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Entergy Nuclear Operations, Inc. Palisades Nuclear Plant 27780 Blue Star Memorial Highway Covert. MI 49043 Tel 269 764 2000 Otto W. Gustafson Licensing Manager PNP-2012-044 May 1,2012 10 CFR 50.55a U. S. Nuclear Regulatory Commission A TIN: Document Control Desk Washington, DC 20555-0001  
Entergy Nuclear Operations, Inc.
Palisades Nuclear Plant
~Entergy                                                        27780 Blue Star Memorial Highway Covert. MI 49043 Tel 269 764 2000 Otto W. Gustafson Licensing Manager PNP-2012-044 May 1,2012                                             10 CFR 50.55a U. S. Nuclear Regulatory Commission ATIN: Document Control Desk Washington, DC 20555-0001


==Subject:==
==Subject:==
Reply to Request for Additional Information Re: Relief Request -Proposed Alternative  
Reply to Request for Additional Information Re: Relief Request -
-Use of Alternate ASME Code Case N-770-1 Baseline Examination Palisades Nuclear Plant Docket 50-255 License No. DPR-20  
Proposed Alternative - Use of Alternate ASME Code Case N-770-1 Baseline Examination Palisades Nuclear Plant Docket 50-255 License No. DPR-20


==References:==
==References:==
: 1. Entergy Nuclear Operations Inc. letter, "Relief Request -Proposed Alternative  
: 1.     Entergy Nuclear Operations Inc. letter, "Relief Request - Proposed Alternative - Use of Alternate ASME Code Case N-770-1 Baseline Examination," dated April 26, 2012
-Use of Alternate ASME Code Case N-770-1 Baseline Examination," dated April 26, 2012 2. 3. 4.  
: 2. NRC Electronic Request, "Palisades - Relief Request - Proposed Alternative - Use of Alternate ASME Code Case N-770-1 Baseline Examination - ME8492," Request for Additional Information, dated April 27, 2012
: 3. NRC Electronic Request, "Palisades - Relief Request - Proposed Alternative - Use of Alternate ASME Code Case N-770-1 Baseline Examination - ME8492," Request for Additional Information, dated April 30, 2012
: 4. Entergy Nuclear Operations Inc. letter, "Reply to Request for Additional Information Re: Relief Request - Proposed Alternative -
Use of Alternate ASME Code Case N-770-1 Baseline Examination," dated April 30, 2012


==Dear Sir or Madam:==
==Dear Sir or Madam:==
NRC Electronic Request, "Palisades
 
-Relief Request -Proposed Alternative
Entergy Nuclear Operations, Inc. (ENO) submitted a request for relief from ASME Code Case N-770-1 on April 26, 2012 (Reference 1). ENO received electronic requests
-Use of Alternate ASME Code Case N-770-1 Baseline Examination
 
-ME8492," Request for Additional Information, dated April 27, 2012 NRC Electronic Request, "Palisades
PNP 2012-044 Page 2 for additional information (RAI) from the Nuclear Regulatory Commission (NRC) concerning the submittal on April 27, 2012 (Reference 2), and on April 30, 2012 (Reference 3). ENO responded to the RAI dated April 27, 2012, on April 30, 2012 (Reference 4).
-Relief Request -Proposed Alternative
The attachment contains the ENO response to the RAI dated April 30, 2012.
-Use of Alternate ASME Code Case N-770-1 Baseline Examination
This letter contains no new commitments and no revisions to existing commitments.
-ME8492," Request for Additional Information, dated April 30, 2012 Entergy Nuclear Operations Inc. letter, "Reply to Request for Additional Information Re: Relief Request -Proposed Alternative
OWG/jse : Reply to Request for Additional Information Re: Relief Request -
-Use of Alternate ASME Code Case N-770-1 Baseline Examination," dated April 30, 2012 Entergy Nuclear Operations, Inc. (ENO) submitted a request for relief from ASME Code Case N-770-1 on April 26, 2012 (Reference 1). ENO received electronic requests PNP 2012-044 Page 2 for additional information (RAI) from the Nuclear Regulatory Commission (NRC) concerning the submittal on April 27, 2012 (Reference 2), and on April 30, 2012 (Reference 3). ENO responded to the RAI dated April 27, 2012, on April 30, 2012 (Reference 4). The attachment contains the ENO response to the RAI dated April 30, 2012. This letter contains no new commitments and no revisions to existing commitments.
Proposed Alternative - Use of Alternate ASME Code Case N-770-1 Baseline Examination, dated April 30, 2012 cc: Administrator, Region III, USNRC Project Manager, Palisades, USNRC Resident Inspector, Palisades, USNRC
OWG/jse Attachment 1: Reply to Request for Additional Information Re: Relief Request -Proposed Alternative  
 
-Use of Alternate ASME Code Case N-770-1 Baseline Examination, dated April 30, 2012 cc: Administrator, Region III, USNRC Project Manager, Palisades, USNRC Resident Inspector, Palisades, USNRC ATTACHMENT 1 REPLY TO REQUEST FOR ADDITIONAL INFORMATION RE: RELIEF REQUEST -PROPOSED ALTERNATIVE  
ATTACHMENT 1 REPLY TO REQUEST FOR ADDITIONAL INFORMATION RE: RELIEF REQUEST -
-USE OF ALTERNATIVE ASME CODE CASE N-770-1 BASELINE EXAMINATION, DATED APRIL 30,2012 Background By letter dated April 26, 2012, Agencywide Documents Access and Management System (ADAMS) Accession Number ML 121118A 144, Entergy Nuclear Operations, Inc., (the licensee) submitted the proposed alternative "Use of Alternate ASME Code Case N-770-1 Baseline Examination" for U.S. Nuclear Regulatory Commission review and acceptance.
PROPOSED ALTERNATIVE - USE OF ALTERNATIVE ASME CODE CASE N-770-1 BASELINE EXAMINATION, DATED APRIL 30,2012
The request proposed an alternative to the requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code) Case N-770-1, as required and conditioned by Title 10 of the Code of Federal Regulations, Part SO (10 CFR SO) paragraph SSa(g)(6)(ii)(F)(3).
 
Specifically, the licensee states that obtaining the required examination coverage of essentially 100 percent is unattainable due to limitations imposed by design and geometry, and requests relief from the required baseline examination coverage on the basis that compliance with the requirements would be a hardship without a compensating increase in the level of quality or safety, per 10 CFR SO.SSa(a)(3)(ii).
===Background===
In order to complete our review, the Piping and NDE Branch requests additional information.
By letter dated April 26, 2012, Agencywide Documents Access and Management System (ADAMS) Accession Number ML 121118A 144, Entergy Nuclear Operations, Inc., (the licensee) submitted the proposed alternative "Use of Alternate ASME Code Case N-770-1 Baseline Examination" for U.S. Nuclear Regulatory Commission review and acceptance. The request proposed an alternative to the requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code) Case N-770-1, as required and conditioned by Title 10 of the Code of Federal Regulations, Part SO (10 CFR SO) paragraph SSa(g)(6)(ii)(F)(3). Specifically, the licensee states that obtaining the required examination coverage of essentially 100 percent is unattainable due to limitations imposed by design and geometry, and requests relief from the required baseline examination coverage on the basis that compliance with the requirements would be a hardship without a compensating increase in the level of quality or safety, per 10 CFR SO.SSa(a)(3)(ii). In order to complete our review, the Piping and NDE Branch requests additional information.
Nuclear Regulatory Commission (NRC) Request 1. In the event that an axial leak were to occur in the PORV 4" DM Weld, please provide a response to the fol/owing:
Nuclear Regulatory Commission (NRC) Request
: a. What would be the effect of an axial flaw through the susceptible material of the DM weld b. Describe the licensee's ability to detect leakage from an axial flaw in this location c. What actions would be available to mitigate the effect of an axial flaw causing leakage at this location on the reactor coolant system Entergy Nuclear Operations, Inc. (END) Response a. Due to the configuration of the susceptible material, an axial flaw could propagate, resulting in a detectable through-wall leak. The total axial extent of the susceptible material is only approximately  
: 1. In the event that an axial leak were to occur in the PORV 4" DM Weld, please provide a response to the fol/owing:
% inch, which limits the length of an axial crack and the possibility of a structural failure. Page 1 of 4 Due to the susceptible material location, the leakage would originate from the steam-filled portion of the pressurizer, and would include only a small amount of entrained boric acid. Discharged steam would condense on cooler surfaces as it depressurized to atmospheric pressure.
: a. What would be the effect of an axial flaw through the susceptible material of the DM weld
As a result, condensation and boric acid deposits would be visible at seams in insulation sheathing.
: b. Describe the licensee's ability to detect leakage from an axial flaw in this location
Leakage would have minimal impact on adjacent equipment within the pressurizer shed because the pressurizer shed fans ventilate the upper pressurizer area, dispersing water vapor to the containment atmosphere.
: c. What actions would be available to mitigate the effect of an axial flaw causing leakage at this location on the reactor coolant system Entergy Nuclear Operations, Inc. (END) Response
: a. Due to the configuration of the susceptible material, an axial flaw could propagate, resulting in a detectable through-wall leak. The total axial extent of the susceptible material is only approximately % inch, which limits the length of an axial crack and the possibility of a structural failure.
Page 1 of 4
 
Due to the susceptible material location, the leakage would originate from the steam-filled portion of the pressurizer, and would include only a small amount of entrained boric acid. Discharged steam would condense on cooler surfaces as it depressurized to atmospheric pressure. As a result, condensation and boric acid deposits would be visible at seams in insulation sheathing. Leakage would have minimal impact on adjacent equipment within the pressurizer shed because the pressurizer shed fans ventilate the upper pressurizer area, dispersing water vapor to the containment atmosphere.
: b. Palisades has various means of detecting leakage from an axial flaw in this location.
: b. Palisades has various means of detecting leakage from an axial flaw in this location.
A small steam leak from an axial flaw in the power operated relief valve (PORV) piping would, over time, result in a rise in containment sump level rate of increase.
A small steam leak from an axial flaw in the power operated relief valve (PORV) piping would, over time, result in a rise in containment sump level rate of increase. Containment sump level is continually monitored, and if a rise in the rate of containment sump level increase is observed, plant procedures direct plant operators to identify the source of the leakage.
Containment sump level is continually monitored, and if a rise in the rate of containment sump level increase is observed, plant procedures direct plant operators to identify the source of the leakage. Operators may also be alerted to a leak from an axial flaw by containment radiation monitoring instrumentation.
Operators may also be alerted to a leak from an axial flaw by containment radiation monitoring instrumentation. This instrumentation, required by the Technical Specifications, is capable of detecting a 100 cm 9/min leak in 45 minutes, based on 1% failed fuel.
This instrumentation, required by the Technical Specifications, is capable of detecting a 100 cm 9/min leak in 45 minutes, based on 1 % failed fuel. The primary coolant system (PCS) is inspected for leaks as the plant is shut down for refueling outages. After refueling, as the plant is returned to power operations, VT -2 visual examinations are performed to detect leakage from the PCS. Operator walkdowns of containment are periodically performed during power operations at lower levels of containment to detect leakage. c. In the event of an axial flaw causing leakage at this location in the PCS, the plant would be shut down and placed in a safe condition in accordance with plant procedures.
The primary coolant system (PCS) is inspected for leaks as the plant is shut down for refueling outages. After refueling, as the plant is returned to power operations, VT -2 visual examinations are performed to detect leakage from the PCS. Operator walkdowns of containment are periodically performed during power operations at lower levels of containment to detect leakage.
NRC Request 2. Provide the basis for hardship to perform an eddy current examination for the PORV 4" OM Weld. Page 2 of 4 ENO Response Eddy current testing (ECT) of the PORV weld would require special tooling, a mockup, procedure development, and qualification of the process, which would take significant time to develop and implement.
: c. In the event of an axial flaw causing leakage at this location in the PCS, the plant would be shut down and placed in a safe condition in accordance with plant procedures.
The PORV nozzle is not directly accessible from the pressurizer manway, because the two openings are on opposite sides of the pressurizer.
NRC Request
The spray head is attached to the underside of the pressurizer head at the topmost location, and would be an obstruction to reaching the nozzle from the manway. In order to access the PORV nozzle without physically entering the pressurizer, an articulated tool would have to be designed.
: 2. Provide the basis for hardship to perform an eddy current examination for the PORV 4" OM Weld.
The tool would be inserted through the manway, manipulated past the spray head, and inserted up into the PORV nozzle. Closed circuit television would likely be needed. The equipment would have to be sufficiently robust to prevent loss of foreign material into the heater region, because dropped equipment could damage installed pressurizer heaters. The dose rate at the pressurizer manway is approximately 800 mrem/hr, and would result in significant worker exposure.
Page 2 of 4
Direct personnel access to the PORV nozzle from inside the pressurizer would require installation of an internal work platform, probably suspended from the manway. This would introduce a significant added risk of foreign material resulting in pressurizer heater damage. Dose rates inside the pressurizer would result in considerable worker exposure, and worker heat stress would be high. NRC Request 3. Provide the basis for hardship for performing addition 10 surface examination for each of the welds covered by this relief request. ENO Response Inside diameter (ID) surface examination would require access to the inside diameter of the pipe. This will create considerable hardship in dose, extended outage time, and undue risk to installed equipment.
 
Intemal surface examination options include liquid penetrant testing and eddy current testing. Liquid penetrant testing requires the pipe interior to be dry. Drying time for cold leg attachment welds would be unreasonably long due to high humidity in the piping system. Eddy current testing would require special tooling, mockup, Page 3 of 4 procedure development, and qualification of the process. This would unduly extend the current refueling outage. PORV weld access from the pressurizer side: PORV access hardship is discussed in the answer to the previous question. Drain nozzle weld access from the PCS cold leg side: Access from the steam generator would require drainage of the PCS to mid-loop in order to open the steam generator manway. This is a high risk activity.
ENO Response Eddy current testing (ECT) of the PORV weld would require special tooling, a mockup, procedure development, and qualification of the process, which would take significant time to develop and implement.
The PORV nozzle is not directly accessible from the pressurizer manway, because the two openings are on opposite sides of the pressurizer. The spray head is attached to the underside of the pressurizer head at the topmost location, and would be an obstruction to reaching the nozzle from the manway.
In order to access the PORV nozzle without physically entering the pressurizer, an articulated tool would have to be designed. The tool would be inserted through the manway, manipulated past the spray head, and inserted up into the PORV nozzle. Closed circuit television would likely be needed. The equipment would have to be sufficiently robust to prevent loss of foreign material into the heater region, because dropped equipment could damage installed pressurizer heaters. The dose rate at the pressurizer manway is approximately 800 mrem/hr, and would result in significant worker exposure.
Direct personnel access to the PORV nozzle from inside the pressurizer would require installation of an internal work platform, probably suspended from the manway. This would introduce a significant added risk of foreign material resulting in pressurizer heater damage. Dose rates inside the pressurizer would result in considerable worker exposure, and worker heat stress would be high.
NRC Request
: 3. Provide the basis for hardship for performing addition 10 surface examination for each of the welds covered by this relief request.
ENO Response Inside diameter (ID) surface examination would require access to the inside diameter of the pipe. This will create considerable hardship in dose, extended outage time, and undue risk to installed equipment.
Intemal surface examination options include liquid penetrant testing and eddy current testing. Liquid penetrant testing requires the pipe interior to be dry. Drying time for cold leg attachment welds would be unreasonably long due to high humidity in the piping system. Eddy current testing would require special tooling, mockup, Page 3 of 4
 
procedure development, and qualification of the process. This would unduly extend the current refueling outage.
PORV weld access from the pressurizer side: PORV access hardship is discussed in the answer to the previous question.
Drain nozzle weld access from the PCS cold leg side: Access from the steam generator would require drainage of the PCS to mid-loop in order to open the steam generator manway. This is a high risk activity. Special tooling and qualification would be required to perform this inspection.
Charging and spray nozzle weld access from the PCS cold leg side: Access from the reactor vessel would require full core offload and core barrel removal. This is a high risk, high dose activity. Special tooling and qualification would be required to perform this inspection.
Access from the steam generator would require drainage of the PCS to mid-loop in order to open the steam generator manway. This is a high risk activity. The inspection device would have to be threaded through the reactor coolant pump.
Special tooling and qualification would be required to perform this inspection.
Special tooling and qualification would be required to perform this inspection.
Charging and spray nozzle weld access from the PCS cold leg side: Access from the reactor vessel would require full core offload and core barrel removal. This is a high risk, high dose activity.
Weld access from the attached piping side: In all cases, piping segments would have to be cut out to allow tool access, and the piping segments would have to be reinstalled after the examinations. Cold leg locations would require plant conditions ranging from reduced inventory to full core offload.
Special tooling and qualification would be required to perform this inspection.
NRC Request
Access from the steam generator would require drainage of the PCS to mid-loop in order to open the steam generator manway. This is a high risk activity.
: 4. Provide the location and size of any PWSCC flaws found in any OM pressurizer, hot leg and cold leg weld temperature locations that would be covered by ASME Code Case N-770-1.
The inspection device would have to be threaded through the reactor coolant pump. Special tooling and qualification would be required to perform this inspection.
END Response ENO did not find any PWSCC flaws in any dissimilar metal (OM) pressurizer, hot leg, or cold leg weld locations covered by ASME Code Case N-770-1.
Weld access from the attached piping side: In all cases, piping segments would have to be cut out to allow tool access, and the piping segments would have to be reinstalled after the examinations.
Page 4 of 4
Cold leg locations would require plant conditions ranging from reduced inventory to full core offload. NRC Request 4. Provide the location and size of any PWSCC flaws found in any OM pressurizer, hot leg and cold leg weld temperature locations that would be covered by ASME Code Case N-770-1. END Response ENO did not find any PWSCC flaws in any dissimilar metal (OM) pressurizer, hot leg, or cold leg weld locations covered by ASME Code Case N-770-1. Page 4 of 4 ENCLOSURE The NRC previously issued an RAI for this relief request on April 27, 2012, and ENO responded to the RAI on April 30, 2012. Enclosure 1 for this RAI response contained a table entitled, "Alloy 600 Susceptible Volume Calculation -1 R22." This table contained a discrepancy.
 
For welds PCS-3-PSS-1 81-1/77 and PCS-3-PSS-2A1-1/27S, the inner 1/3T coverage, circumferential scan for axial flaws, 8 inches of the 11 inch circumference were scanned rather than 7 inches of the 11 inch circumference.
ENCLOSURE The NRC previously issued an RAI for this relief request on April 27, 2012, and ENO responded to the RAI on April 30, 2012. Enclosure 1 for this RAI response contained a table entitled, "Alloy 600 Susceptible Volume Calculation - 1R22."
This discrepancy does not affect the computed coverage percentage as the 72.8% coverage percentage in the table was based on coverage of 8 inches of the 11 inch circumference.
This table contained a discrepancy. For welds PCS-3-PSS-1 81-1/77 and PCS-3-PSS-2A1-1/27S, the inner 1/3T coverage, circumferential scan for axial flaws, 8 inches of the 11 inch circumference were scanned rather than 7 inches of the 11 inch circumference.
The discrepancy is corrected in the attached table. 1 Page Follows Alloy 600 Susceptible Volume Calculation  
This discrepancy does not affect the computed coverage percentage as the 72.8%
-lR22 Inner 1/3T Coverage Outer 2/3T Coverage Weld Axial Scan for Cire Cire Scan for Axial Combined Axial Scan for Cire Sean for Axial Flaws Flaws Coverage Cire Flaws Flaws Notes PC5-4-PR5-1P1-1/
coverage percentage in the table was based on coverage of 8 inches of the 11 inch circumference.
100% 0% 50% 100% 0% 165 (PORV) PC5-2-DRL-2A1-1/
The discrepancy is corrected in the attached table.
100% 62.5% 81.3% 85% 66% 273 415-40.6%
1 Page Follows
100% 60L-100% 43L-0% PC5-2-CHL-1A1-17 68L-100% Combined -20.3% (100% volume 43.4% 92% /274 605-0% (100% volume scanned, with Combined -66.6% credit for scanned, no credit unqualified 43L) for unqualified 43L) 63.6% 72.8% PC5-3-P55-1B1-1/
 
68.2% 60% 73% 77 (100% scanned 7" of (100% scanned 8" of 11" circumference) 11" circumference) 63.6% 72.8% , I PC5-3-P55-2A1-1/
Alloy 600 Susceptible Volume Calculation - lR22 Inner 1/3T Coverage                               Outer 2/3T Coverage Weld         Axial Scan for Cire     Cire Scan for Axial Combined Axial Scan for Cire Sean for Axial Notes Flaws                   Flaws       Coverage Cire Flaws           Flaws PC5-4-PR5-1P1-1/
275 (100% scanned 7" of (100% scanned 8" of 68.2% 60% 73% ---circumferen&
100%                     0%             50%       100%               0%
11" circumference)}}
165 (PORV)
PC5-2-DRL-2A1-1/
100%                     62.5%         81.3%       85%               66%
273 415-40.6%
100%
43L-0%
60L-100%
Combined - 20.3%
PC5-2-CHL-1A1-17       68L-100%                                                           (100% volume 43.4%       92%
      /274             605-0%                                                             scanned, with (100% volume Combined - 66.6%                                                           credit for scanned, no credit unqualified 43L) for unqualified 43L) 63.6%                   72.8%
PC5-3-P55-1B1-1/
68.2%       60%               73%
77       (100% scanned 7" of     (100% scanned 8" of 11" circumference)     11" circumference)
                                                                                                                  ,
63.6%                   72.8%
PC5-3-P55-2A1-1/                                                                                                   I 68.2%      60%                73%
275       (100% scanned 7" of     (100% scanned 8" of
-         --
                ~1" circumferen&           11" circumference)}}

Revision as of 04:09, 12 November 2019

Reply to Request for Additional Information Relief Request - Proposed Alternative - Use of Alternate ASME Code Case N-770-1 Baseline Examination
ML12123A079
Person / Time
Site: Palisades Entergy icon.png
Issue date: 05/01/2012
From: Gustafson O
Entergy Nuclear Operations
To:
Office of Nuclear Reactor Regulation, Document Control Desk
References
PNP-2012-044, TAC ME8492
Download: ML12123A079 (8)
v • d • e


Text

'"

Entergy Nuclear Operations, Inc.

Palisades Nuclear Plant

~Entergy 27780 Blue Star Memorial Highway Covert. MI 49043 Tel 269 764 2000 Otto W. Gustafson Licensing Manager PNP-2012-044 May 1,2012 10 CFR 50.55a U. S. Nuclear Regulatory Commission ATIN: Document Control Desk Washington, DC 20555-0001

Subject:

Reply to Request for Additional Information Re: Relief Request -

Proposed Alternative - Use of Alternate ASME Code Case N-770-1 Baseline Examination Palisades Nuclear Plant Docket 50-255 License No. DPR-20

References:

1. Entergy Nuclear Operations Inc. letter, "Relief Request - Proposed Alternative - Use of Alternate ASME Code Case N-770-1 Baseline Examination," dated April 26, 2012
2. NRC Electronic Request, "Palisades - Relief Request - Proposed Alternative - Use of Alternate ASME Code Case N-770-1 Baseline Examination - ME8492," Request for Additional Information, dated April 27, 2012
3. NRC Electronic Request, "Palisades - Relief Request - Proposed Alternative - Use of Alternate ASME Code Case N-770-1 Baseline Examination - ME8492," Request for Additional Information, dated April 30, 2012
4. Entergy Nuclear Operations Inc. letter, "Reply to Request for Additional Information Re: Relief Request - Proposed Alternative -

Use of Alternate ASME Code Case N-770-1 Baseline Examination," dated April 30, 2012

Dear Sir or Madam:

Entergy Nuclear Operations, Inc. (ENO) submitted a request for relief from ASME Code Case N-770-1 on April 26, 2012 (Reference 1). ENO received electronic requests

PNP 2012-044 Page 2 for additional information (RAI) from the Nuclear Regulatory Commission (NRC) concerning the submittal on April 27, 2012 (Reference 2), and on April 30, 2012 (Reference 3). ENO responded to the RAI dated April 27, 2012, on April 30, 2012 (Reference 4).

The attachment contains the ENO response to the RAI dated April 30, 2012.

This letter contains no new commitments and no revisions to existing commitments.

OWG/jse : Reply to Request for Additional Information Re: Relief Request -

Proposed Alternative - Use of Alternate ASME Code Case N-770-1 Baseline Examination, dated April 30, 2012 cc: Administrator, Region III, USNRC Project Manager, Palisades, USNRC Resident Inspector, Palisades, USNRC

ATTACHMENT 1 REPLY TO REQUEST FOR ADDITIONAL INFORMATION RE: RELIEF REQUEST -

PROPOSED ALTERNATIVE - USE OF ALTERNATIVE ASME CODE CASE N-770-1 BASELINE EXAMINATION, DATED APRIL 30,2012

Background

By letter dated April 26, 2012, Agencywide Documents Access and Management System (ADAMS) Accession Number ML 121118A 144, Entergy Nuclear Operations, Inc., (the licensee) submitted the proposed alternative "Use of Alternate ASME Code Case N-770-1 Baseline Examination" for U.S. Nuclear Regulatory Commission review and acceptance. The request proposed an alternative to the requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code) Case N-770-1, as required and conditioned by Title 10 of the Code of Federal Regulations, Part SO (10 CFR SO) paragraph SSa(g)(6)(ii)(F)(3). Specifically, the licensee states that obtaining the required examination coverage of essentially 100 percent is unattainable due to limitations imposed by design and geometry, and requests relief from the required baseline examination coverage on the basis that compliance with the requirements would be a hardship without a compensating increase in the level of quality or safety, per 10 CFR SO.SSa(a)(3)(ii). In order to complete our review, the Piping and NDE Branch requests additional information.

Nuclear Regulatory Commission (NRC) Request

1. In the event that an axial leak were to occur in the PORV 4" DM Weld, please provide a response to the fol/owing:
a. What would be the effect of an axial flaw through the susceptible material of the DM weld
b. Describe the licensee's ability to detect leakage from an axial flaw in this location
c. What actions would be available to mitigate the effect of an axial flaw causing leakage at this location on the reactor coolant system Entergy Nuclear Operations, Inc. (END) Response
a. Due to the configuration of the susceptible material, an axial flaw could propagate, resulting in a detectable through-wall leak. The total axial extent of the susceptible material is only approximately % inch, which limits the length of an axial crack and the possibility of a structural failure.

Page 1 of 4

Due to the susceptible material location, the leakage would originate from the steam-filled portion of the pressurizer, and would include only a small amount of entrained boric acid. Discharged steam would condense on cooler surfaces as it depressurized to atmospheric pressure. As a result, condensation and boric acid deposits would be visible at seams in insulation sheathing. Leakage would have minimal impact on adjacent equipment within the pressurizer shed because the pressurizer shed fans ventilate the upper pressurizer area, dispersing water vapor to the containment atmosphere.

b. Palisades has various means of detecting leakage from an axial flaw in this location.

A small steam leak from an axial flaw in the power operated relief valve (PORV) piping would, over time, result in a rise in containment sump level rate of increase. Containment sump level is continually monitored, and if a rise in the rate of containment sump level increase is observed, plant procedures direct plant operators to identify the source of the leakage.

Operators may also be alerted to a leak from an axial flaw by containment radiation monitoring instrumentation. This instrumentation, required by the Technical Specifications, is capable of detecting a 100 cm 9/min leak in 45 minutes, based on 1% failed fuel.

The primary coolant system (PCS) is inspected for leaks as the plant is shut down for refueling outages. After refueling, as the plant is returned to power operations, VT -2 visual examinations are performed to detect leakage from the PCS. Operator walkdowns of containment are periodically performed during power operations at lower levels of containment to detect leakage.

c. In the event of an axial flaw causing leakage at this location in the PCS, the plant would be shut down and placed in a safe condition in accordance with plant procedures.

NRC Request

2. Provide the basis for hardship to perform an eddy current examination for the PORV 4" OM Weld.

Page 2 of 4

ENO Response Eddy current testing (ECT) of the PORV weld would require special tooling, a mockup, procedure development, and qualification of the process, which would take significant time to develop and implement.

The PORV nozzle is not directly accessible from the pressurizer manway, because the two openings are on opposite sides of the pressurizer. The spray head is attached to the underside of the pressurizer head at the topmost location, and would be an obstruction to reaching the nozzle from the manway.

In order to access the PORV nozzle without physically entering the pressurizer, an articulated tool would have to be designed. The tool would be inserted through the manway, manipulated past the spray head, and inserted up into the PORV nozzle. Closed circuit television would likely be needed. The equipment would have to be sufficiently robust to prevent loss of foreign material into the heater region, because dropped equipment could damage installed pressurizer heaters. The dose rate at the pressurizer manway is approximately 800 mrem/hr, and would result in significant worker exposure.

Direct personnel access to the PORV nozzle from inside the pressurizer would require installation of an internal work platform, probably suspended from the manway. This would introduce a significant added risk of foreign material resulting in pressurizer heater damage. Dose rates inside the pressurizer would result in considerable worker exposure, and worker heat stress would be high.

NRC Request

3. Provide the basis for hardship for performing addition 10 surface examination for each of the welds covered by this relief request.

ENO Response Inside diameter (ID) surface examination would require access to the inside diameter of the pipe. This will create considerable hardship in dose, extended outage time, and undue risk to installed equipment.

Intemal surface examination options include liquid penetrant testing and eddy current testing. Liquid penetrant testing requires the pipe interior to be dry. Drying time for cold leg attachment welds would be unreasonably long due to high humidity in the piping system. Eddy current testing would require special tooling, mockup, Page 3 of 4

procedure development, and qualification of the process. This would unduly extend the current refueling outage.

PORV weld access from the pressurizer side: PORV access hardship is discussed in the answer to the previous question.

Drain nozzle weld access from the PCS cold leg side: Access from the steam generator would require drainage of the PCS to mid-loop in order to open the steam generator manway. This is a high risk activity. Special tooling and qualification would be required to perform this inspection.

Charging and spray nozzle weld access from the PCS cold leg side: Access from the reactor vessel would require full core offload and core barrel removal. This is a high risk, high dose activity. Special tooling and qualification would be required to perform this inspection.

Access from the steam generator would require drainage of the PCS to mid-loop in order to open the steam generator manway. This is a high risk activity. The inspection device would have to be threaded through the reactor coolant pump.

Special tooling and qualification would be required to perform this inspection.

Weld access from the attached piping side: In all cases, piping segments would have to be cut out to allow tool access, and the piping segments would have to be reinstalled after the examinations. Cold leg locations would require plant conditions ranging from reduced inventory to full core offload.

NRC Request

4. Provide the location and size of any PWSCC flaws found in any OM pressurizer, hot leg and cold leg weld temperature locations that would be covered by ASME Code Case N-770-1.

END Response ENO did not find any PWSCC flaws in any dissimilar metal (OM) pressurizer, hot leg, or cold leg weld locations covered by ASME Code Case N-770-1.

Page 4 of 4

ENCLOSURE The NRC previously issued an RAI for this relief request on April 27, 2012, and ENO responded to the RAI on April 30, 2012. Enclosure 1 for this RAI response contained a table entitled, "Alloy 600 Susceptible Volume Calculation - 1R22."

This table contained a discrepancy. For welds PCS-3-PSS-1 81-1/77 and PCS-3-PSS-2A1-1/27S, the inner 1/3T coverage, circumferential scan for axial flaws, 8 inches of the 11 inch circumference were scanned rather than 7 inches of the 11 inch circumference.

This discrepancy does not affect the computed coverage percentage as the 72.8%

coverage percentage in the table was based on coverage of 8 inches of the 11 inch circumference.

The discrepancy is corrected in the attached table.

1 Page Follows

Alloy 600 Susceptible Volume Calculation - lR22 Inner 1/3T Coverage Outer 2/3T Coverage Weld Axial Scan for Cire Cire Scan for Axial Combined Axial Scan for Cire Sean for Axial Notes Flaws Flaws Coverage Cire Flaws Flaws PC5-4-PR5-1P1-1/

100% 0% 50% 100% 0%

165 (PORV)

PC5-2-DRL-2A1-1/

100% 62.5% 81.3% 85% 66%

273 415-40.6%

100%

43L-0%

60L-100%

Combined - 20.3%

PC5-2-CHL-1A1-17 68L-100% (100% volume 43.4% 92%

/274 605-0% scanned, with (100% volume Combined - 66.6% credit for scanned, no credit unqualified 43L) for unqualified 43L) 63.6% 72.8%

PC5-3-P55-1B1-1/

68.2% 60% 73%

77 (100% scanned 7" of (100% scanned 8" of 11" circumference) 11" circumference)

,

63.6% 72.8%

PC5-3-P55-2A1-1/ I 68.2% 60% 73%

275 (100% scanned 7" of (100% scanned 8" of

- --

~1" circumferen& 11" circumference)

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