination Category B-P, Note 2, requires the pressure retaining boundary during the system hydrostatic test include all Class 1 components within the system boundary.

The alternative requirements of ASME Code Case N-498-4 requires that the boundary subject to test pressurization during the system leakage test extend to all Class 1 pressure retaining components within the system boundary.

4. Reason for Request

The drain, vent, test, and fill lines within the RCPB are typical one-inch nominal pipe size or less. These connections include two manual isolation valves whose purpose is to satisfy the

I U

BSEP 06-0074 Enclosure Page 2 of 23 design requirement for double isolation of the RCPB. During normal operation, these manual isolation valves are maintained in the closed or locked-closed position. Thus, components downstream of the first isolation valve are not subjected to reactor coolant system pressure unless leakage through the inboard valves occurs.

As stated above, Code requirements would require test pressurization to extend to all Class 1 pressure retaining components within the system boundary. To comply with this requirement, CP&L would be required to the open the first isolation valve. Having the first isolation valve open during the pressure test would defeat the design requirement for double isolation of the RCPB. As such, this non-standard configuration would increase the risk for inventory loss. Because of the potential for inventory loss, this configuration also creates safety concerns for the personnel performing the visual examination.

In addition, opening the first manual isolation valve will create a hardship in regards to personnel exposure and contamination. Opening these valves will require personnel to enter radiation fields to position the valves for the test, restore the valves following the test, and to perform the required independent valve position verification. Since these valves are typically located in close proximity to the main RCPB piping, CP&L estimates the dose associated with this effort, for each pressure test, as approximately 1.058 Rem for Unit 1 and 1.308 Rem for Unit 2. Because of the location of these valves, the risk for personnel contamination increases.

Based on CP&L's evaluation of this Code requirement, opening the first isolation valve to allow pressurization of the downstream components will not increase the level of quality or safety at the plant. As such, CP&L has concluded that placing the plant and personnel at risk is unwarranted. For this reason, CP&L is requesting relief from the requirement of the ASME Code, Section X1, Table IWB-2500-1, Examination Category B-P, Note 2, and ASME Code Case N-498-4(a)(5).

5. Proposed Alternative and Basis for Use

Proposed Alternative The VT-2 visual examination of the components downstream of the isolation valves listed in will extend to and include the second closed valve at the boundary extremity.

This visual examination will be performed with the isolation valves in their normal, closed operating position.

Basis for Use Because of the potential safety concerns and hardships, CP&L has concluded that the proposed alternative provides an acceptable level of safety and quality based on the following reasons:

BSEP 06-0074 Enclosure Page 3 of 23

1. The piping, fittings, and valves within these lines were designed and constructed to the highest standards. The components were designed for pressures and temperatures greater than they experience during normal operation. They were constructed to standards commensurate to the requirements of the ASME Code,Section III for Class 1 components. Because of these high standards, there is reasonable assurance that leakage integrity will be maintained during normal operation.
2. The proposed alternative is a proven method for assuring leakage integrity. This alternative is the same requirement that is used during the Code required system leakage test that is performed every refueling outage.
3. Only the isolable portion of these connections will not be pressurized during the test.

Since these lines are in the same configuration during normal operation, approving this alternative poses no new safety concerns. As outlined in the alternative requirement, the VT-2 visual examination will extend to and include the second closed valve at the boundary extremity.

4. Not pressurizing component connections, piping, and valves that are one-inch nominal pipe size and smaller during a pressure test is an acceptable practice per the ASME Code,Section XI. After repairs by welding on the pressure-retaining boundary, a system hydrostatic test is exempt for these small diameter components per paragraph IWA-4700(a).

In summary, CP&L has concluded that extending the system pressure to the components downstream of the first normally closed isolation valve is a hardship and poses safety concerns for the plant and personnel. The components affected by this relief request were designed and constructed to the highest standards available. The test configuration of these components is the same as they experience during normal operations and Code required system leakage test. In addition, extending the test pressure to these components once per ten years is unjustifiable considering these same components would be exempt from pressure testing if repaired or replaced. For these reasons, approving the use of the proposed alternative will provide an acceptable level of safety and quality.

6. Duration of Proposed Alternative

Use of the alternative is proposed for the remainder of the current 10-year inservice inspection interval.

7. Precedents

This proposed alternative is similar, but not identical, to a relief request submitted by the Hatch Nuclear Plant in a letter dated March 30, 2005 (i.e., ADAMS Accession Number ML050940201), as approved by NRC letter dated November 9, 2005 (i.e., ADAMS Accession Number ML052970008).

BSEP 06-0074 Enclosure Page 4 of 23

8. References
1. Title 10 of the Code of Federal Regulations, Part 50, Section 55a, Codes and Standards (i.e., 10 CFR 50.55a).
2. ASME Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, 1989 Edition (no Addenda).
3. ASME Code Case N-498-4, Alternative Requirements for 10-Year System Hydrostatic Testing of Class 1, 2, and 3 System.

i BSEP 06-0074 Enclosure Page 5 of 23 First General Isolation Normal Location Valve Description Position Drawing B21-FOO1 RX Inboard High Point Vent Valve Locked Figure 14 Closed B21-V18 B21-FO11A Inboard Body Drain Valve Closed Figure 2 B21-V20 B21-FO1OA Inboard Body Drain Valve Closed Figure 2 B21-V23 B21-F011B Inboard Body Drain Valve Closed Figure 3 B21-V25 B21-F01OB Inboard Body Drain Valve Closed Figure 3 B21-V160 Test valve for Excess Flow Check Valve Closed Figure 9 B21-IV-2456(Pen. X-49A-A)

B21-V161 Test valve for Excess Flow Check Valve Locked Figure 10 B21-IV-2455(Pen. X-49B-A)

Closed B21-V167 B21-F016 Test Inboard Isolation Valve Locked Figure 5 Closed B32-F025A B32-F023A Vent Root Valve Closed Figure 7 B32-F025B B32-F023B Vent Root Valve Closed Figure 8 B32-F027A B32-F023A Inboard Drain Valve Closed Figure 7 B32-F027B B32-F023B Inboard Drain Valve Closed Figure 8 B32-F029 Reactor Pressure Vessel Drain Inboard Valve Closed Figure 15 B32-F034A B32-FO31A Inboard Vent Valve Closed Figure 7 B32-F034B B32-FO31B Inboard Vent Valve Closed Figure 8 B32-F036A B32-F031A Inboard Body Drain Valve Closed Figure 7 B32-F036B B32-FO31B Inboard Body Drain Valve Closed Figure 8 B32-F046A B32-F043A Inboard Root Valve Closed Figure 7 B32-F046B B32-F043B Inboard Root Valve Closed Figure 8 B32-F048A B32-F043A Inboard Body Drain Valve Closed Figure 7

.1 L

BSEP 06-0074 Enclosure Page 6 of 23 First General Isolation Normal Location Valve Description Position Drawing B32-F048B B32-F043B Inboard Body Drain Valve Closed Figure 8 B32-V36 B32-F032A Vent Root Valve Closed Figure 7 B32-V38 B32-F032B Vent Root Valve Closed Figure 8 B32-FO51A Reactor Recirculation Loop A Inboard Drain Closed Figure 7 (HW)

B32-FO51B Reactor Recirculation Loop B Inboard Drain Closed Figure 8 (HW)

C41-V8 C41-F007 Inboard Test Isolation Valve Locked Figure 16 Closed 1-Ell-V112 El.-F060A Inboard Body Drain Valve Closed Figure 11 Ell-V117 El1-F050A Inboard Body Drain Valve Closed Figure 11 Ell-V130 Ell-F050B Inboard Body Drain Valve Closed Figure 12 1-Ell-V132 E1-F060B Inboard Body Drain Valve Closed Figure 12 Ell-V82 Ell-FO15A Inboard Body Drain Valve Locked Figure 11 Closed Ell-V169 Ell-FO15B Inboard Body Drain Valve Locked Figure 12 Closed Ell-V5000 E11-F009 Inboard LLRT Test Connection Locked Figure 13 Closed E21-V27 E21-F006A Downstream Inboard Body Drain Closed Figure 1 Valve E21-V39 E21-FO06B Downstream Inboard Body Drain Closed Figure 1 Valve 2-E21-V41 Core Spray Div II Inboard Vent Valve Closed Figure 1 E21-V67 E21-FO07A Inboard Body Drain Valve Closed Figure 1 E21-V69 E21-FO07B Inboard Body Drain Valve Closed Figure 1

BSEP 06-0074 Enclosure Page 7 of 23 First General Isolation Normal Location Valve Description Position Drawing E41-V174 E41-F002 Inboard ISI Test Valve Locked Figure 4 Closed E51-V1O1 E51-F007 Inboard ISI Test Valve Locked Figure 6 Closed G31-F002 RWCU Inlet Line Test Valve Locked Figure 15 Closed

BSEP 06-0074 Enclosure Page 8 of 23 Figure 1 - Core Spray (CS) A and B Loops DRYWELL PENETRATION X-16B EL.62'-9" AZ. 28d'

  • V74
  • Unit I only
    • Unit 2 only

BSEP 06-0074 Enclosure Page 9 of 23 Figure 2 - Feedwater A Loop

BSEP 06-0074 Enclosure Page 10 of 23 Figure 3 - Feedwater B Loop

BSEP 06-0074 Enclosure Page 11 of 23 Figure 4 - High Pressure Coolant Injection (IPCI) Steam Line EL. 36'-0" AZ. 2310

BSEP 06-0074 Enclosure Page 12 of 23 Figure 5 - Primary Steam A Line

]FO13B PRIMARY STEAM LINE "A" DRYWELL PEN. X-7A EL. 22'-4", AZ. 50 DRYWELL PEN. X-8 EL. 20'-2 ',, AZ. 00 PRIMARY STEAM LINE "B" PRIMARY STEAM LINE "C" PRIMARY STEAM LINE "D"

BSEP 06-0074 Enclosure Page 13 of 23 Figure 6 - Reactor Core Isolation Cooling (RCIC) Steam Line DRYWELL PENETRATION V86 F043D X-72E EL. 36'-O" AZ.2440

\\

DRYWELL PENETRATION X-72F EL. 36'-0" AZ. 2440 PRIMARY CONTAINMENT DRYWELL PENETRATION ON X-61E EL. 36'-0" AZ. 900 DRYWELL PENETRATION X-61F EL. 36'-0" AZ. 900 DRYWELL PENETRATION X-10 EL. 231-6" AZ. 1850

B BSEP 06-0074 Enclosure Page 14 of 23 Figure 7 - Reactor Recirculation A Loop PENETRATION X-56E

BSEP 06-0074 Enclosure Page 15 of 23 Figure 8 - Reactor Recirculation B Loop

  • Unit Two Only

Figure 9 - Nuclear Steam Supply Instrumentation REACTOR VESSEL BSEP 06-0074 Enclosure Page 16 of 23 62 4

EL. 86'-0" E7 AZ. 2250 EL. 62'-6" AZ. 2250 X69D B21.776 X-69E

-X-69F ID

'D 2D

'U S

EL. 36-0" AZ. 2700 P

M ZB

ýH 3l

.E EL. 38'-0 n

AZ. 2700 At

BSEP 06-0074 Enclosure Page 17 of 23 Figure 10 - Nuclear Steam Supply Instrumentation REACTOR VESSEL

i BSEP 06-0074 Enclosure Page 18 of 23 Figure 11 - Residual Heat Removal (RHR) A Loop r--------

N2A r

r--

N2B jN2C N2D N2E

- NIB PRIMARY CONTAINMENT u.c oLL_

LLI REACTOR

-PRESSURE VESSEL EL

\\

I I

r"--

IREATO o _

I RECII01 V114 V32 I PUMPI Vil V11 COO1A 1`0050A I

VI V112 vI I

-t

- i 4-

-,- F "

FO31A IYIt- _-j:xl-X-13A EL. 23'-6" AZ. 1600

  • Unit One Only

..-I BSEP 06-0074 Enclosure Page 19 of 23 Figure 12 - Residual Heat Removal (RIIR) B Loop GRTG. EL. 38'-0" REACTOR RECIRC.

PUMP C001B PRIMARY CONTAINMENT

"--------V128--V33 4

F,._F050B DRY WELL PEN. (X-13B)

EL. 30'-0"

AZ 2700 F00B 28F 1-4--4 F031B 87-24-900

[-l~l....._

I EL. 23'-6" AZ 2000 GRTG. EL. 17.-0"

  • UNIT I ONLY

BSEP 06-0074 Enclosure Page 20 of 23 Figure 13 - Residual Heat Removal (RHR) Shutdown Cooling V103

BSEP 06-0074 Enclosure Page 21 of 23 Figure 14 - RPV Closure Head Vent Lines 706-3/4-603

.l r, I" BSEP 06-0074 Enclosure Page 22 of 23 Figure 15 - Reactor Water Cleanup (RWCU)

DRYWELL PENETRATION X-14 EL. 62'-9 AZ. 2450 LF004 LF027

[F 1001 1-6-603"-

K 4u--PRIMARY CONTAINMENT G31 Ell I RHR F10-20-6034 F067 "A" LOOP RECIRCULATION SUCTION LINE REACTOR VESSEL

(

0203

-o 050 16-2-603 G31 B

N15 26-2--600 F030 F030 32d

BSEP 06-0074 Enclosure Page 23 of 23 Figure 16 - Standby Liquid Control (SLC)

DRYWELL PEN. X-42 EL. 71'-0" - AZ. 2100 11/2"

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