View source for LR-N10-0165, Response to NRC Request for Additional Information Dated April 15, 2010, Related to Structures and Structures-Related Aging Management Programs for License Renewal Application

{{Adams
| number = ML101390184
| issue date = 05/13/2010
| title = Response to NRC Request for Additional Information Dated April 15, 2010, Related to Structures and Structures-Related Aging Management Programs for License Renewal Application
| author name = Davison P J
| author affiliation = PSEG Nuclear, LLC
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000272, 05000311
| license number = DPR-070, DPR-075
| contact person = 
| case reference number = LR-N10-0165
| document type = Letter type:LR
| page count = 358
| project = 
| stage = Response to RAI
}}

=Text=
{{#Wiki_filter:PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, NJ 08038 0 P LEL Nuclear LLC MAY 13 2010 10 CFR 50 10 CFR 51 10 CFR 54 LR-N10-0165 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Salem Nuclear Generating Station, Unit No. 1 and Unit No. 2 Facility Operating License Nos. DPR-70 and DPR-75 NRC Docket Nos. 50-272 and 50-311 

==Subject:==

==Reference:==

Response to NRC Request for Additional Information dated April 15, 2010, Related to Structures and Structures-Related Aging Management Programs for the Salem Nuclear Generating Station, Units 1 and 2 License Renewal Application Letter from Mr. Donnie Ashley (USNRC) to Mr. Thomas Joyce (PSEG Nuclear, LLC) "REQUEST FOR ADDITIONAL INFORMATION REGARDING ASME SECTION Xl, SUBSECTION IWE FOR THE SALEM NUCLEAR GENERATING STATION UNITS 1 AND 2 LICENSE RENEWAL APPLICATION (TAC ME1836 AND ME1834)", dated April 15, 2010 In the referenced letter, the NRC requested additional information related to certain Structures and Structures-Related Aging Management Programs associated with the Salem Nuclear Generating Station, Units 1 and 2 (Salem) License Renewal Application (LRA). Enclosure A contains the responses to the requests for additional information (RAI). Enclosure B contains updates to LRA Appendix A (UFSAR Supplement) and Appendix B Program descriptions that are affected by these RAI responses.
Enclosure C contains an update to the Salem License Renewal Commitment List, which details several changes being made to PSEG Nuclear's commitments associated with the Salem License Renewal Application.
If you have any questions regarding this submittal, please contact Mr. Ali Fakhar, PSEG Manager -License Renewal, at 856-339-1646.
/4047 LR-N10-0165 Page 2 MAY 13 2010 I declare under penalty of perjury that the foregoing is true and correct.Executed on S 10  Sincerely, Paul J. Davison Vice President, Operations Support PSEG Nuclear LLC 

==Enclosures:==

A. Responses to Request for Additional Information B. Updates to LRA Appendix A and Appendix B C. Updates to License Renewal Commitment List D. MPR Associates Report MPR-2613, Revision 3 cc: S. Collins, Regional Administrator 
-USNRC Region I B. Brady, Project Manager, License Renewal -USNRC R. Ennis, Project Manager -USNRC NRC Senior Resident Inspector 
-Salem P. Mulligan, Manager IV, NJBNE L. Marabella, Corporate Commitment Tracking Coordinator Howard Berrick, Salem Commitment Tracking Coordinator LR-N10-0165 Enclosure A Enclosure A Responses to Request for Additional Information related to Structures and Structures-related Aging Management Programs for the Salem Nuclear Generating Station, Units 1 and 2 License Renewal Application (LRA)RAI B.2.1.28-1 RAI B.2.1.28-2 RAI B.2.1.29-1 RAI B.2.1.29-2 RAI B.2.1.33-1 RAI B.2.1.33-2 RAI B.2.1.33-3 RAI B.2.1.33-4 Page 1 of 38 LR-N10-0165 Enclosure A RAI B.2.1.28-1

===Background===
GALL Report (NUREG-1801), AMP XI.S1, "ASME Section XI, Subsection IWE," Program Element 10 states that implementation of ASME Section Xl, Subsection IWE, in accordance with 10 CFR 50.55a, is a necessary element of aging management for steel components of steel and concrete containments through the period of extended operation.
Issue Program Element 10 for the Salem ASME Section Xl, Subsection IWE aging management program discusses operating experience related to containment steel liner plate corrosion as described in NRC Information Notices IN 97-10 and IN 2004-09.However, Program Element 10 for the Salem ASME Section XI, Subsection IWE aging management program does not discuss operating experience related to liner plate corrosion recently reported at Beaver Valley. In addition, a review of the operating experience of the Salem Unit 1 (PIRS # 950706252-78) in 1995, (Notification 
#20344017) in 2007, and Unit 2 (Notification 
#20235636) in 2005 indicate that borated water was running down the containment liner plate behind the insulation which resulted in indications of corrosion of the containment liner plate and seepage of water into moisture barrier. According to Notification 
# 20344017, borated water has been leaking in one area of containment for last 30 years.Request 1. Provide details of borated water leakage, if any, observed inside Units 1 and 2 containments during the most recent refueling outages.2. Explain why augmented inspection of the Unit 2 liner plate and moisture barrier was not performed in successive inspection intervals as required by IWE-1242 since 1995. According to IWE-1242, augmented inspection is required of areas exposed to standing water, repeated wetting and drying, and persistent leakage.3. Provide a summary of the liner plate degradation, including loss of liner plate thickness due to corrosion, integrity of leak chase channels and condition of moisture barrier, as observed during the most recent inspections of Unit 1 and 2 containments.
: 4. Provide detailed future plans for determining corrective actions, including commitments and completion schedules, for addressing steel liner plate corrosion and moisture barrier deterioration in Unit 1 and 2 containments.
The staff needs the above information to confirm that the effects of aging of the containment pressure boundary metal will be adequately managed so that it's intended function will be maintained consistent with the current licensing basis for the period of extended operation as required by 10 CFR 54.21(a)(3).
Page 2 of 38 LR-N10-0165 Enclosure A PSEG Response: 1. During the most recent Salem Unit 1 outage, in the spring of 2010, no active leakage from the reactor cavity and fuel transfer canal telltales was observed.During the most recent Salem Unit 2 outage, in the fall of 2009, a 60 drip per minute leak of borated water was observed at the fuel transfer canal telltale, above the door to the Letdown Heat Exchanger Room. Borated water was observed on the containment liner plate moisture barrier under the fuel transfer canal. These leaks were attributed to reactor cavity leakage. The containment liner plate and moisture barrier were examined and found to meet the IWE acceptance criteria.There were other leaks identified during walkdowns of the Salem Unit 1 and Unit 2 Containments as part of the Boric Acid Corrosion program, but these other leaks were small and localized, such as at a pipe cap, and were generally inactive, so that there was no impact on the Containment.
: 2. Salem began implementation of containment inservice inspection (CISI) in accordance with ASME Section Xl, Subsection IWE, as mandated by 10 CFR Part 50.55a in April 2000. Since that time, 100% of accessible surface areas of the Salem Unit 2 containment liner plate was examined each Inspection Period of the 1 st CISI Interval in accordance with IWE-3500.
The IWE program and examinations identified no surface areas of the containment liner plate that require augmented examinations as specified in IWE-1 242. The moisture barrier was considered inaccessible and was not inspected prior to 2009. In 2009, Salem implemented a change to make the moisture barrier accessible, to the extent possible, and performed visual inspections of the moisture barrier in accordance with IWE. The 2009 containment liner plate examinations identified areas that require augmented examination.
These augmented examination areas have been identified for inclusion in the Salem plan for the 2nd CISI Interval, which started in April of 2010.Prior to April 2000, inspection of the containment was performed under the Structures Monitoring Program in accordance with 10 CFR 50.65 and 10 CFR Part 50, Appendix J. Augmented Examination requirements of IWE-1 242 did not apply.3. A summary of the containment liner plate degradation, including loss of containment liner plate thickness due to corrosion, integrity of containment liner leak chase channels and condition of the moisture barrier, as observed during the most recent inspections of the Unit 1 and Unit 2 Containments, is summarized below. This degradation was found as a result of Enhancement 
#2 described in the LRA, Appendix B.2.1.28 (ASME Section Xl, Subsection IWE), as shown on page B-132.This enhancement, which involved trimming the stainless steel lagging at the bottom of the containment liner insulation panels, has been implemented at both Unit 1 and Unit 2. As a result, areas that were previously inaccessible for inspection have been made accessible, examinations have been performed, and evaluations have verified the adequacy of existing conditions.
Additionally, corrective actions to address degraded conditions found during the most recent inspections have been developed.
Page 3 of 38 LR-N10-0165 Enclosure A Unit 1: The following examination results were obtained during the refueling outage in the spring of 2010.Containment Location Examination Results Component Liner plate -3/4" The knuckle plate Some local corrosion was observed in the knuckle plate connects the 1/4" area above the floor. As a result, UT liner plate below readings were taken at 1 foot intervals the floor to the around the containment, except where 1/2" wall liner interferences near the containment liner plate. The plate plate impeded UT testing, for a total of 349 extends from 2 readings.
Minimum thickness measured feet below the after cleaning was 0.721". The thickness concrete floor to utilized in the containment analysis is 1/2".approximately 6" All readings met acceptance criteria for above the floor, loss of material less than 10% of the thickness in the analysis.See UFSAR Fig.3.8-7 Liner plate -1/2" The bottom starts Four containment liner plate insulation plate at the top of the panels were removed where corrosion was knuckle plate and observed in the adjacent accessible extends to areas. Visual examination of the approximately 22 containment liner plate covered by the feet above the insulation panels identified local surface floor, corrosion at each of the four panels and local paint blistering at one of the panels.As a result, 102 UT thickness readings See UFSAR Fig. were taken at the four panels in the area of 3.8-1 and 3.8-7 the surface corrosion.
An additional 18 UT thickness readings were taken in the area of the blistered paint. The minimum thickness from UT measurement results was 0.452". All readings met acceptance criteria for loss of material less than 10%of the nominal thickness.
Vertical leak Extends down All of the accessible vertical leak chase chase channels from the lowest channels were examined.
One channel horizontal leak had corrosion that extended through the chase channel into channel wall (hole). The leak chase and under the channel with the hole was cleaned out and containment floor, the channel and containment liner plate were visually examined with a boroscope beneath the containment floor. Only surface scale was observed and no evidence of significant moisture or loss of material was noted. The channel with the hole was cut at the floor and capped to prevent moisture intrusion.
Page 4 of 38 LR-N1O-0165 Enclosure A Containment Location Examination Results Component Moisture Barrier At the containment 100% of the moisture barrier area was liner plate to inspected and repaired or replaced where concrete floor it did not meet the IWE acceptance interface at criteria.elevation 78'.Unit 2: The following examination results were obtained during the refueling outage in the fall of 2009.Containment Location Examination Results Component Liner plate -3/4" The knuckle plate Some local corrosion was observed in the knuckle plate connects the 1/4" area above the floor. As a result, UT liner plate below readings were taken at 1 foot intervals the floor to the 1/2" around the Containment, except where wall liner plate. interferences near the containment liner The plate extends plate impeded UT testing, for a total of 368 from 2 feet below readings.
Minimum thickness measured the concrete floor before cleaning was 0.677". The thickness to approximately utilized in the containment analysis is 1/2".6" above the floor. All readings met acceptance criteria for loss of material less than 10% of the See UFSAR thickness in the analysis.Fig.3.8-7 Liner plate -1/2" The bottom starts Seven UT measurements were taken on plate at the top of the the 1/2" containment liner plate, just above knuckle plate and the 3/4" knuckle plate due to an extends to about interference at the knuckle plate. The 22 feet above the measured thickness was greater than the floor, nominal thickness of 1/2". The minimum thickness measured was 0.509".See UFSAR Fig.3.8-1 and 3.8-7 In addition, containment liner insulation was removed at four panel locations in suspect areas as described in the response to RAI B.2.1.28-2, item #1.Some local surface corrosion was observed at all four of the panels. Some local paint blistering was observed at one of the panels. A total of 4 UT readings were taken at the one insulation panel with the paint blisters.
After cleaning of indications, minimum thickness from UT results was 0.518". All readings met acceptance criteria for loss of material less than 10% of the nominal thickness.
Page 5 of 38 LR-N`10-0165 Enclosure A Containment Location Examination Results Component Vertical leak Extends down All of the accessible vertical leak chase chase channels from the lowest channels were examined.
Six channels horizontal leak had, corrosion that extended through the chase channel into channel wall (hole). The leak chase and under the channels with the holes were cleaned out containment floor, to the extent possible.
The channel and containment liner plate were visually examined with a boroscope beneath the containment floor. The six channels with the holes were cut at the floor and capped to prevent moisture intrusion.
Moisture Barrier At the containment 100% of the accessible moisture barrier liner plate to was inspected and found to have concrete floor performed its intended function but interface at degradation was noted. A short segment elevation 78'. of the moisture barrier was removed in an area with significant corrosion of the 3/4" knuckle plate above the moisture barrier, where the corrosion was suspected to occur below the moisture barrier. The moisture barrier was removed to a depth of approximately 1 ". Some corrosion of the 3/4" knuckle plate was noted below the surface of the moisture barrier at the floor level but the corrosion of the 3/4" knuckle plate did not extend below the portion of the moisture barrier that was removed.The 3/4" knuckle plate met the IWE acceptance criteria.4. Degradation was found as a result of implementation of Enhancement 
#2 to the IWE program as described in the LRA, Appendix B.2.1.28 (ASME Section XI, Subsection IWE), as shown on page B-1 32. As a result, areas that were previously inaccessible for inspection have been made accessible, examinations have been performed, and evaluations have verified the adequacy of existing conditions as described in item #3 of this RAI. Some corrective actions have been completed and additional corrective actions have been specified, as described below.Unit I -corrective actions completed during the refueling outage in the spring of 2010:* Examination of 100% of the accessible 1/2" containment liner plate and moisture barrier.* UT measurements of the 3/4" containment liner (knuckle plate) around the perimeter of the Containment.
Page 6 of 38 LR-N10-0165 Enclosure A* UT measurements of the 1/2" containment liner plate where insulation panels were removed and loss of material was observed.* Coating repairs of the 3/4" containment liner (knuckle plate)." The one vertical leak chase channel with a hole was capped.* Coating repairs at areas where containment liner insulation panels were removed to allow for containment liner plate inspection and corrosion was observed.* The moisture barrier was repaired or replaced.* Evaluation to confirm the identified loss of material is acceptable.
Unit 1 -additional corrective actions to be completed prior to the period of extended operation:
* Perform augmented examinations of the 3/4" containment liner (knuckle plate) at 78' elevation in accordance with IWE-2420.* Perform augmented examinations of the 1/2" containment liner plate behind insulation panels, where loss of material was previously identified, in accordance with IWE-2420.* Remove 1/2" containment liner insulation panels, adjacent to accessible areas where there are indications of corrosion, to determine the extent of condition of the existing corroded areas of the containment liner plate.Unit 2 -corrective actions completed during the refueling outage in the fall of 2009:* Examination of 100% of the accessible 1/2" containment liner plate and moisture barrier.* UT measurements of the 3/4" containment liner (knuckle plate) around the perimeter of the Containment.
* UT measurements of the 1/2" containment liner plate where insulation panels were removed and loss of material was observed.* The six vertical leak chase channels with a hole were capped.* Evaluation to confirm the identified loss of material is acceptable.
Unit 2 -additional corrective actions to be completed prior to the period of extended operation:
* Examine the accessible 3/4" containment liner (knuckle plate). If corrosion is observed to extend below the surface of the moisture barrier, excavate the moisture barrier to sound metal below the floor level and perform examinations as required by IWE.* Perform remote visual inspections, of the six capped vertical leak chase channels, below the containment floor to determine extent of condition.
* Remove the concrete floor and expose the 1/4" containment liner plate (floor)for a minimum of two of the vertical leak chase channels with holes. Perform examinations of exposed 1/4" containment liner plate (floor) as required by IWE. Additional excavations will be performed, if necessary, depending upon conditions found at the first two channels.Page 7 of 38 LR-N10-0165 Enclosure A* Remove 1/2" containment liner insulation panels, adjacent to accessible areas where there are indications of corrosion, to determine the extent of condition of the existing corroded areas of the containment liner plate.* Perform augmented examinations of the 1/2" containment liner plate behind insulation panels, where loss of material was previously identified, in accordance with IWE-2420.* Examine 100% of the moisture barrier in accordance with IWE-231 0 and replace or repair the moisture barrier to meet the acceptance standard in IWE-351 0.Examinations and inspections will be performed in accordance with IWE-2000 and the acceptance standards will be in accordance with IWE-3500.Updates to formalize these Salem commitments, to complete the future corrective actions described above, are made to LRA Appendix A, Section A.5, the License Renewal Commitment List, under line number 28, commitment 
#2, which can be found in Enclosure C.Page 8 of 38 LR-N10-0165 Enclosure A RAI B.2.1.28-2

==Background:==

GALL Report (NUREG-1801), AMP Xl.S1, "ASME Section XI, Subsection IWE," Program Element 1, requires inspection of steel containment components including liners, liner anchors, and integral attachments for loss of material due to general, pitting, and crevice corrosion.
Inservice inspection (ISI) requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code, Section Xl, Subsection IWE for steel containments (Class MC) and steel liners for concrete containments (Class CC) are imposed by 10 CFR 50.55a.Issue: Program Element 10 for the Salem ASME Section XI, Subsection IWE aging management program discusses sampling inspections of normally inaccessible areas of steel liner plate located behind the insulation panels around the lower 30 feet of the Unit 2 containment completed in 2009. Similar inspections are scheduled for Unit 1.However, details of the sampling methodology used for the inspection is not described in the LRA and program basis document.Request: 1. Describe the sampling methodology used in 2009 inspection to select the locations for inspecting containment liner plate and moisture barrier behind the insulating panels.2. The sampling methodology planned for future inspections.
Would the sampling methodology provide a statistical confidence level of at least 95% that the results of inspections will meet the acceptance criteria of IWE 3500.The staff needs the above information to confirm that the effects of aging of the containment pressure boundary metal will be adequately managed so that it's intended function will be maintained consistent with the current licensing basis for the period of extended operation as required by 10 CFR 54.21(a)(3).
Page 9 of 38 LR-N10-0165 Enclosure A PSEG Response: 1. Random sampling was not used in 2009 to select the locations for inspecting the containment liner plate and the moisture barrier behind the containment liner insulation lagging. During the Salem Unit 2 outage in the fall of 2009, the bottom edge of the containment liner insulation lagging was trimmed to allow for visual inspection of 100 % of the liner at the juncture of the containment concrete floor and the moisture barrier; except for areas where access is restricted by permanent plant structures, equipment, or components.
Salem recognized it is prudent to make this area accessible for visual examination prior to and during the period of extended operation to resolve concerns involving corrosion in this area.In addition to inspecting 100% of accessible containment liner plate and 100% of the moisture barrier, one containment liner insulation panel was removed in each quadrant, for a total of four containment liner insulation panels, to evaluate the acceptability of inaccessible areas covered by containment liner insulation.
Selection of the containment liner insulation panels was based on the presence of and the extent of corrosion in accessible areas below the containment liner insulation panels, staining on the containment liner insulation panels, or staining at the floor below containment liner insulation panels. Visual inspection of exposed, inaccessible containment liner plate, during the 2009 Salem Unit 2 examinations, indicated only minor surface corrosion.
Ultrasonic testing (UT) thickness measurements conducted at the insulation panel with the most loss of material showed that the measured containment liner plate thickness was greater than the nominal thickness of 0.5 inches.2. In LRA Appendix A, Section A.2.1.28 (ASME Section Xl, Subsection IWE), Enhancement 
#1, Salem committed to enhance the ASME Section Xl, Subsection IWE, aging management program to require inspections of a sample of the inaccessible containment liner covered by containment liner insulation and lagging prior to the period of extended operation and every 10 years thereafter.
The commitment is clarified as outlined below.Prior to the period of extended operation (PEO)* A sampling plan will be developed based upon guidance in EPRI TR-107514,"Age Related Degradation Inspection Method and Demonstration:
in Behalf of Calvert Cliffs Nuclear Power Plant License Renewal Application".
* The population size of containment liner insulation panels in each unit is approximately 264 panels. A sample size of 57 will meet the statistical requirements of a 95% confidence level that 95% of the containment liner plate behind the containment liner insulation meets the acceptance criteria of IWE-3500.* The samples will be randomly selected.* The examination will be performed by either removing the containment liner insulation panels and performing a visual inspection, or by using a pulsed eddy current (PEC) remote inspection, with the containment liner insulation left in place, to detect evidence of loss of material.
If evidence of loss of material is detected using PEC, the containment liner insulation panel will be subsequently removed to allow for visual and UT examinations.
Page 10 of 38 LR-N10-0165 Enclosure A* If acceptance criteria defined in IWE-3500 is not satisfied, the sampling plan will be modified as recommended in EPRI TR-107514.
During the period of extended operation During the PEO, a reduced sample size will be randomly selected and examined each Containment Inservice Inspection Period contingent upon satisfactory results of the sample examined prior to the PEO.* One containment liner insulation panel will be selected, at random, for removal from each quadrant, during each of the three Periods in an Inspection Interval.Therefore, a total of 12 containment liner insulation panels will be selected, in each unit, during each ten year Inspection Interval, to allow for examination of the containment liner behind the containment liner insulation.
* The randomly selected containment liner insulation panels in each quadrant will not include containment liner insulation panels previously selected.The sampling plan during the PEO is considered to provide reasonable assurance that the containment liner plate, in the inaccessible areas behind the Containment liner insulation, will meet the acceptance criteria in IWE-3500 and the containment liner plate will perform its intended function during the PEO. This is based upon the evidence to date from inspections performed behind the containment liner plate, where all examinations met the acceptance criteria, and the enhanced inspections prior to the PEO described above.Updates to LRA Appendix A and Appendix B as a result of the clarification of the enhancement can be found in Enclosure B. Updates to LRA Appendix A, Section A.5, the License Renewal Commitment List, under line number 28, commitment 
#1, can be found in Enclosure C.Page 11 of 38 LR-N10-0165 Enclosure A RAI B.2.1.29-1

==Background:==

GALL Report, Section XI.S2, Element 6 states that ASME Section Xi, Subsection IWL, Article IWL-3000 provides acceptance criteria for concrete containments.
The GALL Report further states that quantitative acceptance criteria based on the "Evaluation Criteria" provided in Chapter 5 of ACI 349.3R may also be used to augment the qualitative assessment of the responsible engineer.
Salem Generating Station Units 1 and 2, document SA-PBD- AMP-XI.S2, Rev. 2, Section 3.6 also states that quantitative acceptance criteria, developed based on Chapter 5 of ACI 349.3R, are included in the program implementing documents to augment the qualitative assessment by the responsible engineer.Issue: A review of the Salem Units 1 and 2 records indicate that IWL inspections performed in 2005, 2007, and 2009 indicate that Section 5.4 of S-C-CAN-SEE-1353, Rev. 0,"Acceptance Criteria for Containment Concrete Defects", has been used by the applicant for inspection of Salem Units 1 and 2 containment concrete surface examinations.
According to this document, the acceptance criteria for concrete surfaces is significantly different and less stringent from the acceptance criteria specified in Section 5.1 of ACI 349.3R.In addition, Notification 000020234570 describes the actual condition of the concrete on the north side of the Unit 2 containment involving surface spalling ranging up to 6 ft long and 16 inch wide, and spalling at joints that is up to 3 ft long and 4 in. wide. Notification 000020234570 also describes a condition on the north side of the containment between the equipment hatch and the fuel handling penetration area involving the protrusion of a pipe from the penetration wall. The applicant did not describe the purpose for the pipe, but the applicant reported that the pipe is broken at the flange. The notification also describes a piece of wood (1 in. by 8 in. by 4 in.) protruding from the penetration wall in the main steam area.Request: The applicant is requested to provide the following information:
: 1. The basis for the acceptance criteria in Section 5.4 of S-C-CAN-SEE-1 353, Rev. 0, including the reasons for it being significantly less stringent than the ACI 349.3R requirements.
: 2. Provide information about broken pipe and flange protruding from the containment surface, and its impact on the containment leak tightness.
: 3. Confirm that the piece of wood (1 in. by 8 in. by 4 in.) is not embedded in the concrete containment wall.Page 12 of 38 LR-N10-0165 Enclosure A 4. Details of corrective actions that the applicant plans to implement for using the acceptance criteria described in Section 5.4 of S-C-CAN-SEE-1 353, Rev. 0 which does not conform with the current industry practice and ACI 349.3R.The staff needs the above information to confirm that the effects of aging of the concrete containment will be adequately managed so that it's intended function will be maintained consistent with the current licensing basis for the period of extended operation, as required by 10 CFR 54.21(a)(3).
PSEG Response: 1. S-C-CAN-SEE-1 353 is no longer an active document in the Salem document control system. The acceptance criteria in Section 5.4 of S-C-CAN-SEE-1353 were never used since the containment concrete conditions never warranted using the acceptance criteria in Section 5.4 of S-C-CAN-SEE-1353.
The Salem ASME Section XI, Subsection IWL program, examination procedures now use guidance provided in ACI 349.3R.2. Notification 000020234570 describes a condition on the north side of the Salem Unit 2 Containment between the equipment hatch and the fuel handling penetration area involving the protrusion of a pipe from a wall. This pipe actually protrudes from a wall extending outward from the Fuel Handling Building.
The wall extending outward from the Fuel Handling Building encloses the space between the Fuel Handling Building and the Containment.
The pipe does not protrude from the Containment wall. Therefore, there is no impact on Containment leak tightness.
: 3. Notification 000020234570 describes a piece of wood (1 in. by 8 in. by 4 in.) that is not embedded in any concrete and is not touching the Containment.
The piece of wood is wedged between miscellaneous steel and the mechanical penetration area wall of the Auxiliary Building, near the Containment wall. This piece of wood has no impact on containment integrity.
: 4. Corrective actions were initiated as a result of differences between the acceptance criteria provided in Section 5.4 of S-C-CAN-SEE-1 353, Rev. 0, which does not conform with the current industry practice described in ACI 349.3R.S-C-CAN-SEE-1 353 is no longer an active document in the Salem document control system and the acceptance criteria described in Section 5.4 of S-C-CAN-SEE-1353 is no longer being used. The ASME Section Xl, Subsection IWL, aging management program has been enhanced to reflect the guidance provided in ACI 349.3R, which is now being used for the examination of the Salem Unit 1 and Unit 2 containments.
This is a new enhancement to the ASME Section Xl, Subsection IWL, aging management program which improves characterization of Containment degradation and revises the Containment inspection acceptance criteria.
This will allow for more effective trending of degradation for future inspections.
Site procedures have been revised to remove references to S-C-CAN-SEE-1353 and revise the acceptance criteria to reflect the ACI 349.3R acceptance criteria, as well as to add references to ACI 349.3R acceptance criteria, and consideration of long term aging management.
Page 13 of 38 LR-N10-0165 Enclosure A Concrete inspections for both Salem Unit 1 and Unit 2 Containments have recently been completed in April of 2010 using the ACI 349.3R tiered acceptance criteria.Examination results have been reviewed by the site Responsible Professional Engineer and satisfactorily met all ACI 349.3R acceptance criteria.Updates to LRA Appendix A and Appendix B as a result of the enhancement can be found in Enclosure B. Updates to LRA Appendix A, Section A.5, the License Renewal Commitment List, under line number 29, can be found in Enclosure C.Page 14 of 38 LR-N10-0165 Enclosure A RAI B.2.1.29-2

==Background:==

GALL Report (NUREG-1801), AMP XI.S2, "ASME Section XI, Subsection IWL," Program Element 10 states that implementation of ASME Section XI, Subsection IWL, in accordance with 10 CFR 50.55a, is a necessary element of aging management for concrete containments through the period of extended operation.
Issue: Program Element 10 for the Salem ASME Section XI, Subsection IWL aging management program describes results of concrete inspections conducted in April 2001 and October 2005 for Unit 1, and November 2000, May 2005, and August 2009 for Unit 2. In addition to isolated areas of physical damage to concrete surfaces on both units, normal shrinkage cracking was also observed.
Salem Units 1 and 2 containments are constructed from reinforced (non-prestressed) concrete; therefore, cracking of the concrete in some areas is likely and is expected over the 60-year operating life. In Notification 000020234570, the applicant reported cracks in the concrete coating over the entire outside of the Unit 2 containment.
Long-term exposure of concrete cracks to salt spray originating from the Delaware Bay could result in corrosion of the embedded steel reinforcing bars located nearest to the outer surface of the containment concrete during the extended period of operation.
Request: The applicant is requested to provide the following:
: 1. The extent and maximum width of the cracks observed in Salem Unit 1 and 2 containments.
: 2. Actions that are planned to mitigate the consequences of chloride ion penetration to the level of the embedded steel reinforcing bars over the period of extended operation.
This may be necessary since the Salem Units 1 and 2 concrete containment surface inspection reports documented scaling and spalling of up to 3 inches.3. If no actions are anticipated to mitigate the consequences of chloride ion penetration to the level of the embedded steel reinforcing bars, the applicant is requested to provide an assessment of this time-dependent phenomenon and the basis for this decision.The staff needs the above information to confirm that the effects of aging of the concrete containment will be adequately managed so that it's intended function will be maintained consistent with the current licensing basis for the period of extended operation, as required by 10 CFR 54.21 (a)(3).Page 15 of 38 LR-N10-0165 Enclosure A PSE&G Response: 1. Concrete inspections for both Salem Unit 1 and Unit 2 Containment structures were completed in April 2010 using the ACI 349.3R tiered acceptance criteria.Examination results have been reviewed by the site Responsible Professional Engineer, and found acceptable, meeting ACI 349.3R acceptance criteria.
These results, including extent and maximum width of cracks, are summarized below: General: The overall crack patterns are very similar for both Unit 1 and Unit 2 containments.
The concrete surfaces exhibit general pattern cracking over the entire surface as well as cracking at random areas of mortar patches. The mortar patches were originally applied at the construction joints. Minor degradation of the mortar patches was noted.Cylindrical walls: Pattern cracking on about a 15" by 15" grid is evident over most of the cylindrical walls, with crack widths of about 0.015".Dome: There is similar pattern cracking on the tops of the Unit 1 and Unit 2 domes. The crack widths across most of the domes are about 0.015" wide with some areas at the top having cracks up to 0.040" wide.Maximum width and extent of cracking: " At the Unit 2 Containment, El. 130' airlock, some of the cracks were 0.0625" wide at the surface." At the Unit 1 Containment, inside the Penetration area, above the floor at elevation 78', a circumferential crack 0.032" wide was noted.Comparison to original structural integrity tests: The above conditions were compared to those found during the original start-up structural integrity tests. The cracks are characterized as passive and inactive.
The extent of the cracking and maximum crack widths is expected and consistent with the crack patterns exhibited following the original start-up structural integrity tests.Widening of cracks at the surface was identified and evaluated as part of the original structural integrity tests and accepted as a shallow, surface condition that was acceptable.
: 2. Salem will continue to monitor the condition of Unit 1 and Unit 2 containment concrete surfaces for spalling, scaling, cracking, and rust stains which are indicative of reinforcing bars corrosion.
The monitoring activities will be in accordance with the applicable edition and addenda of ASME Section XI, Subsection IWL, as approved in 10 CFR 50.55a and recommended in the GALL Report, Rev. 1. Inspection and acceptance criteria will be in accordance with ACI 349.3R as described in LRA Section B.2.1.29.
If acceptance criteria specified in ACI 349.3R for spalling, scaling, and cracking cannot be met, corrective actions will be implemented as required by the corrective action process to address corrosion of reinforcing bars. These Page 16 of 38 LR-N10-0165 Enclosure A actions may include mitigative measures, such as repairs to scaled and spalled areas of concrete, and sealing of cracks to minimize penetration of chloride ions. If corrosion staining of reinforcing steel is observed on containment concrete surfaces, an engineering evaluation will be conducted to assess the condition of reinforcing bars and the impact of rebar corrosion on containment structural integrity.
As described in the response to item 1 above, the Unit 1 and Unit 2 concrete containment surfaces were not spalled up to 3 inches, but rather had minor scaling and spalling.
Therefore, there is currently no need for specific mitigative actions to prevent the potential of chloride ion penetration to the level of embedded reinforcing bars.3. Chloride ions are common in nature and small amounts can be unintentionally contained in the concrete mixture ingredients.
Potential external sources of chlorides include exposure to seawater or spray, deicing salts, or those from accelerating admixtures.
The penetration and diffusion of chloride ions in concrete and their impact on reinforced concrete has been a subject of tests and studies as documented in the American Concrete Institute (ACI) ACI 222R, "Protection of Metals in Concrete Against Corrosion", ACI 349.3R, "Evaluation of Existing Nuclear Safety-Related Concrete Structures", ACI 365.1 R, "Service Life Prediction---State-of-the-Art Report", NUREG/CR-6927, "Primer on Durability of Nuclear Power Plant Reinforced Concrete Structures 
-A Review of Pertinent Factors", and other literature.
NUREG/CR-6927 also includes the results of concrete condition surveys in ten nuclear power plants. The review of ACI 222R, ACI 349.3R, ACI 365.1R, and NUREG/CR-6927 indicates that chloride ion penetration and diffusion in concrete depends on concrete durability and serviceability which have been incorporated into codes such as ACI 318, "Building Code Requirements for Structural Concrete".
Durability of concrete has been included in ACI 318 through specification of maximum water-cement ratios, cement content, type of cement, entrained air, minimum cover over the reinforcing bars, and control of cracks. The Salem Containments are constructed of concrete that conforms to the applicable ACI 318 requirements.
The minimum concrete clear cover over the reinforcing bars shown on the design drawings is 33/8 inches nominal which is greater than the 2 inch cover required by ACI 318 for concrete exposed to weather. Recent examinations of Unit 1 and Unit 2 Containment concrete surfaces using procedures that are based on ACI 349.3R inspection and acceptance criteria identified only minor spalling and scaling, but none that reduce the concrete cover over the reinforcing bars below the 2 inches required by ACI 318. Cracking is minor as described in the response to the item 1 of this RAI. In addition, the Containment concrete is observed to be free of large penetrating cracks that could permit significant chloride ion penetration to reach the level of reinforcing bars.The primary concern associated with penetration and concentration of chloride ions over time is that it can lead to corrosion of the reinforcing bars. Reinforcing bars with adequate concrete cover should not be susceptible to corrosion because the highly alkaline conditions present within concrete cause a passive iron-oxide film to form on the surface of the reinforcing bars. Chloride ions, however, can destroy this passive film and initiate corrosion.
Corrosion of reinforcing bars (i.e., the transformation of metallic iron to ferric oxide, or rust) is accompanied by an increase in volume of the metallic iron. The volume increase can cause cracking, spalling, and delamination of the concrete that can be visible in the form of loss of concrete material, rust spots and stains, and cracks in the concrete cover along the reinforcing bars. Visual Page 17 of 38 LR-N10-0165 Enclosure A inspection required to be conducted in accordance with ASME Section XI, Subsection IWL, would detect such conditions before the loss of containment intended function.
To date, corrosion of containment concrete reinforcing bars has not been identified as a concern by Salem or industry operating experience.
In summary, chloride ion penetration and diffusion in concrete has been a subject of extensive research and studies. The extent of the penetration and diffusion depend on concrete durability, permeability, and cracking, which have been considered in concrete design codes and standards.
Salem conforms to the applicable concrete codes and standards.
In the event this time dependent phenomenon penetrates to the level of reinforcing bars and initiates corrosion, the increase in volume of the steel due to the creation of rust will result in spalling, cracking, delamination of concrete, and staining of concrete surfaces.
Implementation of ASME Section Xl, Subsection IWL, aging management program described in LRA B.2.1.29 is considered to provide reasonable assurance that these aging effects will be detected and corrective actions will be taken prior to the loss of the Containment intended function.Page 18 of 38 LR-N10-0165 Enclosure A RAI B.2.1.33-1

==Background:==

NRC Information Notice 2004-05, "Spent Fuel Pool Leakage to Onsite Groundwater," notes that leakage of the spent fuel pools has occurred at Salem Unit 1.Issue: The licensee at Salem NGS in 2002 identified evidence of radioactive water leakage through the interior wall of the Unit 1 auxiliary building mechanical penetration room. In the years since initial startup, materials such as boric acid and minerals have accumulated in the leak collection and detection system that restricted normal drainage of fluid. Modification of the tell-tale drains that are used to detect, monitor, and quantify potential leakage from the spent fuel pool liner resulted in inadvertent further restriction of free drainage of leakage from the liner that resulted in accumulation of borated water between the liner and concrete and migration to other locations through penetrations, construction joints, and cracks. The seismic gap was confirmed to contain water with radionuclides characteristic of the Unit 1 spent fuel pool water and leakage into the seismic gap has continued.
Leakage into the tell-tale drains is occurring at a rate of about 100 gallons per day.Request: a. Provide historical data on the leakage occurrence and volume, and available information from chemical analysis performed on the leakage.b. Provide a summary of the root cause analysis that was used to identify the source of leakage through the liner that has resulted in accumulation of borated water between the liner and concrete, including information on the path of the leakage and structures that could potentially be affected by the presence of the borated water.c. Discuss plans for remedial actions or repairs to address leakage through the spent fuel pool liner. In the absence of a commitment to fix the leakage prior to the period of extended operation, explain how the structures monitoring program, or other plant-specific program, will address the leakage to ensure that aging effects, especially in inaccessible areas, will be effectively managed during the period of extended operation.
: d. Provide background information and data to demonstrate that the concrete and embedded steel reinforcement have not been degraded by exposure to the borated water and that the liner will not be impacted.
If experimental results will be used as part of the assessment, provide evidence that the test program is representative of the materials and conditions that exist in the region between the spent fuel pool liner and concrete.
This information should also include the MPR Associates report that documents the details of the tests performed and evaluation of SNP spent fuel pool concrete and rebar.e. If a concrete sampling program (e.g., obtaining concrete cores in region affected)cannot be implemented, please explain why this is not feasible.The staff needs the information to confirm that the potential effect of aging of the spent fuel pool reinforced concrete, liner, and steel reinforcement due to presence of borated Page 19 of 38 LR-N10-0165 Enclosure A The staff needs the information to confirm that the potential effect of aging of the spent fuel pool reinforced concrete, liner, and steel reinforcement due to presence of borated water will be adequately managed so that the intended function of impacted structural members will be maintained consistent with the current licensing basis for the period of extended operation as required by 10 CFR 54.21(a)3.
PSEG Response: a. Unit 1 History: The Unit 1 Spent Fuel Pool (SFP) leakage consists of leakage through the leak chase system drains (telltales) and leakage through concrete cracks and construction joints in the Fuel Handling Building (FHB) walls.In 1980 a small leak was discovered in the Spent Fuel Pool (SFP) telltale drains.Underwater inspections determined that the cause was due to leaking seam welds.The identified leaking seam welds were repaired.
After the repairs were implemented, the observed leakage was reduced to less than 0.2 gallons per day (gpd).In 2002, Salem identified evidence of active water leakage through an exterior wall of the Unit 1 Auxiliary Building (AB) mechanical penetration room. Chemical analysis of the water verified the water was consistent with borated SFP water. Further investigations revealed that the SFP telltales were blocked. The blockage resulted in water accumulation behind the SFP liner and ultimately to migration of leakage through construction joints and small cracks in the Fuel Handling Building walls. The escaped water accumulated in the seismic gap between the FHB and AB and reached groundwater within the plant controlled area. Corrective actions were initiated to remove blockage in the telltales and drainage system. Analysis of the removed material showed that the deposits were largely quartz (Si02) and calcite (CaCO3) with minor amounts of gismondine (CaA12Si2O8.4H20).
These materials are consistent with FHB concrete.
In addition the pH of water collected from the telltales after the removal of the blockage was 7.1 which is consistent with expected values for water in contact with concrete.
The expected pH of SFP water not in contact with concrete should be equivalent to SFP chemistry which is 4.6.In 2003, Salem began monitoring and trending leakage through the telltales.
The volume of leakage varies as indicated in Figure 1 below but on an average is approximately 100 gallons per day. This figure shows average daily leakage up to February, 2008. Based on sump pump run times, daily telltale leakage rates have remained approximately the same for 2008, 2009 and early 2010. The trends of leakage through the telltales are used as an indicator that corrective actions are required to maintain proper drainage through the leak chase system, thus, preventing buildup of borated water between concrete and the liner.Page 20 of 38 LR-N10-0165 Enclosure A Figure 1 Salem Unit 1 SFP Telltale Leakage -Daily Average 250.0 200.0 C. 150.0 100.0 50.0 0.0 Ve %( Ok % ve 0%Unit 2 History: During an inspection of the Salem Unit 2 SFP in April of 2010, evidence of active leakage was identified from the telltale lines. Based on the inspection results, it was determined that for Unit 2 while there is leakage, the actual volume is on the order of a gallon per day and as such the use of measuring the sump pump run time is too small to provide meaningful quantification.
In most instances the Unit 2 leakage actually evaporates before causing the sump pump to run.The evidence obtained from the leakage identified on Unit 2 is consistent with the historical information derived from Unit 1 with respect to chemistry, radionuclide content, and pH information.
These data suggest that the leakage from Unit 2 has been identified at an early stage of discovery and exhibits similar conditions that historically occurred at Salem Unit 1. Boroscopic inspection of the telltales in April 2010 showed that the telltales were open, however, based on the evidence that Unit 2 does have a relatively small but active leak, monitoring and trending will continue.In addition, telltales cleaning will be performed as necessary to ensure that the telltales remain open.Page 21 of 38 LR-N110-0165 Enclosure A b. The evaluation of the leakage source for Unit 1 is documented in a focused self-assessment performed by Salem in 2003. Salem has since monitored and trended the leakage. The following is a summary of the self-assessment and the leakage monitoring.
Leakage Source: Salem has concluded that leakage through the SFP liner is occurring in many small cracks in seam welds or plug welds located throughout the SFP liner. These cracks are too small to be readily identified and located. This leakage enters channels behind the SFP liner either directly from cracks in seam welds or indirectly by migrating over concrete from cracks in plug welds. The leakage drains to the channels then to 1 inch telltales and eventually to a sump.Below are key points that led to the above conclusion." Chemical and isotopic analyses of the telltale leakage indicate the presence of boron and gamma activity at concentrations consistent with the SFP water.* Isotopic analysis of the water discharged from the telltales after they were initially cleaned in 2003 was consistent with the SFP except it had aged several years suggesting that the leakage had been trapped between the SFP liner and concrete for some time.* The leakage rate from the telltales varies, but is on the order of 100 gallons per day. Salem Engineering estimates this leakage rate could be a result of a single or multiple cracks 0.001 inch wide and 6 inches long.* The leakage is likely from small cracks in seam welds (2,100 linear feet) of adjoining liners plates or the plug welds (1,400 total) that connect the liner plates to the steel embedded in the surrounding concrete.
This conclusion is supported by the observation of leakage in all of the telltale drains.* The conclusion that some of the leakage is through plug welds is based on chemical analysis of the deposits that were blocking the telltales.
Analyses showed that the blockage debris was formed by materials consistent with FHB concrete, which can only occur if plug welds are leaking.* The backing bar for the seam welds is tied to the embedded leakage channels and the plug welds are tied to the embedded steel. Therefore, the cause of the cracks is postulated to be due to differential thermal expansion between the liner and the concrete structure.
Page 22 of 38 LR-N10-0165 Enclosure A Leakage Path: The leak path in the Unit 1 SFP begins at liner seam welds or liner plug welds.Leakage through seam welds is immediately directed into leak chase system collection channels and does not contact concrete or any structural elements.Leakage through plug welds may accumulate between the concrete and the SFP liner and migrates to leakage collection channels.
Leakage proceeds down the leakage collection channel to a telltale and drains into the sump collection system.Some time between 1995 and 1998 the telltales became blocked and the leakage accumulated in the space between the SFP liner and the concrete wall. The hydrostatic head associated with this accumulation drove migration of the leakage through the SFP structure along cracks and construction joints. The leakage reached the seismic gap between the Fuel Handling Building and the Auxiliary Building and entered the Auxiliary Building through a wall sleeve. In 2003 corrective actions were initiated to remove blockage in the telltales and drainage system. In addition, Salem has installed a drain in the seismic gap which drains water to the Auxiliary Building, thus, minimizing leakage to the groundwater.
Potential Impact to Salem Structures:
The accumulation of leakage in the seismic gap can impact the Fuel Handling Building and the Auxiliary Building.
A description of the impact on the FHB is provided in the response to item "d" below. An assessment of the impact of borated water accumulation in the seismic gap on the Auxiliary Building was performed in 2009. The assessment used the concrete degradation curve developed from testing in support of the FHB evaluation with corrections to account for the difference in boron concentration and lower temperatures.
The assessment concluded that potential degradation of the Auxiliary Building from exposure to the borated water in the seismic gap is minimal.c. There are currently no plans to perform repairs to the SFP liner seam welds or plug welds. Repairs were considered previously but determined to be impractical because the pool is nearing capacity of stored spent fuel making access challenging.
Also locating the cracks using the existing underwater NDE technology has not been successfully demonstrated.
In 1995 Salem inspected 95% of all seam welds and found no through wall cracks. This suggests the leakage is occurring below the sensitivity of the test.Salem continues to participate in and monitor industry activities to develop crack detection methodologies and repair methods. This includes EPRI, the Material Aging Institute, and Electricit6 de France (EdF) activities.
In addition, Salem will continue to evaluate repair methodologies as they become available for potential implementation.
Salem has conducted extensive laboratory testing to characterize the extent of concrete and rebar degradation as a result of exposure to borated water through the period of extended operation (PEO). The results of the tests were used to evaluate the impact of the leakage on the FHB structural integrity.
The evaluation concluded Page 23 of 38 LR-N10-0165 Enclosure A the impact on structural integrity is not significant and that the FHB will continue to perform its intended function through the PEO.As indicated in the Salem License Renewal Commitment List in LRA Appendix A, Section A.5, page A-68, line item number 33, Enhancement 5 a, b, and c, the long-term aging management strategy for addressing borated water leakage from the SFP includes the following actions related to the spent fuel pool liner: 5. Require the following actions related to the spent fuel pool liner: a. Perform periodic structural examination of the Fuel Handling Building per ACI 349.3R to ensure structural condition is in agreement with the analysis.b. Monitor telltale leakage and inspect the leak chase system to ensure no blockage.c. Test water drained from the seismic gap for boron concentration.
Note -see below response to part B.2.1.33-le for updates to Enhancement 5.d. Salem has completed an assessment of the long-term structural adequacy of the Unit 1 FHB reinforced concrete structure under potential prolonged exposure of the concrete and reinforcing steel to borated water which has leaked from the pool. The results of the assessment are documented in MPR Report 2613, "Salem Generating Station Fuel Handling Building-Evaluation of Degraded Condition," Revision 3. This report is provided in enclosure "D". Below are summaries of the assessment elements and conclusions.
FHB Concrete Degradation Testing: The most severe degradation is most likely occurring behind the SFP liner at the bottom and sides of the pool. It is neither practical nor safe to obtain samples of the degraded concrete.
Instead Salem has performed concrete testing to quantify the potential degradation of the FHB structure.
Beginning in 2003 a series of borated water degradation tests were conducted on core samples taken from the Salem Auxiliary Building and new concrete specimens that were made to the original Salem specifications using aggregate from the original sources. Concrete test specimens were submerged in a borated water bath for varying periods up to 39 months. The borated water bath testing parameters (borated water concentration and temperature) were maintained to be representative of the SFP water. The tests were conducted in a manner which was conservative with respect to the actual conditions expected in the FHB. Test specimens were removed from the bath at periodic intervals and degradation of the exposed concrete surface was measured.
Results show that degradation is diffusion-controlled, which is consistent with published studies of weak acid attack of concrete.
A curve fit of degradation versus time was generated using the measured degradation from test specimens.
The following figure 2 shows the curve fit.Page 24 of 38 LR-N10-0165 Enclosure A S0 I Q0 Q)0.1.0.0.0.0.0.1.5 31 __13 _ 75 38 19l f)0 10 20 30 40 50 60 70 80 90 Time (year)Two Sigma Figure 2 -Extrapolation of Degradation through End of Plant Life Test Result

== Conclusions:==
 " The Borated water attacks the calcium hydroxide component of the cement paste causing loss of bonding of the coarse and fine aggregate." The predicted depth of penetration of concrete after 70 years exposure to spent fuel pool water at 1 00 0 F is 1.30 inches. This includes a two-sigma statistical uncertainty of the test data and an adjustment for temperature.
Since the concrete clear cover for the walls and slab is greater than this projected degradation depth, the borated water will not reach rebar during the 70 year period.* The wicking effect at the rebar/concrete interface was observed to be minor.That is, the degradation rate of the concrete at the rebar/concrete interface is similar to the general rate of attack of concrete without rebar. Therefore, degradation of rebar at the construction joints or cracks will not spread rapidly along the bond with the rebar. Rebar functionality is not compromised from a general loss of bond with the concrete.Assessment Corroboration from the Connecticut Yankee SFP Cores: In the fall of 2007 Salem learned that EPRI had possession of several samples taken from the Connecticut Yankee (CY) SFP during decommissioning.
Salem and EPRI collaborated on evaluation of these concrete cores. The objective of the evaluation was to use actual CY plant observations to corroborate the assessment of the Salem FHB.Page 25 of 38 LR-N10-0165 Enclosure A Overall, the CY cores corroborate the MPR results of testing and projections for Salem. The concrete used at CY is similar to that used at Salem in that the aggregates (coarse and fine) were inert with respect to borated water. The degraded concrete in the CY cores varied from less than 0.05 inch to 0.91 inch which demonstrates that the degradation depth can vary. Over the 37 year life of the CY Spent Fuel Pool, the concrete degradation reached a maximum depth of 0.91 inch.These results are consistent with the above degradation curve developed from the Salem concrete testing.Although the surface closest to the SFP liner of the CY cores was degraded from borated water attack, the embedded reinforcing steel exhibited no corrosion or loss of bond with the cement. The presence of secondary deposits, including secondary deposits in cracks found in the CY cores, provide evidence that water migration occurred.
These results confirm that cracks in concrete and other concrete defects do not promote reinforcing steel degradation.
Published Studies: Published studies demonstrate that attack of Portland cement concrete by weak acids, such as borated water, usually results in low depths of penetration, and are diffusion-controlled.
Curve fits of the test data show that the depth of degradation versus time follows a Fick's Law of Diffusion formulation where depth increases with the square root of time.A published study has also investigated the effects of reinforcing steel corrosion due to borated water entering reinforced concrete through cracks. The tests showed that corrosion increases as crack width increases and pH decreases.
In particular, the tests showed negligible reinforcing steel attack even when specimens were subjected to the most corrosive test environment (pH of 5.2) with the largest crack width (0.4 mm) for a period of two years. Corrosion was limited to surface scarring in the area of the crack.Existing FHB Structure Design Margin: MPR performed an assessment of the FHB given the leakage paths and borated water attack described above. Below is a summary of the assessment results.The concrete and embedded steel degradation rates developed from the core testing were used to calculate the reduction in available design margin. The analysis assumed that leakage continued through the end of plant life, which was defined as 70 years.The slab underneath the SFP has been exposed to borated water leakage since early in plant life. The areas in the vicinity of leaking welds have likely experienced the most degradation, since some leakage at plug welds must migrate along the slab surface to the leakage collection channels, the projected depth of concrete degradation in these areas of the slab is 1.30 inches, assuming exposure to borated water for 70 years. This has no impact on the structural capacity of the slab since the reinforcing steel is protected by a 6 inch thick leveling layer of concrete which is not credited in the structural evaluations and an additional 2.46 inches of concrete clear cover.Page 26 of 38 LR-N10-0165 Enclosure A The walls surrounding the SFP were exposed to borated water during the time period when the telltales were blocked and leakage accumulated in the gap between the SFP liner and the walls. Exposure of the walls to borated water started between 1995 and 1998 when the leakage collection channels and telltales became blocked and extended to early 2003 when drain flow was re-established.
General degradation of the walls from exposure to borated water over this 8 year period had a minimal impact on the capacity of the FHB structure.
Specifically, the depth of degradation over this period is calculated as 0.44 inches. This depth is less than the concrete clear cover of 4.4 inches on wall locations where there are no leakage collection channels, and 2.9 inches where leakage collection channels exist in the walls. Therefore, the reinforcing steel was protected from exposure to borated water.A 0.44 inch reduction in wall thickness reduces the structural margin at the most limiting wall by 0.4% (compared to the code allowable).
Degradation of the concrete walls in the immediate vicinity of a leaking plug weld will continue even though the stored inventory has been drained. Such degradation will be highly localized and has no impact on the overall capacity of the concrete wall. Localized reinforcing steel degradation from borated water migration through cracks and construction joints had a negligible impact on the FHB walls during this period.The assessment also concluded that reinforcing steel located in construction joints and cracks which are exposed to borated water will not be significantly degraded through the PEO. Based on the corrosion rate of carbon steel in de-aerated boric acid, this reinforcing steel may experience a radial reduction of 0.011 inches in 70 years which is insignificant.
Therefore, the structural assessment concluded that there are no structural challenges to the design basis of the FHB through the end of plant life including the period of extended operation.
Independent ACI Structural Assessment:
Salem had an experienced site concrete structural engineer perform an independent structural assessment of the FHB in 2006. The assessment included review of building drawings, a visual inspection of the accessible portions of the FHB exterior walls, and a visual inspection in the Sump Room. The checklist in ACI 201.1 R-92,"Guide for Conducting a Visual Inspection of Concrete in Service" was used to guide the inspections.
Observations were compared to limits in ACI 349.3R, "Evaluation of Existing Nuclear Safety-Related Concrete Structures".
Key conclusions from the independent assessment are excerpted below.* "Overall the concrete appears to be in good structural condition."* The "appearance of leaching or chemical attack and corrosion staining of undefined source on concrete surfaces do not indicate significant structural deterioration at this time."* There were "no indications of concrete surface expansion due to reinforcing steel corrosion." Page 27 of 38 LR-N10-0165 Enclosure A Assessment of SFP Liner: Salem has also assessed the impact of potential degradation of the slab on the integrity of the SFP liner. The primary concern is that local degradation of the slab can create "voids" underneath the SFP liner. If the void corresponds to the location of a rack foot, the foot may no longer be supported on a firm surface. A scoping assessment included in MPR-2613 demonstrates that the liner is sufficiently ductile to accommodate the load from the fuel racks even if the foot of the rack is positioned over an area of local concrete degradation.
: e. Salem has conducted extensive testing to understand and quantify concrete and reinforcing rebar degradations.
The results of these tests were used to evaluate the FHB structural integrity through the PEO. The evaluation concluded the FHB will perform its intended function through the PEO. Thus obtaining concrete cores in the region most affected by borated water is not required.
However, a shallow concrete core will be taken to assess potential degradation of the FHB from borated water.The shallow core sample will be taken in the SFP wall where previous inspections have shown evidence of borated water migration through the concrete.
This action is being added as item d under Enhancement 5 of the Structures Monitoring Program.In preparing this response it was noted that Structures Monitoring Program Enhancement 5.c was not clear with respect to sampling the water taken from the seismic gap for ph, chlorides, and sulfates; although it is generally described in enhancement 
: 11. Therefore enhancement 5.c is revised to clarify this enhancement.
These updates to Enhancement 5 are shown below, and are captured in updates to the Appendix A and Appendix B Structures Monitoring Program description (See Enclosure B) and the License Renewal Commitment List (See Enclosure C). The revisions to item 5.c and the new action, item 5.d, are shown in bolded italic text: 5. Require the following actions related to the spent fuel pool liner: a. Perform periodic structural examination of the Fuel Handling Building per ACI 349.3R to ensure structural condition is in agreement with the analysis.b. Monitor telltale leakage and inspect the leak chase system to ensure no blockage.c. Test water drained from the seismic gap for boron, chloride, and sulfate concentrations; and pH.d. Perform a shallow core sample in the SFP wall where previous inspections have shown ingress of borated water through the concrete.
The core sample will be examined for degradation from borated water.Page 28 of 38 LR-N10-0165 Enclosure A RAI B.2.1.33-2

==Background:==

The LRA states that leakage of borated water has occurred in SNGS Units 1 and 2 reactor cavities during refueling outages, but the leaks have been contained within the Containment Building.Issue: In April 2006 visual structural examinations of the accessible portions of the containment reinforced concrete structures for SNGS Units 1 and 2 indicated that the concrete was apparently in good structural condition.
It is unclear to the staff that leakage of the borated water has not resulted in degradation of either the concrete or embedded steel reinforcement that is inaccessible for inspection.
Request: a. Provide historical data on the leakage occurrence and volume, and available information from chemical analysis performed on the leakage.b. Provide the root cause analysis that was used to identify the source of leakage, including information on the path of the leakage and structures that could potentially be affected by the presence of the borated water.c. Discuss plans for remedial actions or repairs to address leakage. In the absence of a commitment to fix the leakage prior to the period of extended operation, explain how the structures monitoring program, or other plant-specific program, will address the leakage to ensure that aging effects, especially in inaccessible areas, will be effectively managed during the period of extended operation.
: d. Provide background information and data to demonstrate that concrete and embedded steel reinforcement potentially exposed to the borated water have not been degraded.
If experimental results will be used as part of the assessment, provide evidence that the test program is representative of the materials and conditions that exist.The staff needs the information to confirm that the potential effect of aging of the reinforced concrete due to presence of borated water will be adequately managed so that the intended function of impacted structural members will be maintained consistent with the current licensing basis for the period of extended operation as required by 10 CFR 54.21(a)3.
Page 29 of 38 LR-N 10-0165 Enclosure A PSEG Response: a. A summary of available Unit 1 and Unit 2 reactor cavity and fuel transfer canal leakage history is provided below: Unit 1: During the Salem Unit 1 2005 refueling outage white deposits were observed in several locations of the Unit 1 containment.
Chemical analyses indicated that the deposits originated from the reactor cavity or fuel transfer canal borated water leakage. In addition, the deposits contained constituents consistent with concrete.Follow-up walkdowns found evidence of a recent leak (water) in the "N16" Decay Tunnel. Analysis of the water indicated it was consistent with the reactor cavity water. The six fuel transfer canal telltale drain lines were found blocked. After cleaning, four of the telltales showed evidence of leakage ranging from approximately 1 to 60 drops per minute. The leakage rates steadily decreased to zero during the refueling outage.* A review of the corrective action database indicates that no active leakage associated with the reactor cavity and fuel transfer canal liner was documented during the Unit 1 2007 nor during the 2008 refueling outages. However, during the 2008 refueling outage chemical analysis was performed on deposits collected from the "N16" tunnel. Analysis showed that the residue originated from either the reactor cavity or the fuel transfer canal." During the 2010 Unit 1 refueling outage no active leaks were observed.Unit 2:* In November 2003 liquid was observed running down the containment liner plate and lagging under the fuel transfer canal inside containment, and pooling on the concrete floor. No evidence of corrosion was observed at the containment liner to floor joint." In April 2005 during the Salem Unit 2 refueling outage, water was observed dripping down the wall on the containment liner. Also, white deposits were observed in several locations on the Letdown Heat Exchanger Room walls under 2 of the 6 telltales, which were dripping.
Analysis showed the deposits and water were from the reactor cavity. The two dripping telltales were probed and no blockage was found.* During the 2006 Unit 2 refueling outage a small amount of leakage was observed coming from a telltale in the Letdown Heat Exchanger Room. During the 2008 Unit 2 refueling outage no active leakage was documented.
However, during the 2009 Unit 2 refueling outage a 60 dpm active leak was found from the telltale located above the door to the Letdown Heat Exchanger Room. A sample of the water was analyzed.Analysis of the water indicated it was consistent with reactor cavity water.* Evidence of boric acid deposits on the Unit 2 containment liner under the fuel transfer canal have been observed during multiple outages since November 2000.b. Chemical analysis of collected leakage during multiple refueling outages at both Units 1 and 2 shows that the source of the leakage is from the reactor cavity or the Page 30 of 38 LR-N10-0165 Enclosure A fuel transfer canal. Assessments for Unit 1 and 2 have concluded that the potential cause of leakage is very small cracks in the reactor cavity or fuel transfer canal liner.The majority of the leakage enters the leak collection chases and drains through the telltales.
Some of the telltales in the Letdown Heat Exchanger Room associated with the fuel transfer canal liner have been observed with active leaks during refueling outages. A second leakage path occurs in the vicinity of where the fuel transfer canal exits containment.
The leakage path is postulated to be through the reactor cavity and fuel transfer canal liner, then through concrete construction joints and cracks, and then down the sides of the containment liner behind the lagging inside containment.
The leakage only occurs when the reactor cavity and fuel transfer canal are flooded up for refueling.
Active leaks have been observed sporadically only during refueling outages and measured leakage rates are less than 100 drops per minute.This leakage has the potential to impact the reactor cavity and fuel transfer canal reinforced concrete structure.
In addition, the leakage has the potential to impact the containment liner. The impact of leakage on the containment liner is documented in Salem's response to RAI B.2.1.28-1.
: c. Salem has concluded that based on the short duration of the refueling activities and the very long exposures needed to degrade reinforced concrete, that remedial actions are not needed. Salem will continue Structural Monitoring (B.2.1.33) and ASME Section Xl, Section IWE Program (B.2.1.28)
Inservice Inspections to ensure the continued integrity of the in-scope structures.
Salem will enhance the Structure Monitoring Program to perform periodic inspection of telltales associated with the reactor cavity and fuel transfer canal liner to ensure the telltales are free of significant blockage and to periodically monitor for leakage when the cavity is flooded. Keeping the telltales free of blockage will ensure that water that has entered the inaccessible areas between the liner and concrete will only contact the concrete for short durations.
Salem will also inspect concrete surfaces for degradation where leakage has been observed, in accordance with this Program. In addition, the Structural Monitoring Program will be enhanced to require pH testing of leakage from the telltales.
LRA Appendix A, Section A.2.1.33, and Appendix B, Section B.2.1.33, the Structures Monitoring Program descriptions, are being revised to reflect these enhancements.
See Enclosure B to see these updates.Also, refer to Enclosure C of this letter to see the corresponding change to the License Renewal Commitment List, LRA Appendix A, Section A.5.Page 31 of 38 LR-N10-0165 Enclosure A d. Salem has performed an assessment of the long-term structural adequacy of the Salem Unit 1 Fuel Handling Building (FHB) reinforced concrete structure under potential prolonged exposure of the concrete and reinforcing steel to borated water.The results of the assessment are documented in MPR Report 2613, "Salem Generating Station Fuel Handling Building-Evaluation of Degraded Condition," Revision 3, which is attached in enclosure "D".Conclusions of the assessment are summarized below." The predicted depth of concrete degradation after 70 years of continuous exposure to borated water is 1.3 inches.* The degradation rate of the concrete at the rebar and concrete interface is similar to the general rate of attack of concrete without rebar. Therefore, degradation of rebar at the construction joints or cracks will not spread rapidly along the bond with the rebar. Rebar functionality is not compromised from a general loss of bond with the concrete.Salem has concluded that the findings for the FHB are directly applicable to the Unit 1 and 2 the reactor cavity and fuel transfer canal reinforced concrete structure.
The reactor cavity and fuel transfer canal are only filled with borated water during refueling outages, which occurs at each unit approximately 1 month out of every 18 months (about 5% of the operating cycle) since the Salem units perform refueling outages every 18 months. By contrast the Unit 1 FHB assessment assumed continuous borated water exposure for 70 years with a resulting depth of degradation of 1.3 inches. Therefore, the exposure duration of borated water on the reactor cavity and fuel transfer canal concrete is approximately 5% of the 70 year duration used as input to the Unit 1 FHB assessment.
Therefore, the expected depth of concrete degradation on the reactor cavity and fuel transfer canal concrete will be substantially less (0.29 inches) than the 1.3 inches predicted in the Unit 1 FHB assessment.
These insignificant depths of degradation will not approach the reinforcing steel. Therefore, as demonstrated in the Unit 1 FHB analysis, there would be insignificant degradation on the reinforcing steel in the reactor cavity and fuel transfer canal reinforced concrete structure.
In summary, based on reinforced concrete testing, reactor cavity and fuel transfer canal concrete degradation due to the borated water leakage will be insignificant.
Therefore, Salem has concluded that the leakage associated with the reactor cavity and fuel transfer canal liner has no impact on the intended function of these structures.
Page 32 of 38 LR-N10-0165 Enclosure A RAI B.2.1.33-3

==Background:==

The LRA states that groundwater intrusion has been observed through seismic expansion joints, concrete construction joints, and expansion and shrinkage cracks in the concrete.
Underground reinforced concrete structures and structures in contact with raw water at SNGS are subject to an aggressive environment.
Groundwater and raw water chemistry results in 2008 and 2009 indicate chloride levels up to 15,000 ppm that exceeds the GALL Report threshold limit for chlorides 
(< 500 ppm). The applicant stated that inspection of below-grade structures will be done when exposed during plant excavations done for construction or maintenance activities.
The LRA states that the structures monitoring program has been enhanced to require periodic sampling, testing, and analysis of groundwater chemistry for pH, chlorides, and sulfates, and assessing its impact on buried structures.
Also the LRA states that the service water intake structure will be monitored to provide a bounding condition and indicator of the likelihood of concrete degradation for inaccessible portions of concrete structures.
Issue: As noted in the LRA, there are several subgrade exterior walls at SNGS that have evidence of past or present groundwater penetration.
During the on-site audit, the applicant was asked if they had any plans for inspections of inaccessible reinforced concrete areas prior to the period of extended operation to confirm the absence of concrete degradation.
The applicant responded that they did not and that operating experience indicates that there is no evidence of corrosion appearing on the interior surfaces of the concrete structures having inaccessible exterior surfaces.
ACI 349.3R-96 recommends an inspection frequency of ten years for below-grade structures.
It was noted that the thickness of some of these walls however may be on the order of four feet. Since the applicant does not have plans for inspections of inaccessible areas, the groundwater is aggressive, there have been several incidences of groundwater penetration into the structures, and the interior of the walls may not indicate the condition of the exterior walls, it is unclear to the staff that this is an adequate approach to managing aging of inaccessible concrete structures subjected to aggressive groundwater.
Request: a. Provide locations where groundwater test samples were/are taken relative to safety related and important-to-safety embedded concrete walls and foundations and provide historical results (i.e., pH, chloride content, and sulfate content)including seasonal variation of results.b. In locations adjacent to embedded reinforced concrete structures where chloride levels exceed limits in GALL Report, provide any plans for inspections, or if no inspections or coring of concrete is planned to evaluate condition of structures (e.g., presence of steel corrosion or determination of chloride profiles), provide a basis to demonstrate that the current level of chlorides in the groundwater is not causing structural degradation of embedded walls or foundations.
Page 33 of 38 LR-N10-0165 Enclosure A The staff needs the information to confirm that the potential effect of aging of the reinforced concrete due to presence of high chloride levels will be adequately managed so that the intended function of impacted structural members will be maintained consistent with the current licensing basis for the period of extended operation as required by 10 CFR 54.21 (a)3.PSEG Response: a. The following table provides well locations and tabulates the groundwater test results.Salem Groundwater Sampling Sample Chlorides Sulfates Location Well No. pH Date ppm ppm Unit 1 -Containment Strt -soutall AC 1/30/09 -8.36 32.6 Structure 
-south wall Unit 1 -Fuel Handling AM 1/8/09 6.83 6.0 26.8 Building -south wall Unit 1 -Fuel Handling M 1/17/08 6.70 43.3 23.7 Building -west wall M 1/27/09 7.00 125 33 M 11/19/09 -83.2 21.6 N 1/17/08 5.92 7.0 93.5 Unit 1 -Fuel Handling Building -south wall N 1/8/09 5.94 4200 1048 N 11/19/09 -6.2 65 K 1/17/08 6.96 2634 483 Salem -Outside the construction cofferdams K 1/19/09 7.02 2100 239 BA 4/08 6.97 1124 1.21 Unit 2 -Fuel Handling Temporary 4/11/10 80 -Building -south wall Wells 3, 4 & 5 Unit 2 -Fuel Handling Temporary 4/11/10 Building -west wall Well 1 Auxiliary Building -north Temporary 4/11/10 wall Well 2 Aggressive Environment 
-limits per the GALL report <5.5 >500 >1500 The location of the safety related structures at Unit 1 and 2 are surrounded by the underground cofferdams used during construction, as shown on Salem UFSAR Figure 3.8-56. Wells AC, AM, M, N and the Temporary Wells are located within the Page 34 of 38 LR-N10-0165 Enclosure A these cofferdams at a sampling depth of approximately 20 feet and these wells are within 30 feet of the associated building walls. Well K is north of Salem Unit 2 Containment Structure, west of the Fire Pump House, outside of the cofferdams at a sampling depth of approximately 80 feet and is monitoring the Vincentown formation.
Well BA is south of Salem Unit 1 Containment Structure, outside of the cofferdams near the Main Fuel Oil Storage Tank and adjacent to the Shoreline Protection and Dike.Salem power-block structures, except for the Turbine building, were constructed in an area excavated down to the Vincentown Formation, approximately 70-80 feet below grade level and encircled with a system of below grade cellular cofferdams, as discussed in Salem UFSAR Section 3.8.1.6.8.7 (Construction) and shown on Figure 3.8-56. The underground walls and foundations were encased in lean concrete up to the top of the cofferdams, approximately 20 feet below grade level. This precludes routine interaction with the normal groundwater which minimizes exposure to chlorides.
The seasonal variation of using deicing salts at walkways and roadways, within the cofferdam area, has resulted in chlorides above 500 ppm on one occasion, as shown on the table above.The flow of the Delaware River at the Salem site is dominated by the tidal flow from the Delaware Bay, with the tidal flow exceeding the river runoff flow resulting in brackish water. The chlorides in the river are tide dependent as high tide would have the higher chloride levels when the water from the bay flows farther up the river. The Salem NJPDES Permit Renewal Application, February 1, 2006 lists'the river water sulfates as 387 ppm and the ph range is 6.7 -8.3. The Salem UFSAR Section 2.4.12 (Environmental Acceptance of Effluents) discusses the river salinity to be variable and averages from 10 to 15 ppt which converts to chlorides of 5500 to 8300 ppm. The LRA section 3.5.2.2.2.4.2 (Aging Management of Inaccessible Areas for Group 6 Structures) on page 3.5-49, inadvertently listed the river water chlorides as variable and ranging from 10,000 to 15,000 ppm. This was the salinity or salt concentration range. The correct chloride concentration range is 5500 to 8300 ppm.b. The enhanced Structures Monitoring Program is adequate to manage the potential effect of aging of embedded walls and foundations for the elevated levels of chlorides in the groundwater.
The Structures Monitoring Program will inspect the Service Water Intake wall splash zones exposed to waves and river tide changes as a leading indicator for the condition of below-grade embedded walls and foundations.
The use of Service Water Intake splash zones as a leading indicator to identify potential degradation of below-grade embedded walls and foundations provides reasonable assurance that the degradation of embedded walls and foundations will be detected before a loss of an intended function.The groundwater and river water samples as listed above show pH values are above 5.5 indicating a non-aggressive environment, and sulfates are lower than 1500 ppm and would indicate a non-aggressive environment; but chlorides exceed 500 ppm which indicates an aggressive environment.
The river water has higher chloride levels than the groundwater, therefore, the Service Water Intake splash zones are exposed to higher chloride levels than embedded walls and foundations.
Page 35 of 38 LR-N1O-0165 Enclosure A ACI 222R-01 "Protection of Metals in Concrete Against Corrosion" at the end of Section 2.2.5 states parts of a structure in the splash zone experiences particularly aggressive conditions.
The corrosion of steel in concrete is an electrochemical process and the continuous wetting and drying at the splash zones promotes macro-cell formation in the concrete allowing corrosion to occur. Also, high salt levels arise by salt water being transported via capillary action upward through the concrete cover and being concentrated due to evaporation in the splash zone. Therefore, the intake structure splash zone is a leading indicator for the condition of all below-grade embedded walls and foundations.
The intake structure splash zones are presently inspected by the Structures Monitoring Program with an inspection frequency of not greater than 5 years per Maintenance Rule requirements.
The corrosion of concrete reinforcing bars due to penetration and concentration of chloride ions is accompanied by an increase in volume of the steel as it is converted to rust products.
The volume increase can cause cracking, spalling, and delamination of the concrete that can be visible in the form of loss of concrete material, rust spots and stains, and cracks in the concrete cover along the reinforcing bars. Visual inspection conducted in accordance with Structural Monitoring Program would detect such conditions before the loss of an intended function.
In the event the inspection identifies significant concrete degradation at the Service Water Intake Structure, the Structural Monitoring Program and the Corrective Action Program require an evaluation of the condition to include applicability to inaccessible portions of the other structures and determine if excavation for inspection of concrete is warranted.
The embedded walls and foundations of the safety related structures surrounded by the underground cofferdams are encased in lean concrete as shown on USFAR Figure 3.8-56. This precludes routine interaction with the normally aggressive groundwater which minimizes exposure to chlorides.
The seasonal variation of using deicing salts at walkways and roadways, within the cofferdam area, has resulted in chlorides above 500 ppm on one occasion, as shown on the table above.Since 2000 there have been five inspection reports for Unit 1 and Unit 2 Service Water Intake Structures.
These inspections did not identify signs of distress due to aggressive chemical attack or corrosion of embedded steel. Also, based on review of past excavations of below grade embedded walls, the concrete was found to be in good condition.
The most recent excavations included the Unit 1 Service Water Intake Structure, Fuel Handling Building, and the Containment Structure during the Salem Unit 1 refueling outage in April 2010.The Service Water Intake Structure concrete embedded walls and foundation are comparable to all other safety related embedded walls and foundations since the same concrete mix design was used during site construction.
Also, the Service Water Intake Structure is designed with a 2 inch concrete cover which is equal to or less than the concrete cover at all other safety related structures.
Therefore, the Service Water Intake Structure splash zone inspections are considered to be an acceptable leading indicator to identify potential degradation of safety related below-grade embedded walls and foundations before a loss of an intended function.Page 36 of 38 LR-N10-0165 Enclosure A RAI B.2.1.33-4

==Background:==

IN GALL Report AMP XI.S6, program elements 3 and 4 state that for each structure/aging effect combination the specific parameters monitored or inspected are selected to ensure that the aging degradation leading to loss of intended function will be detected and quantified before there is a loss of intended function.Issue: As a result of the field walk-down with the applicant's technical staff on February 12, 2010, the staff noticed minor indications of degradation in several areas (e.g., cracking, efflorescence, leaching, and water). At Salem Unit 1 Auxiliary Building Elevation 64 (below ground water level) there was evidence of water in-leakage through the wall and the area was roped off as an exclusion zone. The applicant was asked about this and informed the staff that the source of the contamination was from in-leakage of groundwater and that the groundwater had picked up the contamination external to the wall.Request: Provide information on how the in-leakage of contaminated groundwater will be addressed under your corrective action program.The staff needs the above information to confirm that the effects of aging such as noted above will be adequately managed so that the intended function of impacted structural members will be maintained consistent with the current licensing basis for the period of extended operation as required by 10 CFR 54.21(a)(3).
PSEG Response: In-leakage of contaminated groundwater into the Auxiliary Building was documented and evaluated under the corrective action program. The groundwater in-leakage has been identified at shrinkage cracks in the below-grade embedded concrete wall. A corrective action report was generated to identify the water intrusion on 10/2009. As part of the corrective action process a qualified structural engineer conducted an initial inspection to evaluate and document the condition.
The structural engineer concluded that the current condition does not adversely impact the structure's intended function, and the current in-leakage will not impact the structure's safety function.
A past corrective action was to inject a sealant into the cracks in this concrete wall. The sealant has deteriorated and is now seeping out of the concrete cracks. Proper housekeeping was used to eliminate standing water to ensure the condition does not create an environment that will promote deterioration of structural members. This below-grade embedded concrete wall shows efflorescence or other mineral deposits.
The crack area is presently in the corrective action program to be cleaned so a detailed engineering inspection can be performed to ensure long term aging issues are identified and any other required corrective actions can be performed.
Page 37 of 38 LR-N10-0165 Enclosure A The Structures Monitoring Program is used to monitor for the effects of groundwater intrusion to ensure the condition does not result in corrosion of rebar, embedded steel, floor mounted component support members, and anchors with the potential of adverse impact on their structural integrity.
The Structures Monitoring Program includes an enhancement to perform a chemical analysis of ground or surface water in-leakage when there is significant in-leakage or there is reason to believe the in-leakage may be damaging concrete elements or reinforcing steel, as shown on LRA page A-27. In summary, implementation of the enhanced Structures Monitoring Program is considered to provide reasonable assurance the aging effect associated with groundwater intrusion will be adequately managed through the period of extended operation.
Page 38 of 38 LR-N10-0165 Enclosure B Enclosure B Salem Generating Station Unit 1 and Unit 2 License Renewal Application (LRA) Appendix A and Appendix B Program Description Updates Note: To facilitate understanding, portions of the original LRA Appendix A and Appendix B Program Descriptions have been repeated in this Enclosure, with revisions indicated.
Existing LRA text is shown in normal font. Changes are highlighted with bolded italics for inserted text and strikethroughs for deleted text.Page 1 of 7 LR-N10-0165 Enclosure B This Enclosure contains portions of LRA Appendix A and Appendix B program descriptions affected by the RAI responses in this package. Some pre-existing text is repeated here to provide context for the changes. The existing LRA text is formatted in normal font; new text is bold and italicized; deleted text is indicated with strikethroughs.
The Appendix A and B changes for a given program are presented together, followed by the next program, etc.A.2.1.28 ASME Section XI, Subsection IWE Starting with the third paragraph, Section A.2.1.28 of the LRA is modified as follows: The ASME Section Xl, Subsection IWE aging management program will be enhanced to include: 1. Inspection of a sample of the inaccessible liner covered by insulation and lagging prior to the period of extended operation and every 10 years thereafter.
Should unacceptable degradation be found additional insulation will be removed as necessary to determine extent of condition in accordance with the corrective action process.Prior to the period of extended operation* The samples shall include 57 randomly selected containment liner insulation panels per unit.* The examination will be performed by either removing the containment liner insulation panels and performing a visual inspection, or by using a pulsed eddy current (PEC) remote inspection, with the containment liner insulation left in place, to detect evidence of loss of material.
If evidence of loss of material is detected using PEC, the containment liner insulation panel will be subsequently removed to allow for visual and UT examinations During the period of extended operation" One containment liner insulation panel will be selected, at random, for removal from each quadrant, during each of the three Periods in an Inspection Interval.
Therefore, a total of 12 containment liner insulation panels will be selected, in each unit, during each ten year Inspection Interval, to allow for examination of the containment liner behind the containment liner insulation.
* The randomly selected containment liner insulation panels in each quadrant will not include containment liner insulation panels previously selected.2. Visual inspection of 100 % of the moisture barrier, at the junction between the containment concrete floor and the containment liner, will be performed in accordance with ASME Section Xl, Subsection IWE program Page 2 of 7 LR-N10-0165 Enclosure B requirements, to the extent practical within the limitation of design, geometry, and materials of construction of the components.
The bottom edge of the stainless steel insulation lagging will be trimmed, if necessary, to perform the moisture barrier inspections.
This inspection will be performed prior to the period of extended operation, and on a frequency consistent with IWE inspection requirements thereafter.
Should unacceptable degradation be found, corrective actions, including extent of condition, will be addressed in accordance with the corrective action process.As a follow-up to inspections performed during the 2009 refueling outage, the following specific corrective actions will be performed on Unit 2 prior to entry into the period of extended operation:
* Examine the accessible 3/4" knuckle plate. If corrosion is observed to extend below the surface of the moisture barrier, excavate the moisture barrier to sound metal below the floor level and perform examinations as required by IWE.* Perform remote visual inspections, of the six capped vertical leak chase channels, below the containment floor to determine extent of condition.
* Remove the concrete floor and expose the 1/4" containment liner plate (floor) for a minimum of two of the vertical leak chase channels with holes. Perform examination of exposed 1/4" containment liner plate (floor) as required by IWE. Additional excavations will be performed, if necessary, depending upon conditions found at the first two channels." Remove 1/2" containment liner insulation panels, adjacent to accessible areas where there are indications of corrosion, to determine the extent of condition of the existing corroded areas of the containment liner plate.* Perform augmented examinations of the areas of the 1/2" containment liner plate behind insulation panels, where loss of material was previously identified, in accordance with IWE-2420.* Examine 100% of the moisture barrier in accordance with IWE-2310 and replace or repair the moisture barrier to meet the acceptance standard in IWE-3510.As a follow-up to inspections performed during the 2010 refueling outage, the following specific corrective actions will be performed on Unit 1 prior to entry into the period of extended operation:
* Perform augmented examinations of the 3/4" containment liner (knuckle plate) at 78' elevation in accordance with IWE-2420.Page 3 of 7 LR-N10-0165 Enclosure B* Perform augmented examinations of the areas of the 1/2" containment liner plate behind insulation panels, where loss of material was previously identified, in accordance with IWE-2420.* Remove 1/2" containment liner insulation panels, adjacent to accessible areas where there are indications of corrosion, to determine the extent of condition of the existing corroded areas of the containment liner plate.These enhancements will be implemented prior to the period of extended operation, with the inspections performed in accordance with the schedule described above.B.2.1.28 ASME Section Xl, Subsection IWE Starting at the top of LRA Appendix B page B-1 31, the Program Description of Section B.2.1.28 is modified as follows: The program will be enhanced to include inspection of a random sample of the containment liner behind the containment liner insulation prior to the period of extended operation.
The sampling plan was developed based upon guidance in EPRI TR-107514, "Age Related Degradation Inspection Method and Demonstration:
in Behalf of Calvert Cliffs Nuclear Power Plant License Renewal Application".
The population size of containment liner insulation panels in each unit is about 264 panels. A sample size of 57 will meet the statistical requirements of a 95% confidence level that 95% of the containment liner plate behind the containment liner insulation meets the acceptance criteria of IWE-3500.
The samples will be randomly selected.
If acceptance criteria defined in IWE-3500 is not satisfied, the sampling plan will be modified as recommended in EPRI TR-107514.
Also, based on the satisfactory completion of the above sample plan, a reduced sample size will be randomly selected and examined each Containment Inservice Inspection Period during the period of extended operation.
In addition, Enhancement 
#1 on LRA page B-132 is clarified as follows: 1. Inspection of a sample of the inaccessible liner covered by insulation and lagging prior to the period of extended operation and every 10 years thereafter.
Should unacceptable degradation be found additional insulation will be removed as necessary to determine extent of condition in accordance with the corrective action process. Program Elements Affected:
Scope of Program (Element 1)Prior to the period of extended operation (PEO)* The samples shall include 57 randomly selected containment liner insulation panels per unit.* The examination will be performed by either removing the containment liner insulation panels and performing a visual inspection, or by using a pulsed eddy current (PEC) remote inspection, with the containment liner insulation left in place, to detect evidence of loss of material.
If evidence of loss of material Page 4 of 7 LR-N10-0165 Enclosure B is detected using PEC, the containment liner insulation panel will be subsequently removed to allow for visual and UT examinations.
During the period of extended operation* One containment liner insulation panel will be selected, at random, for removal from each quadrant, during each of the three Periods in an Inspection Interval.
Therefore, a total of 12 containment liner insulation panels will be selected, in each unit, during each ten year Inspection Interval, to allow for examination of the containment liner behind the containment liner insulation.
* The randomly selected containment liner insulation panels in each quadrant will not include containment liner insulation panels previously selected.A.2.1.29 ASME Section Xl, Subsection IWL LRA Appendix A is revised as follows:* Section A.1.1 (NUREG-1 801 Chapter XI Aging Management Programs), item #29 (ASME Section XI, Subsection IWL (Section A.2.1.29), on page A-6, is revised as follows: 29. ASME Section X1, Subsection IWL (Section A.2.1.29) 
[Existing 
-Requires Enhancement]
* Section A.2.1.29 (ASME Section X1, Subsection IWL), on page A-24, the following is added: The ASME Section XI, Subsection IWL, aging management program will be enhanced to include: 1. Examination and acceptance criteria in accordance with the guidance contained in ACI 349.3R.B.2.1.29 ASME Section Xl, Subsection IWL LRA Appendix B, Section B.2.1.29 is revised as follows: Section B.2.1.29 (ASME Section X1, Subsection IWL), on page B-136, the following change is made to the section on Enhancements:
Enhancements Ne~o. Prior to the period of extended operation, the following enhancement will be implemented in the program elements: 1. Examination and acceptance criteria in accordance with the guidance contained in ACI 349.3R. Program Elements Affected:
Acceptance Criteria (Element 6)Page 5 of 7 LR-N1 0-0165 Enclosure B* Section B.2.1.29 (ASME Section Xl, Subsection IWL), on page B-138, the following change is made to the

== Conclusion:==

Conclusion The ex-istig enhanced ASME Section Xl, Subsection IWL, aging management program will provides reasonable assurance that the identified aging effects are adequately managed so that the intended functions of components within the scope of license renewal will be maintained consistent with the current licensing basis during the period of extended operation.
A.2.1.33 Structures Monitoring Program The fifth Enhancement to the Structures Monitoring Program, as described in LRA Section A.2.1.33, on pages A-26 and A-27 is modified as follows: 5. Require the following actions related to the spent fuel pool liner: a. Perform periodic structural examination of the Fuel Handling Building per ACI 349.3R to ensure structural condition is in agreement with the analysis.b. Monitor telltale leakage and inspect the leak chase system to ensure no blockage.c. Test water drained from the seismic gap for boron, chloride, and sulfate concentrations; and pH.d. Perform a shallow core sample in the Spent Fuel Pool wall where previous inspections have shown ingress of borated water through the concrete.
The core sample will be examined for degradation from borated water.In addition, a new Enhancement 
(#15) is added to page A-27 of LRA Appendix A, Section A.2.1.33, as follows: 15. When the reactor cavity is flooded up, Salem will periodically monitor the telltales associated with the reactor cavity and refueling canal for leakage. If telltale leakage is observed, then the pH of the leakage will be measured to ensure that concrete reinforcement steel is not experiencing a corrosive environment.
In addition, Salem will periodically inspect the leak chase system associated with the reactor cavity and refueling canal to ensure the telltales are free of significant blockage.
Salem will also inspect concrete surfaces for degradation where leakage has been observed, in accordance with this Program.Page 6 of 7 LR-N1 0-0165 Enclosure B B.2.1.33 Structures Monitoring Program Consistent with the changes made to LRA Section A.2.1.33, the fifth Enhancement to the Structures Monitoring Program, as described in LRA Section B.2.1.33, on page B-153 is modified as follows: 5. Require the following actions related to the spent fuel pool liner: a. Perform periodic structural examination of the Fuel Handling Building per ACI 349.3R to ensure structural condition is in agreement with the analysis.b. Monitor telltale leakage and inspect the leak chase system to ensure no blockage.c. Test water drained from the seismic gap for boron, chloride, and sulfate concentrations; and pH.d. Perform a shallow core sample in the Spent Fuel Pool wall where previous inspections have shown ingress of borated water through the concrete.
The core sample will be examined for degradation from borated water.Also consistent with the changes made to LRA Section A.2.1.33, a new Enhancement
(#15) is added to page B-154 of LRA Appendix B, Section A.2.1.33, as follows: 15. When the reactor cavity is flooded up, Salem will periodically monitor the telltales associated with the reactor cavity and refueling canal for leakage. If telltale leakage is observed, then the pH of the leakage will be measured to ensure that concrete reinforcement steel is not experiencing a corrosive environment.
In addition, Salem will periodically inspect the leak chase system associated with the reactor cavity and refueling canal to ensure the telltales are free of significant blockage.
Salem will also inspect concrete surfaces for degradation where leakage has been observed, in accordance with this Program.Page 7 of 7 LR-N10-0165 Enclosure C A.5 License Renewal Commitment List The following table identifies modifications made to license renewal commitments 28, 29 and 33 as a result of the responses to the RAIs contained in this package. Pre-existing text is formatted in normal font; new text is bold and italicized; deleted text is indicated with strikethroughs.
The specific RAIs that led to the commitment modifications are listed in the "SOURCE" column adjacent to beginning of the new text. Any other actions described in this submittal represent intended or planned actions. They are described for the NRC's information and are not regulatory commitments.
UFSAR PROGRAM SUPPLEMENT ENHANCEMENT OR NO. OR COMMITMENT LOCATION IMPLEMENTATION SOURCE OR TOPIC (LRA APP. A) SCHEDULE 28 ASME Section XI, ASME Section XI, Subsection IWE is an existing program A.2.1.28 Program to be enhanced Section Subsection IWE that will be enhanced to include: prior to the period of B.2.1.28 extended operation.
: 1. Inspection of a sample of the inaccessible liner Inspection Schedule covered by insulation and lagging once prior to the identified in Commitment period of extended operation and every 10 years thereafter.
Should unacceptable degradation be found additional insulation will be removed as necessary to determine extent of condition in accordance with the corrective action process.Prior to the period of extended operation Salem letter LR-N10-0165" The samples shall include 57 randomly RAI B.2.1.28-2 selected containment liner insulation panels per unit." The examination will be performed by either removing the containment liner insulation panels and performing a visual inspection, or by using a pulsed eddy current (PEC) remote inspection, with the containment liner insulation left in place, to detect evidence of loss of material.
If evidence of loss of material is detected using PEC, the containment liner insulation panel will be Page 1 of 7 LR-N10-0165 Enclosure C UFSAR SUPPLEMENT ENHANCEMENT OR NO. PROGRAM COMMITMENT LOCATION IMPLEMENTATION SOURCE OR TOPIC (LRA APP. A) SCHEDULE I subsequently removed to allow for visual and UT examinations During the period of extended operation" One containment liner insulation panel will be selected, at random, for removal from each quadrant, during each of the three Periods in an Inspection Interval.
Therefore, a total of 12 containment liner insulation panels will be selected, in each unit, during each ten year Inspection Interval, to allow for examination of the containment liner behind the containment liner insulation." The randomly selected containment liner insulation panels in each quadrant will not include containment liner insulation panels previously selected.2. Visual inspection of 100 % of the moisture barrier, at the junction between the containment concrete floor and the containment liner, will be performed in accordance with ASME Section XI, Subsection IWE program requirements, to the extent practical within the limitation of design, geometry, and materials of construction of the components.
The bottom edge of the stainless steel insulation lagging will be trimmed, if necessary, to perform the moisture barrier inspections.
This inspection will be performed prior to the period of extended operation, and on a frequency consistent with IWE inspection requirements thereafter.
Should unacceptable degradation be found, corrective actions, including extent of condition, will be addressed in accordance with the corrective action process.Page 2 of 7 LR-N10-0165 Enclosure C UFSAR PROGRAM SUPPLEMENT ENHANCEMENT OR NO. OR COMMITMENT LOCATION IMPLEMENTATION SOURCE OR TOPIC (LRA APP. A) SCHEDULE I As a follow-up to inspections performed during Salem letter the 2009 refueling outage, the following LR-N10-0165 specific corrective actions will be performed on RAI B.2.1.28-1 Unit 2 prior to entry into the period of extended operation:
* Examine the accessible 3/4" knuckle plate.If corrosion is observed to extend below the surface of the moisture barrier, excavate the moisture barrier to sound metal below the floor level and perform examinations as required by IWE.* Perform remote visual inspections, of the six capped vertical leak chase channels, below the containment floor to determine extent of condition." Remove the concrete floor and expose the 1/4" containment liner plate (floor) for a minimum of two of the vertical leak chase channels with holes. Perform examination of exposed 1/4" containment liner plate (floor) as required by IWE. Additional excavations will be performed, if necessary, depending upon conditions found at the first two channels.* Remove 1/2" containment liner insulation panels, adjacent to accessible areas where there are indications of corrosion, to determine the extent of condition of the existing corroded areas of the containment liner plate.* Perform augmented examinations of the areas of the 1/2" containment liner plate behind insulation panels, where loss of material was previously identified, in Page 3 of 7 LR-N10-0165 Enclosure C UFSAR SUPPLEMENT ENHANCEMENT OR NO. PROGRAM COMMITMENT LOCATION IMPLEMENTATION SOURCE ORTOPIC (LRA APP. A) SCHEDULE accordance with IWE-2420.Examine 100% of the moisture barrier in accordance with IWE-2310 and replace or repair the moisture barrier to meet the acceptance standard in IWE-3510.As a follow-up to inspections performed during the 2010 refueling outage, the following specific corrective actions will be performed on Unit 1 prior to entry into the period of extended operation: " Perform augmented examinations of the 3/4" containment liner (knuckle plate) at 78'elevation in accordance with IWE-2420.* Perform augmented examinations of the areas of the 1/2" containment liner plate behind insulation panels, where loss of material was previously identified, in accordance with IWE-2420." Remove 1/2" containment liner insulation panels, adjacent to accessible areas where there are indications of corrosion, to determine the extent of condition of the existing corroded areas of the containment liner plate.29 ASME Section X1, rvxi.ting progrAMi.. 
:o.i.td..
A.2.1.29 Section Subsection IWL B.2.1.29 ASME Section Xl, Subsection IWL, is an existing Program to be Salem letter program that will be enhanced to include: enhanced prior to the LR-N1O-0165 period of extended RAI B.2.1.29-1
: 1. Examination and acceptance criteria in operation.
accordance with the guidance contained in ACI 349.3R.Page 4 of 7 LR-N10-0165 Enclosure C UFSAR PROGRAM SUPPLEMENT ENHANCEMENT OR NO. OR COMMITMENT LOCATION IMPLEMENTATION SOURCE OR TOPIC (LRA APP. A) SCHEDULE I 33 Structures Monitoring Program Structures Monitoring is an existing program that will be enhanced to include: 1. Additional structures and components as described in A.2.1.33.2. Concrete structures will be observed for a reduction in equipment anchor capacity due to local concrete degradation.
This will be accomplished by visual inspection of concrete surfaces around anchors for cracking, and spalling.3. Clarify that inspections are performed for loss of material due to corrosion and pitting of additional steel components, such as embedments, panels and enclosures, doors, siding, metal deck, and anchors.4. Require inspection of penetration seals, structural seals, and elastomers, for degradations that will lead to a loss of sealing by visual inspection of the seal for hardening, shrinkage and loss of strength.5. Require the following actions related to the spent fuel pool liner: a. Perform periodic structural examination of the Fuel Handling Building per ACI 349.3R to ensure structural condition is in agreement with the analysis.b. Monitor telltale leakage and inspect the leak chase system to ensure no blockage.c. Test water drained from the seismic gap for boron, chloride and sulfate concentrations; and pH.d. Perform a shallow core sample in the Spent Fuel Pool wall where previous inspections have shown ingress of borated water through the concrete.The core sample will be examined for degradation from borated water.A.2.1.33 Program to be enhanced prior to the period of extended operation.
Section B.2.1.33 Salem letter LR-N10-0165 RAI B.2.1.33-1 Page 5 of 7 LR-N10-0165 Enclosure C UFSAR SUPPLEMENT ENHANCEMENT OR NO. PROGRAM COMMITMENT LOCATION IMPLEMENTATION SOURCE OR TOPIC (LRA APP. A) SCHEDULE 6. Require monitoring of vibration isolators, associated with component supports other than those covered by ASME Xl, Subsection IWF.7. Add an Examination Checklist for masonry wall inspection requirements.
: 8. Parameters monitored for wooden components will be enhanced to include: Change in Material Properties, Loss of Material due to Insect Damage and Moisture Damage.9. Specify an inspection frequency of not greater than 5 years for structures including submerged portions of the service water intake structure.
: 10. Require individuals responsible for inspections and assessments for structures to have a B.S.Engineering degree and/or Professional Engineer license, and a minimum of four years experience working on building structures.
: 11. Perform periodic sampling, testing, and analysis of ground water chemistry for pH, chlorides, and sulfates on a frequency of 5 years. Groundwater samples in the areas adjacent to Unit 1 containment structure and Unit 1 auxiliary building will also be tested for boron concentration.
: 12. Require supplemental inspections of the affected in scope structures within 30 days following extreme environmental or natural phenomena (large floods, significant earthquakes, hurricanes, and tornadoes).
: 13. Perform a chemical analysis of ground or surface water in-leakage when there is significant in-leakage or there is reason to believe that the in-leakage may be damaging concrete elements or reinforcing steel.14. Implementing procedures will be enhanced to include additional acceptance criteria details specified in ACI 349.3R-96.
Page 6 of 7 LR-N130-0165 Enclosure C UFSAR SUPPLEMENT ENHANCEMENT OR NO. PROGRAM COMMITMENT LOCATION IMPLEMENTATION SOURCE OR TOPIC (LRA APP. A) SCHEDULE 15. When the reactor cavity is flooded up, Salem Salem letter will periodically monitor the telltales associated LR-N1lO-0165 with the reactor cavity and refueling canal for RAI B.2.1.33-2 leakage. If telltale leakage is observed, then the pH of the leakage will be measured to ensure that concrete reinforcement steel is not experiencing a corrosive environment.
In addition, Salem will periodically inspect the leak chase system associated with the reactor cavity and refueling canal to ensure the telltales are free of significant blockage.
Salem will also inspect concrete surfaces for degradation where leakage has been observed, in accordance with this Program.Page 7 of 7 LR-N 10-0165 Enclosure D Enclosure D MPR Report MPR-2613, "Salem Generating Station Fuel Handling Building-Evaluation of Degraded Condition," Revision 3 (305 pages)[PSEG Nuclear LLC VTD Number 326367], associated with response to RAI B.2.1.33-1 for Salem Generating Station Unit 1 and Unit 2 License Renewal Application USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES VTD 326367 000 1 PRINTED 20060531 f-r rlj- i VENDOR INFORMATION OS PSEG NUCLEAR LLC VTD NUMBER: 3!2" l ACTIVE Approved Documentation El APCP Approved, Pending Change Package (May only be used for Rev. 1)El CAN Canceled, Not Required[] VOID No longer applicable, superseded by:_Discipline Selection[: Electrical r, l&C [-i Mechanical Other Specify: CIVIL Safety Related: l Yes El No UNIT APPLICABILITY C Salem 1 5? Salem Common 0- Hope Creek & Salem Ml Salem 2 Hope Creek -' PSEG Nuclear LLC Eli Salem 3 System I Title: F"LL/_/..
dIDL/ a1,/z &/AIz , SAP Sys Code: _____Vendor Name: / Aq Vendor Code: Vendor No.: A 4 13 Vendor Category: (Category Codes are listed in DCRMS.)Purchase Order No.Material Master. _Originator 
.A raI, aos _ Dept: 12-SI6/1AJ-W/
Group: tIvtL Date: Ext: __910 If changes are made to this form, initial and date the change and document in the revision summary.Nuclear Common Page 11 of 14 Rev.0 U ORM-2 REVIOlN 

==SUMMARY==
PSEi NC CC-AP.ZZ-OO'@ 
*G NUCLEAR LLC VTD No. -3:26361 PS DOC R SUPERVISOR' DISCIPLINE REV REV DATE EVALUATO INTERFACE REVISION DESCRIPTION V14'SIGN & PRINT NAME< C---i C 0I') r 3OF tO)P HO'(A U P14 z o00 0Oz 0r-0*----rln z L/)H (I zP>U)Nuclear Common Page 12 of 14 Rev. 0'L  
:0*MPR ASSOCIAVS INC, ENGIN ERS MPR-2613 Revision 3 February 2009 (PSEG Nuclear VTD 326367)Salem Generating Station Fuel Handling Building Evaluation of Degraded Condition QUALITY ASSURANCE DOCUMENT This document has been prepared, reviewed and approved in accordance with the Quality Assurance requirements of 10CFR50 Appendix B, as specified In the MPR Quality Assurance Manual.Prepared for PSEG Nuclear LLC Salem Generating Station P.O. Box 236 Hancocks Bridge, NJ 08038 
-Mpi ASSOCIATES INC:N'GI NE ERS Salem Generating Station Fuel Handling Building Evaluation of Degraded Condition MPR-2613 Revision 3 (PSEG Nuclear VTD 326367)February 2009 QUALITY ASSURANCE DOCUMENT This document has been prepared, reviewed and approved In accordance with the Quality Assurance requirements Of 10CFR50, Appendix B, as specified in the MPR Quality Assurance Manual.* ~~Prepared by.M e Prepared by: d/ " 9 Reviewed by: /Join .s fs -r Dr. James E. Nestesl, Jr.Reviewed &Approved by: Robert B. Keating, PE Principal Contributors John W. Simons Robert B. Keating James E. Nestell Matthew C. Frey Prepared for PSEG Nuclear LLC Salem Generating Station P. 0. Box 236* Hancocks Bridge, NJ 08038 320 KING -STREET ALEXANDRIA, VA 22314-32M 703-61W-0200 FAX:. 703-51M.0224 http:klwww,mpr.com RECORD OF REVISIONS Revision Affected Pages Description 0 All Initial Issue 1 All
* Changed terminology from "Spent Fuel Pool Building" to "Fuel Handling Building" throughout report for consistency with plant convention.
Moved the details of concrete and reinforcing steel testing to a separate report (MPR-2634).
2 All Included results and conclusions from additional concrete testing (including long-duration 
-9 months -testing) documented in Revision 1 of MPR-2634.Removed discussion of potential margin recovery from the ultimate strength method, and the associated calculation in Appendix A.Revised the discussion of degradation of embedded rebar based on additional information from the open literature and reports regarding SFP leakage at another PWR.* Updated the degradation model.0 Added an evaluation of the liner above localized degradation of the slab.* Added a "road map" to previous evaluations that form the current design basis for the FHB.* Added discussion of independent structural evaluation by an experienced concrete structural engineer.* Included a description of the mechanism for formation of the channel/telltale obstructions.
* Provided an expanded description of a construction joint.* Added an Executive Summary.3 All 0 Updated the discussion of testing conducted by MPR to reflect results from the 39-month specimens documented in Revision 2 of MPR-2634.
Updated degradation projections accordingly.
0 Updated evaluation of the liner above localized degradation of the slab.* Updated the assessment of potential rebar corrosion.
* Incorporated results of evaluations of concrete cores from the CY SFP as corroboration of degradation modes and projections.
J _____________________ 
.1.MPR-2613 Revision 3.&deg;&deg; 0 Executive Summary This report, together with a companion report (MPR-2634, Boric AcidAttack of Concrete and Reinforcing Steel, Revision 2), represent the culmination of a multi-year effort to assess the structural adequacy of the Salem Unit 1 Fuel Handling Building (FHB). The FHB, a reinforced concrete structure, has experienced and will continue to experience degradation from boric acid leakage from the Spent Fuel Pool (SFP).This report demonstrates the adequacy of the Unit I FHB through the end of plant life (70 years total). This evaluation uses results from long-duration concrete testing as well as other efforts completed after the initialassessment in 2003. These efforts have furthered the understanding of the degradation processes, supported quantification of the long-term concrete degradation rate and provided a more accurate corrosion rate for embedded rebar. This evaluation includes results from concrete cores.removed from the SFP of a decommissioned plant as corroboration of the laboratory testing and degradation postulations for the Salem FHB.CONCRETE DEGRADATION Boric acid reacts with the alkaline constituents of concrete, causing cracking and loss of bonding with the aggregate.
The reacted concrete is soft and porous and has no strength.
There is no impact on concrete strength other than an effective reduction in thickness corresponding to the depth of the corrosion layer.The rate of degradation is controlled by diffusion of boric acid into the concrete.
Results from the long-duration testing show that degradation is diffusion-controlled.
Projections of concrete degradation are made using a square root of time curve fit of the test data, including uncertainty." The slab underneath the SFP has been exposed to boric acid leakage since early in plant life. This degradation is localized to the vicinity of leaking plug welds. However, as telltales became obstructed over time, the area exposed to boric acid increased until the entire slab was exposed. Re-establishing flow in the telltales and draining the stored inventory between the liner and concrete does not fully stop this mode of degradation because! some telltales remain blocked and the leakage must migrate from the plug welds to channels with open telltales.
The-projected depth of concrete degradation in the slab is 1.3 inches assuming exposure to boric acid for 70 years.* The walls surrounding the SFP were exposed to boric acid during the time period when the telltales were fully plugged and SFP leakage accumulated in the gap between the SFP liner and the walls. Exposure of the walls to boric acid started between 1995 and 1998 when the leakage channels and telltales became blocked and extended to early 2003 when drain flow was re-established.
The projected depth of concrete degradation in the walls is 0.44 inch.MPR-2613 Revision 3 iv
* REINFORCING STEEL CORROSION Embedded reinforcing steel can potentially corrode from boric acid that migrates through the concrete.
Since the concrete cover for all walls and the slab is markedly greater than the projected depth of concrete degradation, boric acid penetration into the concrete will not reach the reinforcing steel. Therefore, the only mechanism for degradation of reinforcing steel is migration of boric acid through cracks or construction joints.Migration of boric acid through construction joints or cracks started prior to 2002, possibly as early as the 1995 to 1998 timeframe.
The leakage from the building stopped subsequent to cleaning the telltales in early 2003. This mode of degradation is not expected to recur as PSEG Nuclear has implemented multiple measures to ensure that the telltales do not become entirely blocked again.All evidence indicates that any degradation of reinforcing steel, particularly the outer reinforcing steel (i.e., the reinforcing steel of concern from a structural standpoint), is negligible.
This evidence includes the following." Laboratory studies available in the literature of corrosion of: embedded rebar from boric acid flow through cracks; and corrosion of mild steel in de-aerated boric acid solutions.
* Inspections of potential rebar degradation from flow of boric acid leakage through a concrete crack at another US PWR." No visual indications of rebar degradation in the FHB, based on an independent concrete condition assessment of the Salem Unit 1 FHB by a concrete structural engineer in accordance with ACI-201.It is conservatively estimated that the rebar has experienced a reduction in radius of less than 1 mil (0.001 inch). This is extent of degradation is negligible.
Further, the fact that the actual rebar strength is greater than the specified compensates for the predicted reduction in margin by more than a factor of 10.FHB STRUCTURAL CAPACITY The foregoing discussion shows that projected degradation through the end of plant life is minor and would have a small impact on available structural margin. However, per the current design basis analysis of the Salem FHB, the available margin is as low as 2% depending on the load case and the location.Projected degradation through the end of plant life reduces the available margin in the limiting section by less than half percentage point to 1.6%. Therefore, the design basis analysis of record remains valid even with the postulated degradation.
MPR-2613 v Revision 3 O CORROBORATION BY CY SFP CORES Overall, the CY cores corroborate the results of testing for Salem and the projections for Salem.The maximun, depth of concrete degradation in the CY cores is within that predicted using the correlation deVeloped from the Salem testing. The rebar in the CY cores exhibited no corrosion even though the upper surface of the concrete was degraded by boric acid, the concrete was cracked and, lased on the presence of secondary deposits within the concrete, there was water migration in the concrete.MPR-2613 Revision 3 vi 0 Contents 1 Introduction 
........................................................................................................ 
: 1. -1 1.1 Purpose .....................................................................................................................
1-1 1.2 Scope of Assessment 
........................................
1-1 1.3 Spent Fuel Pool Description 
....................................................................................
1-1 1.4 Background 
..............................................................................................................
1-3 2 Sum m ary .......................................................................................................
2-1 2.1 Boric Acid Attack on Concrete .................................................................................
2-1 2.2 Degradation of Salem FHB ......................................................................................
2-2 2.3 Structural Assessment 
...............................................................................................
2-3 2.4 Corroboration of Salem Assessment with Cores from CY's SFP ...........................
2-4 3 Review of Design Basis ............................................ 
................................
3-3.1 Analysis M ethod .......................................................................................................
3-1 3.2 Design Conditions 
...................................................................................................
3-1 3.3 Current Design M argin ................................................... 
.................................
3-7 4 Potential M argin Recovery ................................................................................
4-1 4.1 Reinforcing Steel Capacity.......................................................................................
4-1 4.2 Concrete Capacity .....................................................................................................
4-2 4.3 Conclusions 
....... a ....................................................................................................
4-3 5 Boric Acid Attack of Concrete and Reinforcing Steel ...............................
5-1 5.1 Chemical Reactions between Acid and Concrete ...........................
5-1 5.2 Literature Studies on Boric Acid Attack ................. 
..........
5-2 5.3 MPR (CRT) Testing of Boric Acid Attack to Support Salem FHB Evaluation 
....... 5-3 5.4 Evaluations from Another PW R ...............................................................................
5-4 5.5 Conclusions 
...............................................................................................................
5-5 6 Assessment of Potential Damage to Structure 
.........................................
6-1 6.1 Leakage Evaluation 
..................................................................................................
6-1 0 MPR-2613 Revision 3 vii 0 Contents (cont'd.)6.2 Telltale Chemistry 
.........................................
6-2 6.3 Chemical Analysis of Material Blocking Telltales/Channels 
...................................
6-3 6.4 L eakage T im eline ..................................................... 
...............................................
6-3 6.5 Independent Structural Assessment per ACI Guidelines 
.........................................
6-5 6.6 Degradation of FHB Structure from Boric Acid Exposure .....................................
6-5 6.7 C onclusions... 
..........................................................................................................
6-10 7 Assessment of Structural Adequacy ...............................................................
7-1 7.1 Concrete Degradation from Boric Acid Exposure .....................
7-1 7.2 Reinforcing Steel Degradation from Migration through Joints/Cracks 
.....................
7-2 7.3 Voided Areas beneath the Liner ...............................................................................
7-3 7.4 C onclusion 
................................................................................................................
7-4 8 Corroboration of Salem Assessment by Cores from the CY SFP ............
8-1 8.1 Evaluation of CY C ores ...........................................................................................
8-1 8.2 Comparison of CY Cores to Salem SFP Assessment 
..............................................
8-5 8.3 C onclusions 
...............................................................................................................
8-6 9 References 
.............................................
9-1 9.1 Specifications 
.............................................
9-1 9.2 R eports ................................................................ 
................................................
9-1 9.3 C alculations 
.............................................................................................. 
...............
9-2 9.4 Drawings ...............................................
9-2 9.5 T echnical Papers, ..................................................................................................
9-3 9.6 Other Documents 
........... 
................
9-3 9.6 O ther D ocu m ents .~ ~.....;-........... 
:..... :................................................ 
:............ 
;........... 
.9 -A Petrographic Examination of Concrete Cores Removed from the Conn-Yankee SFP ...................................................................................................
A-B Statistical Analysis of Rebar Yield and Tensile Strength Tests ........ B-1 C Margin Reduction from Rebar Corrosion 
...................................................
C-1 D Margin Reduction from Concrete Degradation 
....................
DI MPR-2613 viii Revision 3 Tables Table 2-1. FHIB Available Margin Based on Predicted Degradation Modes ..............................
2-4 Table 3-1. Loads and Margins for Each Load Combination 
.......................................................
3-8 Table 3-2. Limiting FHB Design Margins from the Design Basis Analysis ..............................
3-9 Table 4-1. FHB Reinforcing Steel Yield Strength Analysis Results ........................................
4-1 Table 6-1. Projected Degradation of Concrete Structure:
Local Degradation from Weld Leakage an General Degradation from Water Trapped Behind Liner ..................................
6-11 Table 6-2. Projected Degradation of Concrete Structure:
Reinforcing Steel Degradation from Seepage through Joints/Cracks 
..............................................................................
6-12 Table 7-1. FHB Available Margin Based on 0.44" General Concrete Degradation and Working Stress D esign (W SD) M ethods .................................................................................
7-2 Table 8-1. Depth of Affected Paste in CY Cores........................................................................
8-3 MPR-2613 Revision 3 ix Figures Figure 1-1. Location of Fuel Handling Buildings 
..............................................................
1-2 Figure 3-1. Stress Distribution for Working Stress Design ..................................................
3-4 Figure 3-2. Stress Distribution for Ultimate Strength Design ............................................
3-6 Figure 4-1. FHB Reinforcing Steel Yield Strength Distribution of Population Sampled ...........
4-2 Figure 6-1. Leakage Tim eline for Salem Unit 1 ..........................................................................
6-5 Figure 8-1. Core 122-As Received Core prior to Sectioning 
.......................
8-3 Figure 8-2. Core 122-Section Showing Fracture along Crack .......................................
8-4 MPR-2613 x Revision 3 IntrOduction

===1.1 PURPOSE===
This report assesses the structural adequacy of the Salem Fuel Handling Building (FHB)reinforced concrete structure after prolonged exposure of the concrete and reinforcing steel to boric acid, which has leaked from the Spent Fuel Pool (SFP).1.2 SCOPE OF ASSESSMENT This assessment evaluates the potential degradation of the FHB and its impact on the structural capacity by examining:
Conservatisms in the current Salem FHB design basis,* Results of tests, analyses, assessments and research documented in open literature that have reported the effects of boric acid on concrete and reinforcing steel, Results of evaluations of the impact of SFP leakage on the surrounding reinforced concrete structure at another PWR, Results &#xfd;of testing designed to determine the effect of boric acid on concrete and reinforcing steel,* Chemical analyses of the liquid draining from the telltales and the material that blocked the telltales, and* History. of SFP leakage at Salem Unit 1.In addition, results from petrographic examinations of concrete cores from the Connecticut Yankee Atom-ic Power Plant (CY) SFP are reviewed to corroborate the degradation modes and projections developed herein.1.3 SPENT FUEL POOL DESCRIPTION Salem Unit l and Salem Unit 2 have SFPs, which are similar in construction.
Each SFP is located in its corresponding unit's FHB. The Unit 1 FHB construction details are shown in References 9.'4.1 through 9.4.9. The FHBs are reinforced concrete structures located on the west side of the cohtainment structures, and each contains a new fuel storage pit, a spent fuel storage pool, and a ftel transfer pool. The buildings consist of reinforced concrete walls and a MPR-2613 Revision 3 1-1 0 reinforced concrete roof and foundation mat. The walls vary in thickness from between 2'-2" to 10'-0". An outline of the FHBs is shown in Figure 1-1.North no No. 2 Unit No. 1 Unit Administration Building Turbine Area Service Building-+ --+ -- Reactor Containment.I FuelI Ha ndling Auxillary BuilldingFulH n ig B n BFuel Handling Building-Building Service Water -"- Service Water Storage Tanks Storage Tanks Figure 1-1. Location of Fuel Handling Buildings MPR-2613 Revision 3 1-2 At the intersection of the concrete walls and the floor slab is a construction joint. A construction joint is formed any time unhardened concrete is placed against concrete that has become sufficiently rigid that the new concrete cannot be incorporated into the old concrete by vibration.
Because of this, additional steps are taken to insure a bond between the two lifts. For example, the old concrete is usually sand blasted and a mortar or other material is used to ensure a bond.The bond is then considered to be of sufficient stiffness that the construction joint is accepted as monolithic for the purpose of determining the relative stiffness of the elements of the building.Construction joints are prone to water migration because shrinkage and other minor cracks are likely to develop at some locations in the joint, and because the joint extends from the inside to the outside of the wall. This was evidenced by through-wall leakage at various points along the construction joints within the Salem FHB. The development of cracks sufficient to permit small amounts of water migration is not considered significant and does not affect the overall relative stiffness of the elements of the building.The SFPs and transfer pools are lined with stainless steel plate. The liner seams are fully welded. Behind the seam welds, channels are embedded in the concrete walls and slab to collect and drain any leakage through the welds. Between the seams, the liner is attached to studs embedded in the concrete of the SFP and transfer pool structures.
Leakage from the channels is collected in a series of telltales (seventeen one-inch diameter drain pipes per unit). The channels have internal plates so that a given telltale corresponds to specific seam weld locations.
The telltales are piped to a drainage canal in each unit's Sump Room.The Sump Rooms run along the West walls of the SFPs. The telltales enter the Sump Rooms at about the floor elevation to discharge into the drainage canal. The drainage canal is separated.
from the remainder of the Sump Room floor by a barrier wall approximately one foot in height.

==1.4 BACKGROUND==

On September 18, 2002, a technician working at the 78-foot elevation of the Salem Unit 1 Auxiliary Building contaminated his shoe. Investigation into the source of the contamination identified white deposits on the wall and active water inflow into the building.
Further investigation determined that water from the SFP was leaking through the concrete wall into the Auxiliary Building and into the seismic gap between the -buildings.
Also, there was evidence of leakage into the Sump Room in the FHB via a construction joint at the base of the pool.In early 2003 a videoscope of the telltales and leakage channels revealed that white deposits were present inside the telltale drains. Most of the telltale drains were completely blocked with the deposits.
Apparently, the drains had become blocked causing leakage from the pool to accumulate in the gap between the liner and the pool. As the water level in the gap increased, hydrostatic pressure forced water into construction joints and cracks. Note that fluid may weep through the seams via minor gaps or discontinuities, however these seams should not be treated as through-wall cracks with significant amount of flow passing through them.MPR-2613 Revision 3 1-3 PSEG Nuclear snaked the telltales in early 2003 using a power auger to re-establish flow in the telltales and drain the stored inventory in the gap between the SFP liner and FHB structure.
After snakingi the flow rate from the telltales increased to approximately 140 gallons/day (gpd)following the .leaning, and later decreased and held steady at 100 gpd.The initial inc iease in the drainage rate following the cleaning indicates that leakage had collected in the blocked drains and channels and possibly between the liner plate and concrete.Potential long-term exposure of the concrete structure to boric acid leakage from the SFP could degrade the structure as the acid reacts with the concrete and possibly corrodes the embedded reinforcing steel. This raises the issue of whether the potential degradation has challenged the structural adequacy of the FHB with respect to its design basis conditions.
PSEG Nuclear's strategy to address the boric acid leakage from the FHB is to: Demonstrate the structural adequacy of the FHB structure based on a combination of structural analyses and concrete testing; and* Maintain flow in the leakage channels and telltales through periodic cleanings.
This report documents the assessment of the structural adequacy of the FHB.MPR-2613 Revision 3 1-4 W 2 Summary 2.1 BORIC ACID ATTACK ON CONCRETE Formation of corrosion products is typical when concrete is exposed to acids. The corrosion layer is typically soft, cracked, and without bonding properties.
In the drying process, the corrosion layer shrinks, cracks widen, and the layer crushes easily. When a corroded layer is formed, the mechanical properties of a specimen depend primarily on the quality of the 'non-corroded core' of the specimen.The reaction between hardened cement paste or concrete and an acid solution is controlled by the diffusion of the acid into the concrete.
The rate at which the concrete (or paste) degrades decreases over time as the distance acid must diffuse through degraded concrete to reach intact concrete increases.
Degradation of concrete by acids follows Fick's Law of Diffusion formulation in which the depth of degradation varies with the square root of time. Hence, the rate of degradation decreases monotonically, approaching zero asymptotically.
The degradation rate depends on the acid.The acidic solution can also attack reinforcing steel embedded in the concrete.
Studies of reinforcing steel corrosion due to boric acid entering reinforced concrete through a crack have shown negligible reinforcing steel attack. Tests documented in open literature demonstrate that, after an exposure time of two years, reinforcing steel corrosion was limited to scarring in the area of the crack. Studies by EPRI on corrosion of steel in de-aerated boric acid determined that the corrosion ratel under de-aerated conditions (0.004 mm/year, Reference 9.5.5). Migration of boric acid through qracks and construction joints is expected to be dc-aerated; however, corrosion of reinforcing steel will be even lower due to the elevated pH as the boric acid reacts with the alkaline constituents of the concrete.Testing was performed as part of this overall assessment to determine how concrete and reinforcing steel are affected by exposure to boric acid in concentrations typical of SFP chemistry.
More specifically, the testing was conducted to understand the mechanisms of concrete and reinforcing steel degradation and degradation rates. The testing used a combination of cores from the Auxiliary Building at Salem and specimens prepared using the same concrete mix and suppliers as for the concrete used at Salem. Several test series, covering exposure times up to 39 months, were conducted.
The tests were conducted in a manner which is conservative with respect to the actual conditions expected in the FHB (e.g., boric acid bath was periodically refreshed to maintain acidic conditions).
Key insights from the testing include the following.
Boric avid reacts with the alkaline constituents of cement paste, causing cracking and loss of bonding with the aggregate.
The reacted cement paste is soft and porous and has no 0 MPR-2613 2-1 Revision 3 strength.
Fine aggregate particles are easily dislodged.
This type of degradation is consistent with concrete degradation from attack by other acids." The wicking effect at the reinforcing steel/concrete interface is minor. That is, the degradation rate of the concrete at the reinforcing steel/concrete interface is similar to the general rate of attack of concrete away from the reinforcing steel. Hence, degradation of reinforcing steel at the construction joints or cracks with boric acid migration will not spread rapidly along the steel bar; i.e., rebar degradation is localized to the vicinity of the construction joint or crack. Functionality of the reinforcing steel is therefore maintained." The structural impact of boric acid attack on the concrete is reduction of the effective area carrying loads." The rate of concrete degradation follows a square root of time formulation, which is typical for diffusion-controlled processes.
The degradation rate decreased substantially during the 39-month test series. Projected depth of affected paste through the end of plant life is 1.3 inches, including adjustment for uncertainty.

===2.2 DEGRADATION===

OF SALEM FHB The Salem FHB has degraded from prolonged exposure to boric acid due to SFP liner leakage.The degradation has likely occurred in three different modes.2.2.1 -Local Degradation from Weld Leakage Leakage through liner plug welds onto concrete or leakage from welds (seam or plug)overflowing blocked channels results in local degradation of the concrete structure, primarily the slab underneath the pool. The boric acid will attack the cement paste, weakening it and causing it to de-bond from the coarse and fine aggregate.
As the degradation progresses, a rubble bed of coarse and fine aggregate may be formed after a significant time. In essence, local degradation will create a "pothole" with sand and coarse aggregate on top of the remaining concrete.This mode of!degradation most likely initiated prior to 1995 and is ongoing. Re-establishing flow in the telltales and draining the stored inventory between the liner and concrete did not stop this mode of degradation because the leakage must still migrate from the plug welds to channels with open telltales.

====2.2.2 General====
Degradation from Water Trapped between the Liner and Concrete As the leakage channels and telltales became plugged, leakage from the SFP accumulated in the gap between the liner and the concrete on the slab and the walls. The water level in the gap increased and may have equalized with the level in the pool, at which point leakage essentially stopped and conditions in the gap become stagnant.
Degradation of the concrete is similar to that described above for weld leakage, except the degradation is widespread rather than localized.
Virtually the entire structure surrounding the pool is potentially exposed to boric acid and subject to degradation.
MPR-2613 Revision 3 2-2 The period of general degradation started between 1995 and 1998 when the leakage channels and telltales becare blocked and extended to early 2003 when drain flow was re-established and degradation again became localized.

====2.2.3 Rebar====
Degradation from Migration through Joints/Cracks Once channels and telltales plugged and leakage accumulated in the gap between the liner and structure, the hydrostatic head forced the leakage into construction joints and cracks and ultimately into the sump room, the Auxiliary Building and the seismic gap (between the FHB and Auxiliary, Building).
Boric acid migration through the construction joints or cracks passed reinforcing steel, potentially initiating corrosion of the reinforcing steel. Boric acid migration through construction joints or cracks would react with the concrete prior to reaching the reinforcing steel. Hence, the pH of the leakage flow would likely be neutral or basic by the time it reaches the reinforcing steel. As noted above, studies of reinforcing steel corrosion from boric acid seepage through cracks showed negligible corrosion-only minor scarring-after two years of exposure (Reference 9.5.4). Further, measured corrosion rates of steel under static, de-aerated conditions with an acidic pH are low (< 4 microns per year) (Reference 9.5.5).The combination of evidence -studies in the literature, inspections of the FHB, testing conducted for Salem and experience at another US PWR -suggests that corrosion of embedded reinforcing steel from boric acid migration through cracks and construction joints is negligible.

===2.3 StructuralIAssessment===
Assessment of the structural adequacy of the degraded Salem Unit 1 FHB is based on a combination Qf the following.
* Projections of degradation incurred to date: Assessment of degradation drawing upon: evaluation of leakage from the SFP;chemical analyses of water draining through the telltales; and chemical analyses of the material blocking the telltales.
Degradation rates from the testing, with additional insights from other studies available in the open literature,-
and evaluations of similar issues at another PWR.Reviewiof the existing design margin and quantification of the impact of the projected degradation on available margin.Evaluation of the margin that can potentially be recovered through reassessment of the actual strength of materials used (i.e., reinforcing steel and concrete).
Table-2-1 shows the available margin after each of the three degradation modes is taken into account. Positive margin is maintained in all sections.MPR-2613 Revision 3 2-3 Table 2-1. FHB Available Margin Based on Predicted Degradation Modes Percent Available Margin Available FHB Wall Reduction In (WSD, No Location of Margin (WSD, Capacity Degradation)
Limiting Margin Including Degradation)
North 0.4% 4% Middle, Bottom 3.6%South 0.7% 300% West, 299%Toward Bottom East 0.7% 5% Middle, 4.3%Toward Bottom West 0.4% 2% Middle, Top 1.6%Slab 0% 3% Middle, Middle 3%It is concluded that the Salem Unit 1 FHB is currently structurally adequate and can withstand the design basis load combinations for up to seventy years, total plant life. Hence, the design basis analysis of record is not invalidated by the postulated degradation.

===2.4 CORROBORATION===

OF SALEM ASSESSMENT WITH CORES FROM CY's SFP Overall, the CY cores corroborate the results of testing for Salem and the projections for Salem.The maximum depth of concrete degradation in the CY cores is within that predicted using the correlation developed from the Salem testing. The rebar in the CY cores exhibited no corrosion even though the upper surface of the concrete was degraded by boric acid, the concrete was cracked and, based on the presence of secondary deposits within the concrete, there was water migration in the concrete.MPR-2613 Revision 3 2-4 w3 Review of Design Basis The design basis for the FHB structures is provided in Report MPR-1863 (Reference 9.2.3), which documents the structural design analysis of the FHBs as modified to include two 10,000 gallon; Service Water Storage Tanks in each unit. The most recent design analysis for the FHB before the addition of the Service Water Storage Tanks was performed during the FHB high density rack modification (Reference 9.3.1). In addition to the design analysis, the FHB was evaluated for beyond-design-basis thermal loading in EQE Calculation Number 200050-C-01 
.(Reference 9.3.2).This section provides a summary of the analysis performed in Reference 9.2.3.3.1 ANALYSISMETHOD Because the FHB and Service Water Storage Tanks are similar for Unit I and Unit 2, the design basis analysis in Reference 9.2.3 used bounding loads to perform a single analysis for both units.The analysis was performed by solving a three-dimensional finite element model, which* included the SFP, the transfer pool, and the surrounding walls. The model divided the FHB structure into approximately seventy sections.
Linearized shear and bending stresses were obtained for each section and were converted to equivalent shear loads and bending moments, respectively, for comparison with design allowables.

===3.2 DESIGN===
CONDITIONS This section defines the FHB design conditions used in the design basis analysis of Reference 9.2.3. The design conditions include the seismic category, material properties, design loads, load combinations, and design allowables.

====3.2.1 Seismic====
Category The FHB is a Seismic Category I Structure, per Section 3.1 of the Salem Structural Design Criteria (Reference.
9.1.2).3.2.2 Material Properties The following material properties were defined in Reference 9.2.3 for the FHB.* Concrete Compressive Strength:
fc 3,500 psi (Reference 9.4.1)* .Reinforcing Steel Yield Strength:
fy = 60,000 psi (Reference 9.4.1)MPR-2613 Revision 3 3-1 3.2.3 Load! Combinations The design load combinations for reinforced concrete structures at Salem are defined in Reference 9.1.2 and provided below. The acceptance criteria for the FHB structure, defined in Reference 9.1.2, Paragraph 7.2.1, are also listed below for each load combination.
Normal Operating Condition:
S=D+L + I+H+B+C Where: D = Dead Load L = Live Load I = Impact Load H = Normal Operating Thermal Load B = Buoyancy Load C = Commodity Load Stresses are limited to the working stress limits defined in Reference 9.1.3.Operating Basis Earthquake (North-South):
S = D + L + I + H + ENS + Ev + B + C Where: ENS = Operating Basis Earthquake in North-South Direction Ev = Operating Basis Earthquake in Vertical Direction Stresses are limited to the 1-1/3 of the working stress limits defined in Reference 9.1.3.Operating Basis Earthquake (East-West):
S = D + L + I + H + EEW +Ev + B + C Where: EEW = Operating Basis Earthquake in East-West Direction Stresses are limited to the 1-1/3 of the working stress limits defined in Reference 9.1.3.Design Basis Earthquake (North-South):
S = D + L + I + H' + E'Ns+E'v +B +P+J+C MPR-2613 Revision 3 3-2 Where: E'NS = Design Basis Earthquake in North-South Direction E'v = Design Basis Earthquake in Vertical Direction H' = M.ximum Thermal Load during Abnormal Conditions P = Internal Pressure Load J = Pipe Whip Load Loads are based on Ultimate Strength Design, limiting the stresses in the reinforcing steel to 90%of-the yield strength, per Reference 9.1.3.Design Basis Earthquake (East-West):
S=D+L+I!+H'+E'Ew+E'v+B+
P+J+C Where: E'EW = Design Basis Earthquake in East-West Direction Loads are based on Ultimate Strength Design, limiting the stresses in the reinforcing steel to 90%of the yield strength, per Reference 9.1.3.Tornado Loading: S D + L + I+ H',+ W + B + C Where: Wt= Tornado Loading Loads are based on Ultimate Strength Design, limiting the stresses in the reinforcing steel to 90%of the yield strength, per Reference 9.1.3..3.2.4 Design Allowables Working Stress Design Method The method for determining the allowable loads based on working stress design-are presented in Part IV-A of Reference 9.1.3. A summary of the analysis method is provided below.Allowakle Moment (M): The ass mptions for flexural design using the working stress design methods is provided in Referernce 9.1.3, Paragraph 1101. These assumptions are:* A plane section before bending remains plane after bending. Stresses vary linearly with the distance from the neutral axis.* The stress-strain relationship for concrete is a straight line under service loads and*within the allowable stresses.MPR-2613 Revision 3 3-3 0* The steel takes all of the tensile stress due to flexure.* The area of reinforcing steel is replaced by an equivalent area of concrete, scaled by the Modular Ratio (n).* In the tensile stress zone the concrete is assumed cracked and is not supporting tensile stress. Therefore, the compressive stress above the neutral axis is all carried by the concrete, and the tensile stress below the neutral axis is carried by the equivalent area of the steel. The stress distribution is shown in Figure 3-1.I-z6 kd fc kd13-C jd = d -3 0 d nMs-\V11177-1111
-T.....," ( I L Jb Figure 3-1. Stress Distribution for Working Stress Design The allowable moment for working stress design, which is based on yielding of the steel limiting failure rather than the concrete limiting failure, is: M = Asfjd Where: As = equivalent area of steel reinforcement f allowable stress of steel for working stress design 24,000 psi (Reference 9.1.3, Paragraph 1003(a))jd = distance between tension and compression forces d distance from extreme compression fiber to centroid of tension reinforcement MPR-2613 Revision 3&#xfd;3-4 Allowable Shear Load (0): The allowable average shear load across a section is: Vl = v bd where: b = width of concrete section ve= allowable concrete shear stress 1. l~fj' for working stress design (Reference 9.1.3, Paragraph 1201(c))Ultimate Strength Design The method for determining the allowable loads based on ultimate strength design is presented in Part IV-B of Reference 9.1.3. A summary of the analysis method is provided below.Ultimate Moment (Mj: The assumptions for flexural design using the ultimate strength design method are provided in Reference 9.1.3, Paragraph 1503. These assumptions are:* At ultimate strength the stress in the concrete is considered to be 0.85 f,' distributed over a rectangular area bounded by the edges of the cross section and extending a distance (a) into the depth of the cross section.The stress in the steel is assumed to be at yield (note that this is modified by Reference.
9.1.3 to a stress of 90% of yield).In the tensile stress zone the concrete is assumed cracked and is not supporting tensile stress. The assumed stress distribution is shown in Figure 3-2.MPR-2613 Revision 3 3-5 r 0.85f 'Cl C =O.85f' ab T =A~f 6 Figure 3-2. Stress Distribution for Ultimate Strength Design The ultimate moment is given as: Mu = 4A&#xfd;f,.(d _a ](Reference 9.1.3, Paragraph 1601 (a))where:= 0.9 for flexure (Reference 9.1.2, Paragraph 7.2.1)As = equivalent area of steel reinforcement fs, = 90% of yield strength of steel = 0.90fy (Reference 9.1.2, Paragraph 7.2.1)d = distance from extreme compression fiber to centroid of tension reinforcement a depth of compression zone (in.)(Reference 9.1.3, Paragraph 0.85f, b f= concrete compressive strength b = width of compression face a))MPR-2613 Revision 3 3-6 Ultimate Shear Load (Vj): The allowable average shear load across a section is: V=v bd where: v,= allowable shear stress for concrete= 2&#xfd;fc' for ultimate strength design (Reference 9.1.3, Paragraph 1701(c))3.3 CURRENT DESIGN MARGIN Twenty-four design margins (equal to the applied load over the allowable load) were calculated in Reference 9.2.3 for each section of the FHB; one for each of the six load combinations under each of the following four load types.* Horizontal Shear Load* Vertical, Shear Load* Horizontal Bending Moment* Vertical Bending Moment Results are provided in Table 3-1 by load type. The most limiting design margin from among all FHB sections ifor each of the six load combinations is provided.
Also, the applied and allowable loads are listed. Recall that the allowable loads for the normal operating condition and OBE are based on working stress design whereas the allowable loads for the other conditions are based on ultimate strength design. As shown, the normal operating condition and the OBE have the lowest margins for each load type. The low margins for the normal operating condition and OBE result from lower allowable loads associated with the working stress method.MPR-2613 Revision 3 3-7 Table 3-1. Loads and Margins for Each Load Combination Applied Allowable Load Limiting Location Load 3  Margin 2 Combination LaI La1 Horizontal Shear Load Anal, _sis Normal Operating South Wall -East, Middle -21.4 -42.4 1.99 East-West OBE South Wall -East, Middle -24.5 -56.6 2.31 North-South 
!OBE North Wall -West, Top -___31.0 -104. 3.36 East-West DBE South Wall -East, Middle -26.9 -65.5 2.44 North-South 
!DBE North Wall -West, Top -40.0 -120. 3.00 Tornado East Wall -North, Top _30.7 80.1 2.61 Vertical Shear Load Analysis Normal Operating East Wall -Middle, Bottom 15.0 51.8 3.45 East-West OBE West Wall -Middle, Bottom -25.5 -114. 4.47 North-SouthKOBE East Wall -Middle, Bottom 15.2 69.1 4.55 East-West DBE West Wall -Middle, Bottom -28.6 -132. 4.62 North-SouthIDBE East Wall -Middle, Bottom 14.7 80.1 5.44 Tornado East Wall -Middle, Bottom 19.9 80.1 4.02_Horizontal Bending Moment Analysis Normal Operating Slab- Middle, Middle -191.2 -197. 1.03 East-West OBE West Wall -Middle, Top -293.6 -299. 1.02 North-South OBE Slab -Middle, Middle -188.9 -263. 1.39 East-West DBE West Wall -Middle, Top -369.7 -469. 1.27 North-SouthIDBE Slab -Middle, Middle -256.1 -409. 1.60 Tornado I East Wall -Middle, Top -207.4 -285. 1.37_ Vertical Bending Moment Analysis.Normal Operating North Wall -Middle, Bottom -147.7 -154. 1.04 East-West OBE North Wall -Middle, Bottom -142.2 -206. 1.45 North-South OBE East Wall -Middle, Toward the Bottom -97.2 -137. 1.41 East-West DBE North Wall -Middle, Bottom -185.5 -321. 1.73 North-South'DBE West Wall -Middle, Toward the Bottom -203.8 -353. 1.73 Tornado West Wall -Middle, Toward the Bottom -216.3 -353. 1.63 Notes::, 1. Negative loads indicate compression on the pool side of the wall.2. Ma=rgn is defined as Allowable Load / Applied Load.3. Normal Operating and OBE allowables are based on working stress design methods, whereas the DBE and Tornado allowables are based on ultimate strength design methods.Table 3-2 identifies all sections having design margins less than 10% from among all load combinations and load types. Also included in the table are the applied loads and allowable loads used to calculate the design margins, the locations of the limiting sections, and the load combinations and load types that produced each applied load and allowable load. All of the cases with less than 10% margin are for normal operation and OBE. Once again, the low margins result from low allowable loads associated with the working stress method.MPR-2613 Revision 3 3-8 Table 3-2. Limiting FHB Design Margins from the Design Basis Analysis Load Load Applied Allowable Design Combination Type Limiting Section Location Load' Load 1  Margin 2 (kip-ftlft) (klp-ftlft)
West Wall -Middle, Towards Bottom -215 -225 1.04 West Wall -Middle, Towards Top -216 -225 1.04 Horizontal West Wall -Middle, Top -208 -225 1.08 Normal Moment East Wall -Middle, Towards Top -125 -136 1.09 Operation Slab -Middle, West -184 -197 1.07 Slab- Middle, Middle -191 -197 1.03 Vertical North Wall -Middle, Bottom -148 -154 1.04 Moment East Wall -Middle, Towards Bottom 103 1.05 East-West Horizontal West Wall -Middle, Towards Top -279 -299 1.07 West Wall -Middle, Top -294 -299 1.02 OBE Moment West Wall -South, Top -274 -299 1.09 Notes: 1. Negative loads indicate compression on the pool side of the wall.2. Margin is defined as Allowable Load / Applied Load.Review of Tables 3-1 and 3-2 indicates that most of the limiting margin cases, including all of the cases with less than 10% margin, have negative loads which denotes compression on the pool side of the wall (or slab) and tension on the outside of the wall (or slab). Since reinforcing steel carries the tensile loads, the reinforcing steel of primary concern with regard to structural margin is the rebar near the outside of the wall -the side farthest from the pool and farthest from the spent fuel pool water which may reside in the gap between the liner and the wall. Because concrete carries compressive loads, the concrete of primary concern with respect to structural margin is that beside the liner gap.MPR-2613 Revision 3 3-9 4 Potential Margin Recovery As discussed in Section 3, the current design basis analysis of the FHB shows that little design margin (less than 10%) exists in several areas of the FHB, allowing for very little degradation of the FHB concrete and reinforcing steel. This section evaluates the margin that can potentially be recovered through- assessment of measured material properties.

===4.1 REINFORCING===

STEEL CAPACITY Per Reference 9.4. 1, the reinforcing steel in the FHB structure has a specified minimum yield strength of 60Iksi. However, the actual yield strength of reinforcing steel is typically higher than the specified minimum value. Using the actual yield strengths of the reinforcing steel is a potential method to recover margin in the FHB.During construction of the Salem units, tensile testing of reinforcing steel was performed to verify that the yield and ultimate strengths met or exceeded the minimum specified value. MPR Calculation 108-275-02 (Reference 9.3.4, provided in Appendix B) documents a statistical analysis on a sample of this yield strength test data. The total sample population was comprised of sub-sample~s of each reinforcing steel size present in the FHB structure.
The statistical analysis determined the mean yield strengths for each sub-sample and the total sample population.
The analysis also characterized the distribution of yield strengths in terms of the percentage of each sub-sample and the total sample population greater than a given yield.Results from the statistical analysis are provided in Table 4-1. The yield strength distribution for the total sample population is shown graphically in Figure 4-1.Table 4-1. FHB Reinforcing Steel Yield Strength Analysis Results Rebar Mear Yield Yield Strength Sample 80% Lower 85% Lower 90% Lower 95% Lower Type Stre ngth Std. Dev Size Bound' Bound' Bound' Bound 1 Total 69,840 6,370 394 64,100 63,300 62,200 61,300 No. 6 67,692 5,027 13 64,100 63,850 62,550 62,550 No. 8 69,'410 7,376 123 63,500 63,000 61,500 60,750 No. 9 70,.815 T 6,591 95 63,800 63,000 62,500 61,600 No. 10 70,934 5,024 47: 66,150 65,400 62,200 62,100 No. 11 69,363 5,490 116 64,950 63,300 62,065 60,900 Notes: 1. The indicated percentages of the sample sizes have yield strengths greater than those shown.2. All yield strengths are in psi.MPR-2613 Revision 3 4-1 60 95%90% 85% 80%50 40 30---0 U.20-10-Yield Strength (psi)Figure 4-1. FHB Reinforcing Steel Yield Strength Distribution of Population Sampled The above table and figure show that the actual yield strengths of the reinforcing steel in the Salem FHB arie larger than the minimum specified 60 ksi. Based on the total population evaluated, the mean yield strength is almost 70 ksi and the 95% lower bound is 61.3 ksi (i.e., 95% of the data are greater than 61.3 ksi).In the working stress design method, the allowable reinforcing steel stress is equal to 40% of the nominal yield strength of the material.
For reinforcing steel with a yield strength of 60 ksi, the allowable stre~s is then 24 ksi. To assess the margin recovered using measured yield strength, the 95% iowe" bound value of 61.3 ksi is used to conservatively bound a significant portion of the steel. Using 40% of 61.3 ksi (24. 52 ksi) as the yield strength in the methodology described in Appendix D increases the design margin by about 2%.4.2 CONCRETE CAPACITY The testing documented in Reference 

====9.2.4 included====
compressive strength tests for concrete specimens prepared using the same mix design and same raw material suppliers as the concrete used in the F1B. The tests showed that the concrete mixture used at Salem has a compressive strength of ab ut 6,000 psi, compared to a specified design value of 3,500 psi. The impact of concrete strength on the concrete capacity can be assessed by review of the calculation in Appendix D. ihe moment capacity of the concrete is not sensitive to the actual concrete strength.
Specifically, the increase in concrete strength from 3,500 psi to 6,000 psi provides a very small (<< 1%) increase in the available margin. Accordingly, the potential for margin recovery from measured concrete strength is not-considered further.0 MPR-2613 Revision 3 4-2 

==4.3 CONCLUSION==
S Design margins in the Salem FHB under normal operating conditions and OBE conditions may be slightly improved through the use of measured material properties as opposed to specified or nominal properties.
In the working stress design method, the allowable reinforcing steel stress is equal to 40% of the nominal yield strength of the material.
Using 40% of the actual yield strength of the reinforcing steel in the working stress design calculations in lieu of the specified normal allowable recovers 2% margin. Using the actual compressive strength of the concrete recovers a negligible amount of margin.The limiting margin in the FHB can be increased from 1.02 to at least 1.04 by taking credit for the actual yield strength of the reinforcing steel. Subsequent sections will show that degradation expected from boric acid attack is less than that required to challenge the structural capacity of the FHB.MPR-2613 Revision 3 4-3 5 Boric Acid Attack of Concrete and Reinforcing Steel Several activities were performed to assess the impact of boric acid on concrete and reinforcing steel. First, MPR performed a review of industry literature regarding the effects of boric acid and other acids on concrete and reinforcing steel. The results of the review were previously provided to PSEG Nuclear via Reference 9.6.3. Second, MPR conducted testing to determine how concrete and reinforcing steel are affected by exposure to boric acid. -The testing was research-oriented in nature with the goals of understanding the mechanisms of concrete and reinforcing stlel degradation and quantifying degradation rates. Details of the testing are documented in MPR-2634 (Reference 9.2.4). Third, MPR reviewed evaluations of concrete and embedded reblar degradation from SFP leakage at another US PWR. Insights identified from the literature review and testing are provided in the following sections.5.1 CHEMICAL REACTIONS BETWEEN ACID AND CONCRETE References 
: 9. .2 and 9.5.3 provide excellent discussions of the mechanisms of concrete degradation finom acid attack. Cement paste in concrete is easily attacked by acidic solutions due to its high alkhlinity.
As the acid attacks the concrete, the cement constituents are altered by decalcification, leading to degradation of the concrete properties.
Portlandite (Ca(OH)2) is the first cement c nstituent to react with the acid. Calcium silicate hydrate also reacts with the acid.In most cases of acidic attack, the chemical reactions result in the formation of calcium salts.The corrosive!
effect of an acid depends on the solubility of these salts; a higher solubility contributes to the progression of attack.The reaction l&#xfd;etween hardened cement paste or concrete and an acid solution is controlled by diffusion of t!he acid into the concrete.
The rate at which the concrete (or paste) degrades decreases over time as the distance acid must diffuse through degraded concrete to reach intact concrete increases.
Degradation of concrete by acids follows a Fick's Law of Diffusion formulation iA which the depth of degradation varies with the square root of time. Hence, the rate of degradation decreases monotonically, approaching zero asymptotically.
The degradation rate depends on the acid. The typical signs of acidic attack include a gradual loss of alkalinrity, loss of mass, and loss of strength and rigidity.Reference 

====9.5.3 provides====
additional insight regarding the degradation of concrete due to acid attack. Althoi!gh concrete degradation is typically higher when soluble salts, as opposed to insoluble or nearly insoluble salts,- are formed during the reaction process, formation of insoluble or nearly insoluble salts can create microcracks during crystallization, which can lead to spalling of the concrete.MPR-2613 Revision 3 5-1 If a cement matrix is continuously immersed in an acidic solution rather than exposed to alternate wetting and drying cycles, the expansion caused by salt crystallization is less, and may not occur at a01. The paper also states that in the first few days or weeks of exposure to acid, cement basedimaterial can become denser with corresponding increases in weight and compressive Otrength.
These phenomena, attributed to small amounts salt crystallization and deposition of corrosion products in the relatively open pore structure of the cement based material, are reported to be temporary until the salt crystallization is high enough to show deteriorating effects (spalling, cracking, etc.).5.2 LITERATURE STUDIES ON BORIC ACID ATTACK 5.2.1 Degradation of Concrete and Cement Paste Reference 9.51.2 investigated the effect of various acids (boric acid was not included) on concrete.
Results of testing showed that formation and growth of a layer of reaction products is typical for concrete exposed to acids. The degraded layer is usually soft, cracked, and without bonding properties.
In the drying process, the degraded layer shrinks, cracks widen, and the layer can be crushed easily. When a degraded layer is formed, the mechanical properties of a specimen depend primarily on the quality of the 'non-degraded core' of the cement paste.The testing documented in Reference 9.5.2 also demonstrated that attack of Portland cement concrete by weak acids, such as boric acid, usually results in low depths of penetration, and is diffusion-controlled.
Curve fits of the test data show that the depth of degradation versus time follows a Fick's Law of Diffusion formulation-depth increases with the square root of time.Further, this reference states that corrosion rates are dependent upon the pH value of the solution.Reference 

====9.5.1 documents====

the results of testing on hardened Portland cement paste specimens that were cured in boric acid solutions, and on concrete exposed to boric acid in the field. The cement paste &#xfd;pecimens were exposed to boric acid solutions for up to 127 days. The results of the testing showed that the weight, bulk density, and compressive strength of the specimens increased due, to boric acid exposure, and the porosity of the specimens decreased.
No specimen degradation was reported.
The paper attributed these results to the formation of low-soluble hydrated calcium borates from the reaction between the boric acid solution and Portlandite, which filled up the pore system of the cement paste. The paper noted that the test results were different frorr typical acid attack, which usually results in a loss of weight, decrease in density and compress ve strength, and increase in porosity.Applying the discussion of Reference 9.5.3 to the results of the cement paste specimen testing documented in Reference 9.5.1, the cement paste specimens may have increased in density, weight, and compressive strength and showed a decrease in porosity because the reaction rate did not have the opportunity to increase to the point where salt crystallization could have deteriorating effects on the specimens.
The slow increase in pH after the first week of testing shows that th4 reaction rate was low. As previously discussed, Reference 

====9.5.3 indicates====

that increases in dnsity and compressive strength after exposure to acid is a temporary phenomena that can occur during the first days or weeks of exposure.MPR-2613 Revision 3 5-2 

====5.2.2 Compressive====

Strength of Concrete Reference 

====9.5.1 documents====

the results of compressive strength tests performed on concrete exposed to boric acid in the field. The results showed that boric acid had no effect on the compressive strength of the concrete, or any other properties that were tested. No degradation of the concrete was reported.
While the boric acid appeared to have no effect on the concrete, factors that affect attack, such as the pH of the solution, whether or not the solution was refreshed, and length of time the concrete was exposed to the solution, were not provided in the reference.
Therefore a strong conclusion related to the effect of boric acid on concrete can not be made with respect to this test. The paper did report that the concrete aggregate was limestone.
Because the limestone aggregate can react with the acid, the findings may not be applicable to concrete mixes using different aggregates.

====5.2.3 Corrosion====

of Rebar Reference 

====9.5.4 reports====
on testing performed to study the effects of reinforcing steel corrosion due to boric acid entering reinforced concrete through cracks. The tests showed that corrosion increases as crack width increases and pH decreases.
In particular, the tests showed negligible reinforcing steel attack even when specimens were subjected to the most corrosive test environment (pH of 5.2) with the largest crack width (0.4 mm) for a period of two years.Corrosion was limited to scarring in the area of the crack.5.3 MPR (CRT) TESTING OF BORIC ACID ATTACK TO SUPPORT SALEM FHB EVALUATION Reference 

====9.2.4 documents====

testing conducted by MPR to support the FHB structural evaluation.
Specifically, the testing determined how concrete and reinforcing steel are affected by exposure to boric acid. , The testing used concrete cores from the Salem Auxiliary Building and additional specimens prepared using the same concrete mix and suppliers as for the Salem concrete.The specimens were soaked in a boric acid bath with a boron concentration consistent with the SFP. The bath was periodically refreshed to maintain acidic conditions.
The testing was research-oriented in nature with the goals of understanding the mechanisms of concrete and reinforcing steel degradation and quantifying degradation rates, rather than to closely replicate the conditions behind the SFP liner. The testing was performed under MPR's lOCFR50 Appendix B Quality Assurance Program.5.3.1 Boric.Acid Attack on Concrete The testing included exposure of concrete specimens to a boric acid solution for up to 9 months.The testing used a combination of cores from the Auxiliary Building at Salem and cylinder-rebar specimens prepared using the same concrete mix and suppliers as for the concrete used at Salem.Microscopic examinations and chemical analyses were performed on the specimens after exposure to the solution.
The results showed that boric acid reacted with the alkaline constituents of the cement paste of concrete.
This reaction caused cracking and a loss of bonding MPR-2613 Revision 3 5-3 from the fine and course aggregate, and left the reacted paste soft and porous with no strength;fine aggregate particles were easily dislodged.
The rate of degradation decreased with the square root of time, as is expected for diffusion-controlled processes.
Based on Reference 9.5.2, the projected depth of degradation after 70 years 1 of exposure is 1.30 inches, including adjustments for temperature and uncertainty.
Compressive strength testing was performed to assess the impact of boric acid degradation on concrete strength after 56 days of exposure.
The apparent compressive strength for specimens soaked in boric acid was lower than that for control specimens soaked in tap water. However, the difference:
in compressive strengths can be explained by accounting for the reduction in cross-sectional area from boric acid attack.5.3.2 Boric Acid Attack on Reinforcing Steel Boric acid attack of reinforcing steel was investigated by MPR (CRT) using concrete specimens with embedded rebar. The specimens were soaked in boric acid for up to 56 days. The cylinder-rebar specimens provided insights on rebar corrosion beneath the concrete surface and the wicking rate of boric acid along the rebar.Only one of the specimens exhibited any reinforcing steel corrosion below the concrete surface. This specimen showed very minor surface corrosion just beneath the concrete surface. This specimen had a surface discontinuity at the rebar-to-concrete interface, which allowed the boric acid solution to contact the rebar below the nominal concrete surface. Hence, the observed corrosion is not indicative of the corrosion of embedded rebar.The wicking rate along the concrete/rebar interface was minor. That is, the degradation of concrete at the concrete/rebar interface is similar to the general rate of attack of concrete without rebar. Therefore, any degradation of reinforcing steel will remain localized to the region where boric acid contacts the rebar.5.4 EVALUATIONS FROM ANOTHER PWR Other PWRs Iave also experienced SFP leakage and evaluated the impact of boric acid on the concrete structure surrounding the SFP. Reference 

====9.2.5 documents====

the evaluation atone of these plants. The plant in question experienced leakage from the SFP which migrated through a crack in the concrete to an adjacent space underneath the SFP. The leakage occurred over a period of several years. Concrete was chipped away to expose rebar in the vicinity of the crack.The crack ran parallel to the rebar, directly next to the rebar. Inspections of the exposed rebar revealed no discernable corrosion of the rebar. This observation is consistent with the negligible observed corrosion of rebar exposed to boric acid via concrete cracks determined in Section 5.2.3, above.Degradation was projected over 70 years to provide a bounding projection that envelopes potential license renewal and storage of fuel in the SFP for 10 years after cessation of operations.
Use of 70 years in the evaluation should not be interpreted as a commitment by PSEG Nuclear to pursue license renewal of Salem.MPR-e613 Revision 3 5-4 

==5.5 CONCLUSION==
S Key insights from the literature review, and the testing, as well as the experience at other PWRs are as follows.Formation and growth of a layer of reaction products is typical for concrete exposed to acids, including boric acid. The degraded layer is usually soft, cracked, and without bonding properties.
In the drying process, the corrosion layer shrinks, cracks widen, and the layer can be crushed easily. When a degraded layer is formed, the mechanical properties of a specimen depend mainly on the quality of the 'non-degraded core' of the cement paste." The attack of Portland cement concrete by weak acids, such as boric acid, usually results in low depths of penetration, and is controlled by diffusion of the acid into the concrete.Curve fits of the test data show that the depth of degradation versus time follows a Fick's Law of Diffusion formulation-depth increases with the square root of time.* The extent of reinforcing bar corrosion in reinforced concrete depends primarily on the crack width and the pH value of the solution.
An increasing corrosion rate is observed for larger crack widths and lower pH values." Significant reinforcing steel corrosion is not expected when boric acid is introduced to the steel through a crack in the concrete because the rebar is protected by the alkaline concrete matrix. Laboratory studies and experience at another PWR show negligible corrosion of embedded rebar after 2 years of exposure to boric acid.* The wicking effect at the interface of concrete and rebar is minor. Therefore, degradation of rebar at the construction joints or cracks with migration will not spread rapidly along the rebar; i.e., rebar degradation is localized to the vicinity of the construction joint or crack.MPR-2613 Revision 3 5-5 6 Assessment of Potential Damage to Structure This section assesses the nature of potential degradation of the FHB structure from exposure to the boric acid solution leaking from the SFP. The assessment draws upon:* Evaluation of leakage from the SFP,* Chemical analyses of water draining through the telltales,* Chemical analyses of the material blocking the telltales, and* Independent structural assessment per ACI guidelines by an experienced concrete structural engineer.Based on these evaluations, degradation of the structure is estimated for portions of the concrete that have potentially been exposed to boric acid from the SFP.O 6.1 LEAKAGE EVALUATION It is generally considered that the source of liner leakage is cracking of the liner seam welds and/or the plug welds attaching the liner to the studs embedded in the concrete.
Since the backing bar for the seam welds is tied to the embedded leakage channels and the plug welds are tied to embedded studs, these welds can be highly stressed due to differential thermal expansion between the liner and the concrete structure.
Given the large number of seam welds (about 2,100 feet) and plug welds (about 1,400), it is likely that there are multiple leaking cracks as opposed to a single large crack. The plug welds are considered more likely to crack and leak on the basis that the differential thermal expansion loads are more concentrated resulting in high stresses.6.1.1 Crack Size/Leakage Rate Scoping calculations were performed to estimate the crack size necessary to produce the nominal leakage rate of 100 gpd. The required crack length varies with the hydrostatic head across the crack (i.e., elevation of crack and water level in gap) and the crack width. If the crack is on the bottom of the, pool and there is no water in the liner/wall gap, the crack length necessary to produce 100 gpd ranges from about 0.5 inch for a 0.003 inch wide crack to about 6 inches for a 0.00 1 inch wide crack.0 MPR-2613 Revision 3 6-1 W The scoping calculations suggest that the crack or cracks causing the leakage are very small, particularly in comparison to the total length of seam welds and number of plug welds. Small, tight cracks are difficult to locate. It is unlikely that video inspections with underwater cameras or vacuum box testing would be able to successfully locate such cracks.2 6.1.2 Leakage from Seam Welds versus Leakage from Plug Welds As discussed above, cracks could occur in the liner seam welds and/or the plug welds to embedded studs. The flow path for each leak location is described below.* Seam Weld Leakage. Leakage through seam welds collects in the leakage channel embedded in the concrete and flows out the telltale to a trough in the Sump Room.Provided the leakage channels and telltales are not obstructed, the boric acid solution from the SFP does not contact the concrete of the FHB structure." Plug Weld Leakage. Leakage associated with a plug weld exposes concrete to the boric acid solution from the SFP. Leakage from a weld on the pool bottom drips onto the concrete slab, forming a puddle, which grows until it, overflows into a leakage channel and is routed to a telltale.
For leakage through a plug weld in a wall, thc boric acid solution runs down to the slab forming a puddle which grows until it overflows into a leakage channel. Exposure of the concrete to the boric acid solution is limited to the flow path from the leak to an open channel.When the leakage channels and telltales are obstructed, leakage from the SEP accumulates in the gap between the FHB structure and the SEP liner, exposing much larger areas of the structure to the boric acid: solution and potential degradation.

===6.2 TELLTALE===
CHEMISTRY PSEG Nuclear's Chemistry Department has analyzed samples of the liquid discharge from the telltales.
The samples have been subjec ted to both chemical analysis and isotopic analysis.
The analyses are documented in Reference 9.2.2; key results and insights are provided below.* Isotopic analysis of liquid samples collected in December 2002 just prior to snaking of the telltales shows that the isotop ic signature is consistent with the SFP chemistry with about five years of decay.* The average pH of SEP telltale samples collected after cleaning the telltales was 7. 1, compared to an expected pH of 4.6. The transfer pool telltales showed a similar trend: measured pH of 7.*8 compared to an expected pH of 4.8. The high pH values indicate that the boric acid solution has reacted with alkaline constituents of the concrete.2In 1995, the Salem Unit 1 SFP was inspected for weld leaks using vacuum box testing. Almost 95% of the seam welds were inspected using vacuum box testing with no indications of a crack; the remainder of the welds could not 0 be inspected due to limited access under the fuel racks (Reference 9.3.6).Revision 3 6-2 O The above results indicate that the liquid accumulated in the gap was about 5 years old and that the boric acid- had reacted with the concrete structure.
Since basic solutions are relatively benign to carbon steel, the high pH values suggest that active degradation of any carbon steel exposed to the boric solution had largely slowed over the time period the liquid was trapped behind the liner.Analysis of telltale samples taken on 11/20/03 show pH values ranging from 6.1 to 7.2, with an average of 6.8 (Reference 9.6.4). As discussed above, pH values are higher than the typical SFP pH of 4.6 indicate that the boric acid is reacting with the concrete.
Hence, even though the stored inventory of has been eliminated from behind the liner, boric acid is coming into contact with concrete.
This indicates that concrete degradation is continuing.

===6.3 CHEMICAL===
ANALYSIS OF MATERIAL BLOCKING TELLTALES/CHANNELS PSEG Nuclear obtained samples of the solid material that was obstructing the channels/telltales and contracted with Framatome-ANP for analysis of the samples. The analyses indicate that the deposits are largely quartz (Si0 2) and calcite (CaC03) with minor amounts of gismondine (CaAl 2 Si 2 O 8.4H 2 0) (Reference 9.6.5). In other words, the material obstructing the telltales and leakage channels derives from the concrete of the FHB.The mechanism for formation of the blockages in the telltales/leakage channels is not well understood.
The calcite likely precipitates out of solution as dissolved calcium compounds are carbonized by reaction with carbon dioxide from the air in the leakage channels and telltales.
The dissolved calcium compounds derive from the concrete via one of the following processes.
* Attack of the concrete by boric acid leaking from the SFP plug welds results in the generation of dissolved calcium compounds such as calcium hydroxide and calcium boratesj GroUnlwater in-leakage through cracks or constructions can leach calcium species from the concrete as it migrates through and over the concrete on its way to the leakage channels and tellfales.
Precipitation of calcite -and other species can also be impacted by evaporation of water and solubility changes as SFP leakage cools to the concrete temperature.
In short, the process for forming the blockages is similar to the formation of stalactites and stalagmites in a cave.The fact that the obstructions include materials derived from concrete indicates that concrete degradation was occurring prior to-the drains becoming plugged. This is clear evidence that the liner plug welds are leaking (see Section 6.1.2), because seam weld leakage would be diverted immediately to the drain channels without contact with concrete.6.4 LEAKAGE TIMELINE Drainage frori the telltales has been noted since 1980. Key milestones regarding SFP leakage*are identified below and presented as a timeline in Figure 6-1.MPR-2613 Revision 3 6-3 1981. A modification to address seam weld leakage was implemented in 1981. This modification consisted of positioning seam encasements over leaking seam welds, and welding the encasements to the liner (Reference 9.6.1).1995. A project to install high-density fuel racks to increase SFP storage capacity was implemented in 1995. Significant leakage through the telltales was noted during the project (Reference 9.3.6), particularly during rack moves. This was likely the result of water being pushed out from between the liner and concrete slab during the changes in floor loading. Therefore, it appears that portions of the liner to concrete gap in the slab were flooded with boric acid from the pool.0 1998. Isotopic analysis of telltale samples from when the drains were cleaned in early 2003 suggest that the liquid accumulated in the gap between the liner and concrete was about 5 years old.* 2002. In the Fall of 2002, PSEG Nuclear identified leakage from the FHB into the Auxiliary Building and into the seismic gap. In addition, leakage into the sump room via a construction joint was noted. This indicates that the liner to concrete gap was flooded up to an elevation above the fuel pool slab.* 2003. In January 2003, PSEG Nuclear cleaned the drains to re-establish flow from the channels and telltales.
The drain flow increased significantly following cleaning.
Later in the year, PSEG Nuclear performed hydrolazing to remove more blockages.
It appears that leakage from the plug welds likely initiated sometime before the 1995 rerack project when water had already puddled or accumulated in the liner floor-to-slab gap. However, at that time at: least some of the channels and telltales were not obstructed, so complete flooding of the gap had not occurred.Blockage of the drains and accumulation of water in the gap between the pool and the walls occurred sometime between 1995 and 1998. In 1995 some of the channels and drains were open for flow. However, radioisotopic analysis indicates that by 1998 SFP leakage was accumulating in the gap rather than flowing out the telltales.
The 1995 re-rack project may have contributed to formation of telltale blockages.
In general, the velocity of the leakage overflowing into the channels is low, too low to entrain sand and other particulate matter. However, leakage flow observed during the re-rack as water was forced into the channels likely was high enough to carry particulates into the channels.0 MPR-2613 Revision 3 6-4 1981 1995 1998 Sep 2002 Jan 2003 Seam Weld. Rerack -Origin of Accumulated Leakage Into Aux. Bldg, -Initial Snaking to Encasement (Portion(s) of Water based on Sump Room, Seismic Re-establish Drain Jun 1977 Repair Floor Wet) Isotope Analysis Gap Identifed Flow Commercial Operation A General Degradation from Water Accumulated in Gap between Liner and Concrete Local Degradation Near Plug Welds Figure 6-1. Leakage Timellne for Salem Unit I 6.5 INDEPENDENT STRUCTURAL ASSESSMENT PER ACI GUIDELINES PSEG Nuclear had an experienced concrete structural engineer perform an independent structural assessment of the FHB in 2006. The assessment included review of building drawings, a visual inspection of the accessible portions of the FHB exterior walls, and a visual inspection in the Sump Room. The checklist in ACI 201.1R-92 was used to guide the inspections.
Observations were compared to limits in ACI 349.3R. The assessment is documented in Reference 9.2.6.Key conclusions from the independent assessment are excerpted below."Overall the concrete appears to be in good structural condition."* The "appearance of leaching or chemical attack and corrosion staining of-undefined source on concrete.
surfaces do not indicate significant structural deterioration at this time."* There were "no indications of concrete surface expansion due to reinforcing steel corrosion." It also recomnends periodic inspections to trend the condition of the building.6.6 DEGRADATION OF FHB STRUCTURE FROM BORIC ACID EXPOSURE The above evaluations and the leakage timeline provide important insights into potential degradation of the FHB structure from exposure to the boric acid solution leaking from the SFP.As discussed above, seam weld leakage in and of itself is not a concern as the leakage is collected in the channels and discharged via the telltales.
Plug weld leakage results in local degradation of the concrete structure.
Once the channels and telltales become plugged, weld leakage (plugi weld and seam weld leakage) accumulates in the gap between the liner and the structure and jresults in general wetting and degradation of the concrete structure.
MPR-2613 Revision 3 6-5 Hydrostatic pressure of the water in the gap ultimately forced water through construction joints or cracks to thle Auxiliary Building, the Sump Room and the seismic gap (between the FHB and the Auxiliary Building).
Water migration through the construction joints or cracks flowed past reinforcing steel, potentially corroding the steel.Once drain flow from the telltales was re-established, the accumulated leakage drained out and the hydrostatic pressure to force the boric acid solution into construction joints or cracks was eliminated.
Consequently, general degradation from the stored inventory ceased and potential degradation of reinforcing steel from migration through joints or cracks ceased. However, active leakage from the pool continues to degrade concrete as it migrates to an open telltale.
Since some telltales are partially or fully blocked, seam weld leakage could be contributing to concrete degradation.
Detailed discussions of the expected structural degradation are provided below. The discussion is divided by degradation mechanism:
local degradation from weld leakage, general degradation from water accumulated in the gap, and reinforcing steel corrosion in construction joints or cracks.6.6.1 Local Degradation from Weld Leakage Local degradation of concrete from weld leakage occurs in two different ways." Boric acid leakage from a plug welds contacts the concrete in the vicinity of the leaking welds and degrades the concrete.* If a telltale is blocked, weld leakage that normally collects in the channel and flows out the telltale, overflows the channel and migrates across concrete to a channel with an open telltale.
The boric acid degrades the concrete along this migration path.Each mechanism is described below and then the extent of degradation over remaining plant life is projected.
Plug Weld Lgakage Leakage associated with liner plug welds results in local degradation of the concrete structure, primarily the slab underneath the pool. Leakage from plug welds, whether on the wall or the bottom of the liner, puddles on the slab until it overflows into the channels.
For a leaking plug weld in the bottom of the pool, the puddle will be located in the vicinity of the leaking weld.For a leaking plug weld in the pool wall, the puddle will be in the comers where the wall and slab intersect.'
Concrete under the puddles will degrade from exposure to the boric acid solution.
The boric acid will attack the cement paste, weakening it and causing it to de-bond from the coarse and fine aggregate.
As the degradation progresses, a rubble bed of coarse and fine aggregate may be formed on top of the concrete as the cement in the top layer fully degrades.
In essence, the local degradation will create a "pothole" with sand and coarse aggregate on top of the remaining concrete.MPR-2613 Revision 3 6-6 O In the case of plug weld leaks in the FHB walls, there may be some local degradation of the wall as the leakage! flows down the liner or the concrete wall. Such damage is expected to be limited to the immediate vicinity of the flow path of the leakage flow down the wall. Weakening of the cement paste and de-bonding of the aggregate results in debris falling down the wall to the slab.Falling debris increases the gap between the liner and the concrete in the upper portion of the wall reducing ;the potential for boric acid leakage to contact the concrete.
Accumulation of debris in the gap in the lower portion of the wall increases the distance that boric acid must diffuse to come into contact with concrete, thereby reducing the degradation rate.As previously discussed, leakage associated with plug welds started prior to the 1995 re-rack.However, pinpointing a date is very difficult.
Degradation associated with this leakage will continue into the future even though drain flow has been re-established.
Migration of Leakage to Open Telltale Once a given telltale is blocked, any weld leakage-seam weld leakage or plug weld leakage-that normally collects in the channel cannot flow out the telltale.
This leakage overflows the channel and migrates across concrete to a channel with an open telltale.
The concrete along the migration path to an open telltale is subject to degradation similar to that described above for plug weld leakage. This degradation mode primarily impacts the slab as leakage-from plug welds on the liner walls or the liner floor-collects on the slab and overflows into an open telltale.Pinpointing the date that this mode of degradation started is difficult.
However, it is clear that it started sometime prior to 1995 as some of the telltales were plugged at the time the re-racking was performed in 1995. This mode of degradation will continue into the future as not all telltales have been fully cleaned at this time and some may re-block between periodic telltale cleanings.
Projected Degradation To assess the structural implications of this degradation mode, it is assumed that local areas of the concrete structure have been subjected to degradation for a period of 70 years. The 70-year time period is a conservative value that spans the entire plant life including:
license renewal 3;and maintaining fuel in the SFP for 10 years after cessation of operations.
Per Reference 9.2.4, the projected depth of local concrete degradation after 70 years is 1.30 inches. Since the concrete cover for all walls and slab is greater than the projected depth of concrete degradation, no reinforcing steel degradation is expected for this mode of degradation.
The projectedi depth of local degradation is applied to the entire slab. While plug weld leakage results in degradation of only a local area, blockage of the telltales expands the area of the slab subjected to long term exposure to boric acid as the boric acid migrates to an open telltale.Applying the maximum depth of degradation to the entire slab is conservative.
3 Consideration of license renewal in determining the plant operating life is not an indication that PSEG Nuclear has committed to pursuing license renewal. Instead, it is included to provide a bounding assessment.
MPR-2613 Revision 3 6-7 

====6.6.2 General====
Degradation from Water behind Liner As the leakage channels and telltales became plugged, leakage from the SFP accumulated in the gap between the liner and the concrete on the floor and the walls. The water level in the gap likely increased until it equalized with the level in the pool, at which point leakage essentially stopped and conditions in the gap became stagnant.General degradation of the concrete is similar to that described above for plug weld leakage, except the degradation is widespread rather than localized.
Virtually the entire structure surrounding the pool is exposed to boric acid and subject to degradation.
The period of general degradation starts sometime between 1995 and 1998 when the leakage channels and telltales became blocked, and extends to early 2003 when drain flow was re-established.
This mode of degradation is not expected to recur as PSEG Nuclear has implemented multiple measures to ensure that the telltales do not become entirely blocked (trending of telltale leakage rates, periodic videoprobe inspections and cleanings).
Reference 

====9.2.4 contains====
a calculation which projects degradation over a 70-year span. Using the projected degradation curve therein and the temperature adjustment, the projected general degradation over an 8-year interval is about 0.44 inch. Since the concrete cover for all walls and the slab is markedly greater than the projected depth of concrete degradation, no reinforcing steel degradation is expected for this mode of degradation.
The projected depth for general degradation is more appropriate to use in structural assessments of the walls than the local degradation projection.
Although local areas of the walls near leaking plug welds could be degraded to deeper depths, there is no mechanism for expanding these local areas to a significant area. Further, structural margin is driven by the condition of the general area, not small localized areas.6.6.3 Degradation from Migration through Construction Joints and Cracks Once channels and telltales plugged and leakage accumulated in the gap between the- liner and structure, the hydrostatic head forced the leakage into construction joints and cracks and ultimately into the Sump Room, the Auxiliary Building and the seismic gap. Migration through the construction joints or cracks passed reinforcing steel, potentially initiating corrosion of the reinforcing steel.The combination of evidence-studies in the literature, inspections of the FHB, testing conducted fot, Salem and experience at another US PWR-indicates that reinforcing steel degradation in the FHB is minimal and structural capacity has not been impacted.
The key points that support this case are as follows.The reinforcing steel of concern from a structural standpoint is the reinforcing steel near the outside of the wall (or slab)-the side farthest from the pool and farthest from the SFP water which may reside in the gap between the liner and the wall.MPR-2613 Revision 3 6-8
* Review of Tables 3-1 and 3-2 indicates that most of the limiting margin cases, including all of the cases with less than 10% margin, have negative loads which denotes compression on the pool side of the wall (or slab) and tension on the outside of the wall (or slab). Since reinforcing steel carries the tensile loads, the reinforcing steel of primary concern with regard to structural margin is the rebar near the outside of the wall. Hence, boric acid must migrate through the walls, which are several feet thick, to reach the reinforcing steel of concern.Boric acid migration through construction joints or cracks would react with the concrete prior to reaching the reinforcing steel. Hence, the pH of the boric acid would likely be neutral or basic by the time it reaches the reinforcing steel. Further, migration through the construction joint or crack would become de-aerated, which would markedly reduce the steel corrosion rate. Note that the reaction with concrete contributes to the observation of negligible corrosion of embedded of steel noted in References 9.5.4 and 9.2.5-see bullets below.-A study conducted by EPRI (Reference 9.5.5) concluded that the corrosion rate of steel in a de-aerated boric acid solution is 0.004 mm/year (0.157 mils/year).
The study considered a range of temperatures and acid concentrations.
The corrosion rate of 0.157 mils/year is for a 2400 ppm boron solution, which is consistent with SFP chemistry.
This is conservative with regard to the situation in the FHB because the pH when the boric acid reaches the rebar will increase from reaction with the* concrete.-Testing documented in Reference 

====9.5.4 showed====
negligible reinforcing steel attack from boric acid flow through a simulated crack after a period of two years; corrosion limited to scarring in the area of the crack. The tests covered a range of pressures, crack sizes and pH. The tests showed that corrosion increases as crack width increases and pH decreases.
The observation of negligible corrosion was for the most aggressive conditions-widest crack (0.4 mm) and lowest pH (5.2). The lowest pH tested is similar to the pH of the SFP.-Experience at another US PWR showed no visible corrosion of embedded reinforcing steel from boric acid migration through a crack over several years (Reference 9.2.5). The source of the boric acid was SFP leakage and the concrete was six feet thick, which are similar to the situation at Salem.The Salem FHB does not show any signs of significant degradation of rebar from exposure to boric acid.-Rust staining on the walls in the sump room is very minor and the result of very small amounts of iron oxide.-An independent structural examination by an experienced concrete structural engineer concluded that the structure is sound and that there are "no indications of concrete surface expansion due to reinforcing steel corrosion was would be MPR-2613 Revision 3 6-9 evidenced by a pattern of cracking, spalling or bulging of the concrete" (Reference 9.2.6).Migration through construction joints or cracks is a relatively recent event at Salem that stopped in 2003. Migration through the construction joints or cracks started prior to 2002 (when leakage into the Auxiliary Building was noted), possibly as early as the 1995 to 1998 timeframe.
Given the thickness bf the walls, boric acid migrating through the walls would not have reached the outer reinforcing steel until well after the 1995 to 1998 timeframe.
Reports of leakage into the Auxiliary Building and sump room stopped subsequent to cleaning the telltales in early 2003.Hence, the outer reinforcing steel was exposed to boric acid which migrated through construction joints or cracks for much less than 5 to 8 years. This mode of degradation is not expected to recur as PSEQ Nuclear has implemented multiple measures to ensure that the telltales do not become entirely blocked (trending of telltale leakage rates, periodic videoprobe inspections and cleanings).
The preponderance of the evidence is that any degradation of reinforcing steel, particularly the outer reinforcing steel, is negligible.
Based on Reference 9.5.5, the corrosion rate of the reinforcing steel in a de-aerated boric acid solution is 0.157 mils/year.
For an exposure duration less than 7 years, the rebar has experienced a reduction in radius of less than 1 mil (0.001 inch).This is conservative estimate with regard to the situation in the FHB because the pH when the boric acid reaches the rebar will increase from reaction with the concrete.It is important to note that any rebar degradation is limited to the immediate vicinity of the crack or construction joint. Testing documented in Reference 

====9.2.4 showed====
that wicking rate of boric acid along the reinforcing steel/concrete interface is about the same as the rate boric acid penetrates into the concrete.
Any de-bonding of concrete from the reinforcing steel is localized.
Accordingly, reinforcing steel functionality is maintained.
6.7 CONCl.USIONS The conditiori of the FHB at the end of plant is projected using the foregoing discussions of the expected nature and timeline of building degradation.
The projections consider degradation that has occurred to date and anticipated future degradation.
The tables below summarize the projections; note that local degradation from weld leakage and general degradation from accumulation 
!of boric acid behind the liner are combined into a single table. The depth of concrete degradation could be as high as 1.30 inches in the slab and 0.44 inch in the walls. It is estimated thai the reinforcing steel has experienced a reduction in radius of less than 0.001 inch, which is negligible.
0 MPR-2613 Revision 3 6-10 Table 6-1. Projected Degradation of Concrete Structure:
Local Degradation from Weld Leakage and General Degradation from Water Trapped Behind Liner Parameter Value Basis Concrete Degradation Effective Los s of 0.44 inch For the walis, loss of concrete is based on general degradation.
Concrete in Wails (based on Aithough locai areas near leaking piug welds could be degraded general to deeper depths, there is no mechanism for expanding these degradation) iocal areas to a significant area. Further, structural margin is driven by the condition of the general area, not smali locaiized areas.The period of general degradation started sometime between 1995 and 1998 and extended to 2003. The maximum time period of 8 years is used.Using the projected degradation curve in Appendix D of Reference 9.2.4 and the temperature adjustment, the projected degradation over an 8-year interval is about 0.44 inch.Effective Loss of 1.30 inches For the slab, loss of concrete is based on local degradation.
Concrete in Slab (based on local Plug weld ieakage results in locai degradation.
However, degradation) blockage of the telltales expands the area of the slab subjected to long term exposure to boric acid as the boric acid migrates to an open telltale.Local exposure of the wall and slab to boric acid leakage started sometime before 1995. Local degradation from plug weld leakage and local degradation from leakage migration will continue into the future. For conservatism, this mode of degradation is assumned to occur over a period of 70 years. As the boric acid leakage may have puddled on the slab and created "potholes" of Indeterminate size, the depth of local degradation is applied to the entire slab.Reference 

====9.2.4 projects====
a depth of degraded concrete of 1.30 inches after 70 years exposure to boric acid.Reinforcing Steel Degradation
___________________________
Reinforcing Steel None No degradation of the reinforcing steel is expected because the Corrosion.
depth of concrete cover (3 Inch minimum) exceeds the effective loss of concrete.MPR-261 3 1 Revision 3 6-11 Table 6-2. Projected Degradation of Concrete Structure:
Reinforcing Steel Degradation from Migration through Joints/Cracks Parameter Value Basis Reinforcing Steel Possibly Boric acid migration through the construction cracks started prior Exposure Time <7 years to 2002 (when leakage into the Auxiliary Building was noted), possibly as early as the 1995 to 1998 timeframe.
However, given the thickness of the walls, the boric acid would not have reached the outer bar until well after the 1995 to 1998 timeframe.
Reports of leakage into the Auxiliary Building and sump room stopped subsequent to cleaning the telltales in early 2003.Reduction in <1 mil Since the outer rebar was exposed to boric acid longer than two Reinforcing Steel (<0.001 inch) years, degradation may be greater than the "negligible" noted in Radius the Reference 

====9.5.4 study====
and the experience at another US PWR. Based on the study documented in Reference 9.5.5, a reduction in radius of 0.001 inch (1 mil) is predicted.
This is conservative with regard to the situation in the FHB because the pH when the boric acid reaches the rebar will increase from reaction with the concrete.Length of Localized The testing documented In Reference 

====9.2.4 showed====
that the Degradationi at wicking rate was low In acidic conditions.
Therefore, the Concrete/
degradation would not spread considerably within the seven Reinforcing Steel years the reinforcing steel is assumed to have been exposed to Interface boric acid; i.e., degradation is localized to the immediate vicinity of the joint/crack.
MPR-2613 Revision 3 6-12 7 As~essment of Structural Adequacy i The structural adequacy of the FHB can be evaluated using the estimated concrete and reinforcing steel degradation levels along with structural calculations for the FHB structure.
Each of the degradation mechanisms discussed in Section 6.6 is addressed below to assess the current condition of the FHB.7.1 Concrete Degradation from Boric Acid Exposure The discussion in Section 6.6 concludes that the depth of concrete degradation may reach up to 0.44 inch on the walls and 1.30 inches on the slab. As shown in Reference 9.3.5 (provided in Appendix D) the impact of the degradation on the structure is contingent upon the section location within the SFP. The effects of degradation on the slab and walls are considered below.7.1.1 Slab Degradation As documented in Reference 9.2.3, the structural analysis of the slab does not credit the 6-inch layer of leveling concrete shown in Reference 

====9.4.6. Although====
no credit is taken for this concrete in any of the previously performed structural analyses, this layer of concrete is critical to understanding degradation depths.The maximunli estimated degradation depth of 1.30 inches would not penetrate this leveling layer and thus has -no impact on the structural capacity of the slab. Accordingly, no structural concrete is lost and thel margins documented in Appendix C of Reference 9.2.3 are unaffected.
7.1.2 Wall Degradation For the FHB walls, the following equations were developed in Reference 9.3.5 to relate the percent reduction in allowable moment (y) to a concrete degradation level (x), in inches.North Wall: y =l.O0x South Wall: y- 1.55x East Wll: y= 1.50x West Wall: y = 0.92x The pool-side:
of the FHB structure experienced concrete degradation to a depth of 0.44 inch during the time when boric acid leakage was trapped behind the liner. While local areas may experience more severe degradation from plug weld leakage, there is no mechanism for expanding thdse local areas to a significant area.0 MPR-2613 Revision 3 7-1 Further structural margin is driven by the condition of the general area, not small localized areas.Accordingly, general degradation of the wall is a more meaningful value to use for structural integrity calculations.
Using each of the above equations along with the predicted depth of general concrete degradation, percent reductions in allowable moments are obtained.
Table 7-1 compares the percent reductions to the limiting available margins of each FHB wall to obtain the available margin in the FHB structure considering projected degradation from boric acid between the liner and concrete..
Limiting available design margins, taken from Appendix C of Reference 9.2.3, are based on working stress design methods.The limiting available margin is 1.6% in the middle section at the top of the West wall. At this region of the pool, concrete degradation is expected to be minimal due to the limited time the section should have been exposed to the boric acid. The concrete in this region would have only experienced sustained contact with boric acid during the time when the telltales were plugged and the pool had completely filled. As previously discussed, any leakage from plug welds in the FHB walls would be expected to only degrade concrete in the immediate vicinity of the flow path of the leakage down the wall. Weakening of the cement paste at these localized areas would not significantly impact the structural integrity of the wall.Positive margin is maintained for all walls given the projected depth of degradation.
Hence, the design basis analysis of record is not invalidated by the postulated degradation.
Table 7.1. FHB Available Margin Based on 0.44" General Concrete Degradation and Working Stress Design (WSD) Methods Available Percent Available Margin Location of Margin (WSD, FHB Wall Reduction in (WSD, No Lmitin Mg Mnruing Capacity Degradation)
Limiting Margin Including~Degradation)
North 0.4% 4% Middle, Bottom 3.6%South 0.7% 300% West,9%Toward Bottom 299_" Middle, East 0.7% 5% Mdl,4.3%Toward Bottom West 0.4% 2% Middle, Top 1.6%7.2 Reinforcing Steel Degradation from Migration through JointslCracks Reinforcing steel degradation from boric acid migration through cracks and construction joints has a negligible impact on the FHB structural capacity.MPR-2613 Revision 3 7-2 Key considerations are as follows.As noted in Table 6-2, migration of boric acid through construction joints and cracks has potentially degraded reinforcing steel by less than 0.001 inch (radial reduction).
Using the equations provided in Appendix C, the maximum calculated reduction in margin is 0.2%, which is insignificant and well below the accuracy of the calculations.
* Fabrication tolerances for reinforcing steel are specified as 94% of total weight (Reference 9.1.5). For the reinforcing steel sizes used in the FHB, this is equivalent to a diametral variation of approximately 3% or about 30 times the estimated degradation.
In light of this tolerance, it is apparent that any predicted reinforcing steel degradation is negligible relative to the imperfections that are inherent to the steel in its original form.* As discussed in Section 4, the available margin for all sections under consideration may be increased by 2% if the actual yield strength of the FHB reinforcing steel is used in the working stress design calculation.
The actual yield strength compensates for the predicted reduction in margin by more than a factor of 10.Based on the above, there is no reduction in structural margin from potential reinforcing steel degradation from boric acid leakage through cracks and construction joints.The conclusion on the adequacy of the reinforcing steel does not change even if reinforcing steel corrosion is assumed to occur over the entire 70-year period considered herein. Using the corrosion rate.of carbon steel in de-acrated boric acid from Reference 9.5.5, the radial reduction is 0.011 inch after 70 years. Using the equations in Appendix C, the maximum calculated reduction in margin is about 2%, which is equal to the increase in margin that can be recovered by crediting the actual yield strength of the reinforcing steel in the working stress design calculation.

===7.3 Voided===
Areas beneath the Liner As discussed previously, boric acid will attack the cement paste, weakening it and causing it to-de-bond froml the coarse and fine aggregate.
As the degradation progresses, a rubble bed of coarse and firie aggregate may be formed on top of the concrete as the cement in the top layer fully degrades.
In essence, the local degradation will create a "pothole" with sand and coarse aggregate on top. of the remaining concrete.
This effect may produce a small voided depth below the 1/4-inch stainless steel liner, but above the. sand and rubble layer. With this void there is a concern that the load of the fuel racks may no longer be supported on a firm surface.As stainless steel is a highly ductile material, it is. expected to strain and deform to the voided depth without failure. Adequacy of the liner with the degraded under-layer was verified in a scoping assessment.
The assessment considered both the water pressure load and fuel rack foot load. As discussed below, neither of these mechanisms are considered likely to cause liner failure.MPR-2613 Revision 3 7-3 Reference 

====9.2.4 calculated====

a degraded paste depth of 1.30 inches. This value considers the depth of cement that would be affected by the boric acid, but is not representative of the voided depth.The coarse and fine aggregate constitute approximately 71% of the volume of the concrete and 79% of the mass of the concrete.
Although a small portion of the concrete constituents may have migrated to the telltales, the majority of the constituents (including almost all of the aggregate) are expected tp remain in place. Assuming that 71% of the concrete constituents remain, the voided depth is expected to be no greater than 0.38 inch.At a depth of 0.38 inch, the water pressure load (17.77 psi from Reference 9.3.1) and the single foot load (maximum of 62,600 lbs over a 12-inch by 12-inch pad, Reference 9.3.1) will likely plastically deform the liner to the rubble bed. The amount of strain experienced by the liner over this small depth is expected to be significantly less than the limiting strain of the material (<10%)and will not cause failure.7.4 Conclusion The FHB is structurally adequate through the end of plant life. As Table 7-1 shows, the structural capacity of the FHB is maintained for all degradation modes. The provided values are for the highest degradation conditions in the most limiting location in the pool; all other areas of the pool show higher available margin. Positive margin is maintained at all locations in the structure.
Therefore, the design basis analysis of record is not invalidated by the postulated degradation.
A scoping assessment further demonstrates that the liner is sufficiently ductile to accommodate the load from the fuel racks even if the foot of the rack is positioned over an area of local concrete degradation.
MPR-2613 Revision 3 7-4 8 Corroboration of Salem Assessment by Cores from the CY SFP This section summarizes results from evaluation of cores removed from the floor of the Connecticut Yankee Atomic Power Plant (CY) SFP and demonstrates how the results corroborate the degradation modes and degradation projections in preceding sections.In the fall of 2007 PSEG Nuclear learned that EPRI had possession of several samples taken from the CY SFP during decommissioning.
The samples were taken in the form of cores that included the liner, leakage channel and concrete.
B&W Technical Services Group, under contract from EPRI, had evaluated the liner welds to attempt to locate the leakage source and to support development of inspection and repair techniques.
A thorough evaluation of the concrete had not been performed.
PSEG Nuclear and EPRI agreed to collaborate on evaluation of the concrete.
The objective of the evaluation was to use actual plant observations to corroborate the assessment of the Salem FHB.8.1 EVALUATION OF CY CORES The evaluation of the CY Cores was performed by Concrete Research & Testing (CRT), MPR's subcontractor for the testing described in Section 5.3, with assistance from B&W Technical Services Group. Reference 

====9.2.7 documents====

the examination of the CY cores.8.1.1 Overview of CY SFP The CY SFP was a reinforced concrete structure with a 1/4-inch stainless steel liner. The pool had leakage collection channels located behind the liner seams. The channels are 3-inch wide by 1/22-inch thick stainless steel plates with a 1-inch wide by 1/4-inch deep groove at the centerline.
The liner plates were plug welded to the channel near the seam weld to align and support the plates for the closure weld. The channel is held in place by Nelson studs embedded into the concrete.It is likely that there were other embedded studs located in between the channels or alternate means for supporting the liner between seam welds, but the exact construction details are not known.In leakage of water from behind the liner was noted during decommissioning.
Specifically, after the pool was drained and dried, pools of water were noted in multiple locations on the floor. It was suspected that cracking in the liner allowed water from behind the liner to leak into the pool.The source of the water was either SFP leakage that had been trapped behind the liner or ground water.MPR-2613 Revision 3 8-1 Personnel who worked at CY during plant operation and decommissioning indicate that the SFP leakage was believed to have started early in plant life. CY began commercial operation in 1968 and was shutdown in 1996. Removal of fuel from the pool was completed in 2005. Therefore, the leakage occurred for approximately 37 years.8.1.2 Description of Cores Three cores were provided to EPRI and subsequently made available to PSEG Nuclear for examination.
All three cores were from the floor of the SFP. The cores were taken from locations where water pooled after the fuel was removed and the pool was drained. The specimens are 6-inch diameter by 10-inch long cores. The cores included the liner, channel, concrete and embedded reinforcing steel. Two cores had been cut into disc specimens prior to shipment to EPRI. Note that the cores as examined by CRT did not include the liner sections as the liner had been removed as part of EPRI's evaluation of the liner welds.8.1.3 Evaluation of Concrete The cores were examined to characterize the concrete and to evaluate potential degradation of the concrete from boric acid attack. The examinations included petrographic examination of the cores by CRT personnel and chemical analyses by B&W Technical Services Group. The chemical analyses determined the presence of boron at varying depths from the top surface and to characterize secondary deposits observed in the concrete.
The discussion below is based on Reference 9.2.7.Petrographic Examination Examination of the cores showed that both the coarse and fine aggregates are non-reactive 4 with respect to acid attack. Specifically, the coarse aggregate is diabase igneous rock and the fine aggregate is primarily quartz. CRT judged the cement paste to be fair quality with a water-cement ratio of 0.60.The upper surface of the cores (i.e., the surface underneath the liner) showed evidence of boric acid attack. The concrete exhibited a light color and the paste was weak. In some cases, aggregate particles were exposed from loss of cement paste. Chemical analyses confirmed the presence of boron.Table 8-1 lists the depth of degradation from each of the cores. As shown, the depth of degradation was typically minor (<0.3 inch). However, in some local areas the depth of degradation was markedly higher (up to 0.91 inch). The deepest areas were adjacent to the channel. Figure 8-1 shows Core 122 before it was sectioned.
As shown, the deepest degradation is on one side of the channel where it appears that the channel had debonded from the concrete allowing the boric acid to access the concrete underneath below the channel. When the core was being sectioned for petrographic examination, it failed at a pre-existing crack. Figure 8-2 shows 4 In the context of this report, "non-reactive aggregate" is used to denote either coarse or fine aggregates that do not react with acids. Non-reactive coarse aggregates include igneous rock and non-reactive fine aggregates include silica sand. Limestone and carbonate-based aggregates are considered reactive, MPR-2613 Revision 3 8-2 0 this fracture and its orientation is consistent with the deepest degradation. (Note that Figure 8-2 is from the opposite perspective as Figure 8-1.)Table 8-1. Depth of Affected Paste in CY Cores Core Depth of Affected Paste (inch)Away from Channel Adjacent to Channel Below Channel 122 0.06-0.12 0.28-0.91 0.20-0.31 123 0.05-0.12 
<0.05 124 0.12 -0.16 0.30-0.67 0.03- 0.14 Figure 8-1. Core 122-As Received Core prior to Sectioning MPR-2613 Revision 3 8-3 Figure 8-2. Core 122-Cross-Section View Showing Fracture along Crack The three cores exhibited cracking as noted below.* Core 122 had a vertical crack at the comer of the channel.* Core 123 had a horizontal crack at the location of the top layer of reinforcing steel.* Core 124 had three vertical cracks under the leakage channel.Only the crack in Core 122 contributed to degradation of the concrete.
This crack was the only crack that connected to the top surface of the concrete in an area wetted by boric acid. The horizontal crack in Core 123 did not connect to the surface. The vertical cracks in Core 124 connected to the surface of the concrete underneath the channel, but the lack of concrete degradation under the channel indicates that this area was not wetted by boric acid.These results demonstrate that boric acid attack of concrete can be highly localized depending on where the boric acid pools. Cracks may provide a means to expand the degraded area, but only if they connect to the surface in an area where boric acid is present. Cracks did not lead to widespread degradation.
Chemical Analyses Chemical analyses were performed on powder concrete samples drilled at various depths from the surface of the core. The analyses showed that boron was present and the boron concentration decreased with depth. These results confirm that the observed degradation is from boric acid attack.The secondary deposits in the concrete were analyzed as well. The deposits were typically ettringite and calcite. The only location where the deposits contained boron was the vertical crack in Core 122. Recall that boric acid penetration into the crack led to an expanded area of degradation.
MPR-2613 Revision 3 8-4 

====8.1.4 Evaluation====

of Reinforcing Steel.Cores 122, 123, and 124 all contained reinforcing steel. The cores were sectioned perpendicular to the reinforc~ing steel so reinforcing steel corrosion and the bond with the cement paste could be evaluated.
Examination of the rebar is documented in Reference 9.2.7.No corrosion was noted in any of the sections.
However, the examinations showed areas where the concrete separated from the underside of the rebar. This is considered to be the result of settlement of the concrete prior to hardening and insufficient consolidation of the concrete around the rebar; it is not due to boric acid attack.8.1.5 Evaluation of Liner Welds EPRI performed non-destructive and destructive examinations of the liner welds in cores removed from the CY SFP. The scope of the examinations included liner seam welds and a plug weld to the channel. The cores did not include any plug welds to embedded studs that may have been located between channels.
Reference 

====9.2.6 provides====
the details of some weld quality issues.Specifically, thiere was lack of fusion in the liner seam weld, an open root weld. Also, the plug welds were not completely filled. No through-wall defects were identified in the metallurgical evaluations.

===8.2 COMPARISON===

OF CY CORES TO SALEM SFP ASSESSMENT

====8.2.1 Comparison====

of Concrete The concrete u'sed at CY and Salem can be compared as follows.* The concrete at both CY and Salem has non-reactive aggregates.
* CY cores used fly ash while laboratory-prepared specimens used in the Salem long-term testing did not, Fly ash promotes hydration of the concrete and increases density. (Note that concrete used in structures at Salem contains fly ash.)* The CY concrete had a higher water-cement ratio than Salem (0.6 versus 0.5).Permealility of concrete increases significantly for water-cement ratios above 0.5.Additioi ially, the strength of concrete is. reduced as water ratio increases.
The net result of the differences identified in the second and third bullet likely increases the porosity of the CY specimens compared to the Salem test specimens.

====8.2.2 Concrete====
Degradation Degradation Modes The degraded concrete in the CY cores varied from less than 0.05-inch to 0.91-inch demonstrating that the degradation can be highly localized.
This is consistent with the postulated degradation of the Salem FHB as described in Section 6 of this report.MPR-2613 Revision 3 8-5 Depth of Degradation Over the 37 year life of CY's SFP concrete degradation reached a maximum depth of 0.91 inch.The correlation developed in Reference 9.2.4 from the Salem testing predicts 0.94 inch of degradation at 37 years. Therefore, the CY degradation is within the expectation for the given exposure time. The increased porosity of the CY cores, as described in Section 8.2.1 should yield a result deeper than the model from Reference 9.2.4.8.2.3 Rebar Corrosion Although the upper surface of the CY cores was degraded from boric acid attack, the embedded reinforcing steel exhibited no corrosion or loss of bond with the cement from boric acid attack.However, it appears that the embedded rebar was not exposed to boric acid. The concrete degradation did not extend to the depth of the rebar, which would expose the rebar to boric acid.Further, the cracks present in the concrete CY cores did not connect from surface or the degraded concrete to the rebar.It is important to note that the presence of secondary deposits, including secondary deposits in cracks found in the CY .cores, provides evidence that water migration occurred.
Yet the reinforcing steel exhibited no corrosion.
These results confirm that cracks in concrete and other concrete defects do not promote reinforcing steel degradation.

==8.3 CONCLUSION==
S Overall, the CY cores corroborate the results of testing for Salem and the projections for Salem.The maximum depth of concrete degradation in the CY cores is within that predicted using the correlation developed from the Salem testing. The rebar in the CY cores exhibited no corrosion even though the upper surface of the concrete was degraded by boric acid, the concrete was cracked and, based on the presence of secondary deposits within the concrete, there was water migration in the concrete.MPR-2613 Revision 3 8-6 
.9 References

===9.1 SPECIFICATIONS===
9.1.1 Deleted.9.1.2 PSEG Nuclear Technical Standard SC.DE-TS.ZZ-4201 (Q), "Salem Structural Design Criteria," Revision 2.9.1.3 ACI 318-63, "Building Code Requirements for Reinforced Concrete," American Concrete Institute, June 1963.9.1.4 ACI 349-80, "Code Requirements for Nuclear Safety Related Concrete Structures," American Concrete Institute, April 1981.9.1.5 ASTM A615, "Standard Specification for Deformed and Plain Billet-Steel Bars for Concrete Reinforcement." O 9.2 REPORTS 9.2.1 Deleted.9.2.2 PSEG Nuclear Chemistry Technologies 
& Support Final Report, "Investigations of Salem Unit 1 Fuel Pool Leakage: Phase II Analyses," 2/21/2003.
9.2.3 MPR-1863, "Salem Generating Station Spent Fuel Pool Building Structural Design Analysis," Revision 0. (PSEG Nuclear VTD 326116)9.2.4 MPR-2634, "Boric Acid Attack of Concrete and Reinforcing Steel," Revision 2.(PSEG Nuclear VTD 326561)9.2.5 PSEG Nuclear Record Transmittal No. DES-060005; contains documents related to evaluation of concrete degradation from boric acid at another PWR.9.2.6 PSEG Nuclear VTD 327194, "Salem Units I and 2 Structural Examination of Spent Fuel Pool Structures," Revision 1.9.2.7 CRT Report No. R-140, "Petrographic Examination of Concrete Cores Removed from the Conn-Yankee Spent Fuel Pool," dated September 11, 2008. (included in Appendix A)MPR-2613 Revision 3 9-1 

===9.3 CALCULATIONS===
9.3.1 PSEG Nuclear Calculation 6S0-1674, "Structural Analysis Report for the Salem Generating Station Spent Fuel Pool Storage," Revision 0, Holtec International.
9.3.2 EQE Calculation 200050-C-01, "Salem Spent Fuel Pool Evaluation for Beyond Design Basis Thermal Load," Revision 0, EQE Engineering.
9.3.3 Deleted.9.3.4 MPR Calculation 108-275-02, "Statistical Analysis of Rebar Yield & Tensile Strengths for Salem Nuclear Generating Station," Revision 0 (included in Appendix B).9.3.5 MPR Calculation 0108-0275-34, "Salem Spent Fuel Pool Structure Capacities Based on Degraded Concrete Conditions," Revision 0 (included in Appendix D).9.3.6 PSEG Nuclear Calculation 6S 1-1836, "Justification for Acceptability of Leakage from the Spent Fuel Pool -Salem Unit 1," Revision 0.9.3.7 MPR Calculation 0108-0275-35, "Salem Spent Fuel Pool Reinforcing Steel Load Capacity at Degraded Conditions," Revision 0 (included in Appendix C).9.4 DRAWINGS 9.4.1 PSEG Nuclear Drawing No. 201075 A 8706-2, "No. 1 Unit -Fuel Handling Area, Plan at Elevation 78'-0", Revision 2.9.4.2 PSEG Nuclear Drawing No. 201076 A 8706-4, "Salem Nuclear Generating Station, No. 1 Unit -Fuel Handling Area, Plan at El. 84'-0"." 9.4.3 PSEG Nuclear Drawing No. 201077 A 8706-8, "Salem Nuclear Generating Station, No. 1 Unit -Fuel Handling Area, Plan at El. 100'-0" and 116'-0"." 9.4.4 PSEG Nuclear Drawing No. 201078 A 8706-4, "Salem Nuclear Generating Station, No. 1 Unit -Fuel Handling Area, Plan at El. 130'-0"." 9.4.5 PSEG Nuclear Drawing No. 201079 A 8706-3, "Salem Nuclear Generating Station, No. 1 Unit -Fuel Handling Area, Roof Plan." 9.4.6 PSEG Nuclear Drawing No. 201080 A 8706-7, "Salem Nuclear Generating Station, No. 1 Unit -Fuel Handling Area, Sections A-A & B-B." MPR-2613 Revision 3 9-2 9.4.7 PSEG Nuclear Drawing No. 201081 A 8706-6, "Salem Nuclear Generating Station, No. 1 Unit -Fuel Handling Area, Sections C-C, D-D & E-E." 9.4.8 PSEG Nuclear Drawing No. 201082 A 8706-5, "Salem Nuclear Generating Station, No. 1 Unit -Fuel Handling Area, Sections F-F & G-G." 9.4.9 PSEG Nuclear Drawing No. 201085 A 8706-5, "Salem Nuclear Generating Station, No. 1 Unit -Fuel Handling Area, Elevation P-P & Str. Bar Schedule." 9.5 TECHNICAL PAPERS 9.5.1 A. Bajza, I. Rousekova, & M. Dubik, "Can Boric Acid Corrode Concrete?," International Symposium on the Non-Traditional Cement and Concrete, Brno, Czech Republic, June 11 -13, 2002.9.5.2 V. Pavlik, "Corrosion of Hardened Cement Paste by Acetic and Nitric Acids. Part I: Calculation of Corrosion Depth," Cement and Concrete Research, Vol. 24, No. 3, pp.551-562, 1994.9.5.3 A. Allahverdi and Frantisek Skvara, "Acidic Corrosion of Hydrated Cement Based Materials.
Part 1 -Mechanism of the Phenomenon," Ceramics -Silikaty, Vol. 44, No.3, pp. 114- 120, 2000.9.5.4 W. Ramm and M. Biscoping, "Autogenous Healing and Reinforcement Corrosion of Water-penetrated Separation Cracks in Reinforced Concrete," Nuclear Engineering and Design, No. 179, pp. 191 -200, 1998.9.5.5 R. E. Nickell, "Degradation and Failure of Bolting in Nuclear Power Plants," Volume 1, Electric Power Research Institute, NP-5769, April 1988.9.5.6 G. Frederick, "Fuel Pool Inspection and Repair," RRAC Technical Program Meeting, December 2007.9.6 OTHER DOCUMENTS 9.6.1 Deleted.9.6.2 PSEG Nuclear DCR 1EC0995, "Provide Weld Seam Encasements Along the Seal Weld Joints in the FH#1 Building Spent Fuel Pool." 9.6.3 Letter from J. Simons (MPR Associates) to A. Roberts (PSEG Nuclear), "Results from Literature Search; Salem Nuclear Generating Station Spent Fuel Pool Building Assessment," 3/10/03.9.6.4 Email from T. Taylor (PSEG Nuclear) to J. Simons (MPR), 

==Subject:==
Spent Fuel Pool Telltales, dated 12/15/03.MPR-2613 Revision 3 9-3 W 9.6.5 Email from T. Taylor (PSEG Nuclear) to J. Simons (MPR), 

==Subject:==
Telltale Pipe Deposit Status, dated 11/25/03.0 MP-61 -MPR-2613 Revision 3 9-4 A Petrographic Examination of Concrete Cores Removed from the Conn-Yankee SFP This appendix contains the following CRT report.CRT Report No. R- 140, "Petrographic Examination of Concrete Cores Removed from the Conn-Yankee Spent Fuel Pool," dated September 11, 2008.(This appendix originally contained a calculation that assessed the potential margin recovery if ultimate strength design was used for normal operation and OBE cases. The calculation was removed for Revision 2 of this report.)MPR-2613 Revision 3 A-1 REPORT NO. R-140 ON PETROGRAPHIC EXAMINATION OF CONCRETE CORES REMOVED FROM THE CONN-YANKEE SPENT FUEL POOL TO MPR ASSOCIATES ALEXANDRIA, VIRGINIA SEPTEMBER 11, 2008 0 REPORT NO. R-140 ON PETROGRAPHIC EXAMINATION OF CONCRETE CORES REMOVED FROM THE CONN-YANKEE SPENT FUEL POOL INTRODUCTION Concrete cores were removed from the spent fuel pooi of the decommissioned Connecticut Yankee nuclear powcr plant. Three of the cores were sent to B&W Technical Services Group in Lynchburg, Virginia for examination and testing of the stainless steel liner. All of the cores received at B&W were removed from the floor of the fuel pool.Subsequently, it was decided to perform petrographic examinations on the cores to determine if and to what extent the concrete has been affected by exposure to boric acid from within thc spent fuel pool.Results from the evaluation of the Conn-Yankee specimens will be used to augment MPR Associates previous assessment of the Salem Spent Fuel Pool leakage and Fuel Handling Building structure, including the long-term testing program.Nick Scaglione of Concrete Research & Testing was contracted to examine the concrete cores at the B&W facility in Lynchburg, Virginia.DESCRIPTIONS OF THE CORES The concrete cores were labeled 122, 123 and 124. The cores have a diameter of 5 %/ in. Core No. 122 is comprised of two sections each having a thickness of roughly 2 in. Section 122-1 represents the upper end of the core. and contains the imbedded stainless steel channel. Section 122-2 represents a depth of 2 to 4 in. below the top surface of the core. This section contains a No. 7 steel rebar oriented parallel to the end surfaces.
Photographs of the core are shown in Figure 1 of Appendix A.
Core 123 has a length of about 18 in. The core is separated into two pieces by a horizontal fracture located at a depth of 10 in. The top end of the core contains the embedded stainless steel channel.The core contains a No. 9 rebar at a depth of about 2 4 in. below the top surface. Photographs of this core are shown in Figure 2 of Appendix A.Core No. 124 is comprised of two sections each having a thickness of roughly 2 in. Section 124-1 represents the upper end of the core and contains the imbedded stainless steel channel. Section 124-3 represents a depth of 4 to 6 in. below the top surface of the core. This section contains a No. 9 steel rebar oriented parallel to the end surfaces.
Photographs of the core are shown in Figure 3 of Appendix A.EXAMINATION AND TEST METHODS Preliminary stereomicroscopic examinations were performed on the exterior surfaces of each core.Following this initial examination, the cores were saw-cut perpendicular to the end surfaces.
For the Core 122 and Core 124 sections the saw cuts were made perpendicular to the orientation of the steel channel and perpendicular to the orientation of the rebar. Core 123 was initially saw cut parallel to the end surfaces at depths of 1 2 in. and 4. 1/4 in. below the top surface of the core. The upper section containing the steel channel will be referred to as Core 123, Section 1. The lower section contains the steel rebar and will be referred to as Core 123, Section 2. These two sections were then saw cut perpendicular to the end surfaces of the core. The saw cuts were made perpendicular to the orientation of the steel channel and perpendicular to the orientation of the rebar. The saw-cut sections of each core were prepared for microscopic examination by lapping with silicon carbide pads.The lapped sections of the cores were examined under a stereomicroscope at magnifications of up to 50X. The examinations were performed following the guidelines outlined in ASTM C 856, "Standard Practice for Petrographic Examination of Hardened Concrete." 2 The water-cementitious ratios of the core specimens were estimated based on observations and qualitative assessments of cement paste hardness, textural features, color, relative amount of unhydrated cement particles and rate of water absorption.
Features of the cement paste when probed (steel probe) were also used in this assessment.
The cores were tested to determine the depth of the boron penetration into the concrete.
The testing for boron concentration was performed using Inductively Coupled Plasma / Mass Spectroscopy (ICP/MS).
For this testing powdered concrete samples were obtained from each core by drilling with a 1/4 in. diameter carbide drill bit at various depths below the top surfaces of the cores. Photographs showing the drilling procedure are shown in Figure 4. The boron testing was performed by B&W.Secondary deposits observed in the concrete specimens were analyzed using a Scanning Electron Microscope (SEM) with Energy Dispersive X-Ray Spectroscopy (EDS) capability.
EDS analyses were performed to determine chemical compositions of the secondary deposits.
The SEM/EDS analyses were performed by B&W.EXAMINATION RESULTS General Description of Concrete The concrete represented by the cores can be described as a non air entrained concrete containing a 1 in. maximum size crushed coarse aggregate and a natural sand. The coarse aggregate particles are comprised predominantly of diabase igneous rock particles.
The fine aggregate is comprised predominantly of quartz particles.
The sand also includes smaller amounts of igneous rock lithics, feldspar particles, siltstone particles and hematite particles and minor amounts of mica, pyroxene and amphibole particles.
The concrete contains both portland cement and fly ash as the cementitious constituents.
The cement paste is judged to be of fair quality having an estimated water-cementitious ratio of 0.60. The concrete represented by the cores is well consolidated.
3 Core No. 122 -Section 1 Preliminary Exam The top surface of the core exhibits a light color relative to the bulk of the core. This surface exhibits a light loss of cement paste, with fine aggregate particles and portions of a few coarse aggregate particles exposed. The cement paste comprising the top of the core is very weak and highly absorptive.
The exterior surface of the core shows cement paste near the surface of the core which exhibits a distinctly lighter color than the cement paste lower in the core. Photographs showing this feature are presented in Figure 5 and Figure 6. These light colored areas represent cement paste that has been degraded by the boric acid exposure.
As can be seen by the photographs, the depth of the degraded cement paste is significantly deeper adjacent to the steel channel compared to the areas away from the steel channel. The depths of affected cement paste are shown in Figure 5 and Figure 6.The core fractured while being clamped into place for saw cutting. Secondary deposits observed on the fracture surface indicate that the fracture was due to a pre-existing crack present in the concrete.The fracture surface of the core showing the secondary deposits is presented in Figure 7.Examination of Lapped Section The lapped section of the core is shown in Figure 8. As can be seen in the photograph the stainless steel channel debonded from the concrete.Depth of Degraded Cement Paste The examination of this section shows degraded cement paste away from the channel to depths ranging from 1.5 mm to 3 mm below top surface of the core. Adjacent to the channel the cement paste is degraded to a depth of about 7 mm on one side and a depth of about 23 mm on the other side.The deeper depth of degradation correlates with the area of the pre-existing crack. Below the channel 4 the cement paste is degraded to typical depths of 5 to 8 mm, although one area is degraded to a depth of less than 1 mm.The cement paste affected by the penetration of boric acid is lighter in color and is significantly weaker and more absorptive than the cement paste in the bulk of the core.Secondary Deposits The concrete contains white colored secondary deposits within the air voids. These deposits typically have a fibrous habit. The secondary deposits are very common in air voids just below the degraded cement paste (see Figure 9). Occasional secondary deposits were observed throughout the full depth of the core section.Examination of Fractured Surface The fractured surface of the core was examined under the stereomicroscope.
As previously mentioned, the fractured surface contains white secondary deposits.
These deposits can be seen with the unaided eye and are present on only the lower portion of the section (see Figure 7). The concrete section has a thickness of about 1.8 in. The cement paste comprising the surface of the fracture plane has been degraded by boric acid penetration to a depth of about 0.9 in. The area of the affected cement paste does not contain the secondary deposits.The crystal morphology of the secondary deposits could not be discerned under the stereomicroscope.
SEM/EDS analyses were performed to identify the composition of the material.
The SEM/EDS analyses identified boron as a major component of the secondary deposit material.
The boron compound was not identified.
An EDS spectrum showing the presence of boron is shown in Figure 10.5 
.Examination of Bottom Lapped Surface The bottom surface of Core 122, Section 1 was lapped to examine the concrete surrounding the steel stud that was used to anchor the stainless steel channel. The steel stud has a diameter of roughly Y2 in.and is shown in Figure 11. The concrete is separated from the majority of the stud circumference (see Figure 11). The cement paste surrounding a portion of the stud appears slightly softer than typical.The thickness of this cement paste is only about 1 mm. It is unclear if this area has been affected by boric acid.Two parallel cracks extend from the stud into the concrete.Core No. 122 -Section 2 The lower section of Core 122 contains a No. 7 steel reinforcing bar (% in. diameter).
The core* section was saw cut perpendicular to the rebar. The examination of the lapped section revealed that the rebar does not exhibit corrosion.
Void areas are present below the rebar and at the lower sides of the rebar. These voids appear to be related to both settlement of the concrete and to insufficient consolidation of the concrete around the rebar.Cracks are present in the concrete that extend from the sides of the rebar up to the top of the core section. A vertically oriented crack was also observed within the concrete.
This crack extends from the bottom of the rebar to the bottom of the core section.0 6 Core No. 123 Preliminary Examination The preliminary examination of the core revealed the presence of a horizontal crack located at a depth of 2 3/4 in. below the top surface of the core. The crack is present at the midpoint level of the No. 9 rebar. This crack is shown in Figure 12.This examination also revealed the presence of a vertically oriented crack extending down from a comer of the stainless steel channel. The crack extended to a depth of 5 mm below the channel.Core No. 123 -Section 1 Examination of Lapped Section The lapped section of the core is shown in Figure 13. As can be seen in the photograph the stainless O steel channel debonded from the concrete.Depth of Degraded Cement Paste The examination of the section shows degraded cement paste at the surface of the core to depths ranging from 1.3 mm to 3.0 mm below top surface of the core. Below the channel the cement paste is degraded to a maximum depth of 1.3 mm. Some areas of the cement paste below the channel show no degradation.
The cement paste affected by the penetration of boric acid is lighter in color and is significantly weaker and more absorptive than the cement paste in the bulk of the core.Secondary Deposits The concrete contains white colored secondary deposits within occasional air voids throughout the* depth of the section. These deposits typically have a fibrous habit, although some having a blocky habit were also observed.7 Cracking Distress There was no cracking distress observed in the lapped section.Core No. 123 -Section 2 The lower section of Core 123 contains a No. 9 steel reinforcing bar (1 /8 in. diameter).
The core section was saw cut perpendicular to the rebar. The examination of the lapped section revealed that the rebar does not exhibit corrosion.
The concrete is separated from the underside of the rebar. The gap between the rebar and the concrete is about 0.8 mm. This feature is shown in Figure 14. The gap is judged to be due to settlement of the concrete below the rebar. A crack extends from one side of the rebar, upward at an angle of roughly 45'. This crack looks like a continuation of the settlement gap. The gap narrows into a tight crack (see Figure 14). The crack extends to the top of the section.On the other side of the rebar a horizontal crack extends from the rebar (see Figure 14). This crack is relatively tight and passes through a coarse aggregate particle.
This is the same crack shown in Figure 12.Core No. 124 -Section 1 Examination of Lapped Section The lapped section of the core is shown in Figure 15. In this core the stainless steel channel remained bonded to the concrete.Depth of Degraded Cement Paste In areas away from the channel, the cement paste is degraded to depths of 3 to 4 mm. The areas adjacent to the channel exhibit deeper levels of degradation.
One side of the channel has degraded cement paste to a depth of 7.5 mm, while the other side of the channel shows degraded cement paste 8
* to depths of up to 17 mm below the core surface (see Figure 15). Below the channel, the depth of the degraded cement paste ranges from 0.8 mm to 3.5 mm. The areas of the degraded cement paste are shown in Figure 15.The cement paste affected by the penetration of boric acid is lighter in color and is significantly weaker and more absorptive than the cement paste in the bulk of the core.Secondary Deposits The concrete contains white colored secondary deposits within occasional air voids throughout the depth of the section. These deposits typically have a fibrous habit, although some having a blocky habit were also observed.
An example of fibrous secondary deposits taken under the stereomicroscope is shown in Figure 16.0 SEM/EDS analyses were performed on some of these secondary deposits to identify their composition.
The fibrous secondary deposits were identified as ettringite (see Figure 17). Analyses of the blocky secondary deposits were not performed, although the deposits are likely calcite.Cracking Distress The core section contains three vertical cracks below the channel. One of these cracks is very tight and does not extend through the full depth of the section. Two of the cracks are adjacent to each other and extend the full depth of the section (see Figure 15). These cracks have an actual width of about 0.06 mm. The width of the cracks at the lapped surface ranges from 0.3 to 0.5 mm. The larger width than actual is due to material lost within the cracks during the specimen preparation.
The cement paste on either side of each crack does not appear to have been adversely affected by boric acid penetration.
It is possible that the missing material within the cracks was affected by boric acid.9 Other Features Several large entrapped air voids are present below the channel. These voids are due to insufficient consolidation of the concrete against the steel channel. The voids can be seen in Figure 15.Core No. 124 -Section 3 The lower section of Core 124 contains a No. 9 steel reinforcing bar (I '/ in. diameter).
The core section was saw cut perpendicular to the rebar. The examination of the lapped section revealed that the rebar does not exhibit corrosion.
Similar to Core 123, a settlement gap is present below the lower portion of the rebar. The gap has a width of 0.5 mm. Cracks extend from both sides of the rebar at 450 angles towards the top of the core section. One of the cracks extends to the top of the core section, while the other crack terminates about 13 mm from the rebar.A very tight vertical crack is present below the rebar. The crack extends from the rebar to the bottom of the core section.Examination of Channel Debris Minor amounts of debris were observed within the stainless steel channel of each core. The debris was examined under the stereomicroscope.
The material is comprised of a combination of cement paste particles, quartz particles and remnants of oxidized steel (rust). The rust particles are present throughout the debris and would indicate that this material was derived from the coring operation.
BORON CONCENTRATION Boron concentration testing was performed on the concrete at various depths in the cores. This testing was performed by B&W Technical Services Group using ICP/MS. The results of this testing as reported by B&W are provided in Appendix B.10 One measurement was performed on a sample taken from Core 123 at a depth of 14 in. below the top surface of the core. This sample was tested to determine the baseline level of the boron concentration for the concrete.
The boron concentration of this sample was measured at 36 ppm.For each core the boron concentration is highest near the top surface of the core. The core with the deepest depth of degraded cement paste (Core 122) exhibits the highest boron concentration (measured at 4370 ppm), while the core with the least amount of degraded cement paste (Core 123)shows the lowest boron concentration (measured at 806 ppm). These values are based on the measurements taken near the surface of the cores.Although the boron concentrations were always lower at the deeper depths compared with the top measurement, the concentrations did not always decrease with depth. In Core 122, the measured boron concentration at the depth of 1 9/16 in. is significantly higher than the measured boron concentration at the depth of 1 1/16 in. This unusual result can possibly be explained by the fact that the sample taken at the 1 1/11 in. depth was from only one drilling location (see photo in Appendix B).Although the sample location was an area of cement paste at the drilled surface, it is possible that the majority of the sample was taken through an underlying coarse aggregate particle, yielding a lower than expected boron concentration.
The test results show that boron has penetrated into the concrete to significantly greater depths than shown by the degraded cement paste. The measurements of the samples taken from the lowest-levels of the upper core sections (about 1 to 1 V2 in. depth) were significantly higher than the baseline level.Based on the examination, the boron penetration has not adversely affected the cement paste at the lower depths in the cores.11 Measurement on a sample taken at the depth of the steel rebar in Core 123 (2 2 in. below the core surface) showed a very low boron concentration of 64 ppm.COMPARISON OF THE CONN-YANKEE AND SALEM CONCRETES The concrete used for the Salem spent fuel pool was a non air entrained, portland cement concrete containing a diabase coarse aggregate and a natural siliceous sand. The concrete had a designed water-cementitious ratio of 0.49. The concrete used for the laboratory study at CRT used the same constituents and the same concrete mix design as the concrete of the Salem spent fuel pool.The concrete represented by the Conn-Yankee cores contain fine and coarse aggregates that are similar to the aggregates of the Salem concrete.
The major difference of the concrete represented by the Conn-Yankee cores with respect to boric acid penetration is the water-cementitious ratio. The water-cementitious ratio of the concrete represented by the Conn-Yankee cores was estimated at 0.60, which is significantly higher than the Salem concrete.
The permeability of cement paste is directly related to the water-cementitious ratio. As the water-cementitious ratio increases, permeability increases.'
A graph showing this relationship is shown in Figure 18. The Conn-Yankee concrete contains fly ash as a cementitious constituent, whereas laboratory prepared specimens for the Salem testing do not. The presence of fly ash in concrete decreases the permeability of the cement paste relative to a straight portland cement concrete having the same water-cementitious ratio.

==SUMMARY==
 & CONCLUSIONS Three concrete cores removed from the floor slab of the spent fuel pool at the decommissioned Connecticut Yankee Nuclear Power Plant were examined petrographically.
The examination was performed to determine the effect of the concrete's exposure to boric acid solution.12 It is apparent from the examination of the Conn-Yankee cores that boric acid solution leaked from the fuel pool and came into contact with the underlying concrete.
The solution apparently overflowed the stainless steel channels and flowed over the concrete surface. The depth of the affected cement paste is typically significantly deeper adjacent to the channel than away from the channel. This is due to the penetration of the boric acid solution along the channel/concrete interface.
The penetration of the solution at this site was likely facilitated by poor bonding of the concrete to the steel channel. If the rate of boric acid solution flowing out of the channel was low, much of the solution would have penetrated along the channel/concrete interface as opposed to flowing over the concrete away from the channel. Vertical cracks located below the side of the channel in two of the cores (Core 122 and Core 123) would have contributed to deeper penetration of the solution in this area.The boric acid solution has chemically attacked the near surface cement paste of the concrete.
In general, acids attack portland cement paste by decalcification (calcium leaching) of the hydrated cement compounds, in particular Calcium Hydroxide and Calcium Silicate Hydrate. The chemical attack has not adversely affected either the diabase coarse aggregate or the natural sand. The cement paste affected by the chemical attack has been significantly weakened relative to the unaffected cement paste. The attacked cement paste is also highly absorptive.
The cement paste comprising the immediate top surface of the cores is typically more severely deteriorated than the cement paste affected at lower depths.The cement paste affected by the boric acid in the Conn-Yankee cores is not as severely deteriorated as the cement paste attacked by the boric acid in the laboratory studies previously performed by CRT (CRT Report No. R-125 dated 12-12-03 and CRT Report No. R-125-3 dated 2-24-06).Based on the ICP/MS results (see Appendix B), boron has penetrated into the concrete well beyond the depth of the chemically attacked cement paste. This finding was also seen in the previous lab 13 work performed at CRT (CRT Report No. R- 125 dated 12-12-03).
Following the reaction of the boric acid solution within the near surface concrete, the reacted solution further penetrates into the concrete carrying the boron. The penetration of the boron has not adversely affected the quality of the cement paste at the lower depths.Another indication of solution moving through the concrete is the presence of secondary deposits (ettringite) observed in air voids at depths of up to 2 in. below the top surface of the cores. As water moves through concrete soluble salts are often dissolved out of the cement paste and re-deposited in air voids upon drying.Through the SEM/EDS analyses boron was detected as secondary deposits within a vertical crack plane of Core 122. The crack extended from the top of the core to the bottom of the section (1.8 in.depth). Although the boric acid would have had direct access to this crack, the cement paste was adversely affected to a depth of only 0.9 in.Nick Scaglione, PWsident Concrete Research & Testing, LLC.14 References
: 1. Powers, T.C., Copeland, L.E., Hayes, J.C., Mann, H.M. "Permeability of Portland Cement Paste" Proceedings, American Concrete Institute, Volume 51, 1954, pp. 285-297.0 15 APPENDIX A Figure 1. Photographs of Core 122. The upper photograph shows Section 1 (upper 2 in. of core). The stainless steel channel can be seen imbedded in the concrete.
The lower photograph shows Section 2 (2 to 4 in. depth).A-1 Figure 2. Photographs of Core 123. The upper photograph shows the top 10 in. of the core.The lower photograph shows the core from a depth of 10 to 18 in. These two pieces were originally connected at surface "A".A-2 Figure 3. Photographs of Core 124. The upper photograph shows Section 1 (upper 2 in. of core). The lower photograph shows Section 3 (4 to 6 in. depth).A-3 Figure 4. The photographs show the procedure used to remove the powdered concrete samples for the boron concentration testing.A-4 Figure 5. Exterior surfaces of Section 122-1. The light colored cement paste near the top surface of the core represents the degraded cement paste. It can be seen that the depth of the degraded cement paste in the vicinity of the channel is significantly greater than the depth of the degraded cement paste away from the channel.A-5 S Ahm Figure 6. Exterior surfaces of Section 122-1. The light colored cement paste near the top surface of the core represents the degraded cement paste. It can be seen that the depth of the degraded cement paste in the vicinity of the channel is significantly greater than the depth of the degraded cement paste away from the channel.A-6 Figure 7. Fractured surface of Section 122-1. This portion of the core broke from the larger section during the saw cutting operation.
White secondary deposits can be seen on only the lower portion of the sample where the concrete has not been affected by the boric acid penetration.
A-7 0 aFigure 8. Cross-section view of Core 122, Section 1 (lapped surface).A-8 Figure 9. Cross-section view of Core 122, Section 1 taken under the microscope.
The 0 photograph shows small air voids filled with white secondary deposits.
The air voids within the cement paste affected by the boric acid penetration does not contain the secondary deposits.A-9 133405-1 Specimnen:
Concrete Piece 12Z Uncoated, 5 kW.Elernemiiw.
B, r, q. Al, SI, Ca Probable Trace: Na, Mg, S B Ca C 1 Cummo--5.00S keW 0 cat ID -
indlJaw0.00S 
-40.955= 17155icn1
*Figure 10. SEM Photograph and EDS analysis of secondary deposits on the fracture surface of Core 122 (see Figure 7). The EDS analyses of the secondary deposit material shows boron as a major element.A-10 0 Figure 11. The upper photograph shows the bottom surface of Core 122, Section 1. The lower O photograph was taken under the microscope and shows the stud/concrete interface.
Note the separation between the concrete and the stud (black arrows) and the crack (blue arrows)A-11 0* l,1igtire 
: 12. I'Iiotograpli of (ore 123 showinag cracking distress at the level of the relbar. The crack was highlighllte(l with black ink so it could boe seen iii the photograph.
A-17 0 Figure 13. Cross-section view of Core 123, Section 1 (lapped surface).
The upper left portion of the specimen was fractured during the removal of the stainless steel liner (see Figure 2).A-13 S Figure 14. Cross-section views of Core 123, Section 2. The photographs show the cross-section of the No. 11 steel rebar. A gap can be seen at the lower rebar/concrete interface.
This gap is due to settlement of the concrete prior to hardening.
Cracking on either side of the rebar can also be seen.A-14 Figure 15. Cross-section view of Core 124, Section 1 (lapped surface).
The dashed yellow line shows the area of the degraded cement paste. Two vertically oriented parallel cracks are shown by the arrows.A-15 S Figure 16. Photograph taken under the stereomicroscope showing the presence of fibrous, white secondary deposits present within an air void.A-16 Specimen:
Concrete Section I EDSot area 1, mage 132972.Uncoated, 15 kV'.Elements:
C, 0, Mg, Al. SI, S, K, Ca, Fe Probable Trace: Na l'osnibte Trace; fI (Below ML3L)0 Light Elements using 2 kVW beam.0 I Al I Boron marked, not positively detected.IiA LFC K Ha ria I i i m Fe Cu rror~,XL5 koeV 0 cri ID=Vgrt=755S 
%Viiaw0o .C.35 -4G.95= 3.3476c nt I 6 0 Figure 17. SEM Photograph and EDS analysis of fibrous secondary deposits present in an air void of Core 124. The photograph is the same area shown in Figure 16. The high sulfur and aluminum content indicates that the material is ettringite (Ca 6 AI2(SO4)3 (OH)1 2 026(H 2 0).A-17 0*1F 0~0 140 120 100 80 60 40 20--0 0.2 0.3 0.4 0.5 0.6 0.7 0.8 Water-Cement Ratio 0 Figure 18. Relationship between permeability and water-cement ratio of mature portland cement paste (reference 1).A-i8 APPENDIX B Summary of Boron Results Ilyll~~~~ 
~ ~ 1111 Sj &#xfd;1 y l1 11j~ l1111 Core 122 -Boron Sample Locations Sample Distance below top Boron concentration, ID surface of channel ppm 122-1-1 11/16" 4370 122-1-2 1-1/16" 593 122-1-3 1-9/16" 1410 B-I Summary of Boron Results Core 123 -Boron Sample Locations Sample Distance below top Boron concentration, ID surface of channel ppm 123-1-1 1/4" 806 123-1-2 1/2" 285 123-1-3 13/16" 387 123-1-4 1-1/16" 549 123-2-1 2-1/2" 64 123-baseline 
-14" 36 0 B-2 Summary of Boron Results Core 124 -Boron Sample Locations Sample Distance below top Boron concentration, ID surface of channel ppm 124-1-1 1/4" 3670 124-1-2 1/2" 1000 124-1-3 11/16" 109 124-1-4 1" 136 0 B-3 W B Statistical Analysis of Rebar Yield and Tensile Strength Tests This appendix contains the following MPR Calculation.
.MPR Calculation 108-275-02, "Statistical Analysis of Rebar Yield & Tensile Strengths for Salem Nuclear Generating Station," Revision 0.0 MPR-2613 Revision 3 B-I 
.9 MPR MPR Associates, Inc.320 King Street Alexandria, VA 22314 CALCULATION TITLE PAGE Client: PSEG Nuclear Page 1 of 8+ Appendices Project: Task No.Salem Spent Fuel Pool Leakage 108-0303-275 Title: Calculation No.Statistical Analysis of Rebar Yield & Tensile Strengths for Salem Nuclear Generating Station 108-275-02 Preparer / Date Checker / Date Reviewer & Approver / Date Rev. No.Lisa Lichtenauer Michelle Heinz Robert Keating QUALITY ASSURANCE DOCUMENT This document has been prepared, checked, and reviewed/approved in accordance with the Quality Assurance rcquircmcnts of IOCFR50 Appendix B, as specified in the MPR Quality Assurance Manual.0 MPR-QA Form QA-3.1-1, Rev. 0 
@1 UMPR MPR Associates, Inc.320 King Street Alexandria, VA 22314 RECORD OF REVISIONS Calculation No. Prepared By Checked By Page: 2 108-275-02 Pags esritio Revision I Affected Pages Description 0 All Initial Issue I I I Note: The revision number found on each individual page of the calculation carries the revision level of the calculation in effect at the time that page was last revised.a MPR OA Form OA-3.1-2, Rev. 0 MPR Associates, Inc.*MPR 320 King Street.. M Alexandria, VA 22314 Calculation No. Prepared By Checked By Page: 3 108-275-02 Revision:
0 Table of Contents Table of Contents .....................................................................................................
3 1.0 Purpose ............................................................................................................
4 2.0 Summary of Results ......................................................................................
4 3.0 Calculation 
.....................................................................................................
5 4.0 References 
........................................................................................................
8 A Appendix A: Rebar Strength Data for Salem Nuclear Generating Station .. A-I B Appendix B: PSEG Nuclear Records of Reinforcing Steel Test Results ..... B-1 S I MPR OA Form: QA.3.1-3, Rev. 0
* UMPR: MPR Associates, Inc.320 King Street Alexandria, VA 22314 Calculation No. Prepared By Checked By Page: 4 108-275-02 Y--AAa A-A& Revision:
0 1.0 PURPOSE The purpose of this calculation is to document the statistical analysis of the yield and tensile strengths of the rebar used in the construction of structures at Salem Nuclear Generating Station.The means, standard deviations, and percentages of rebar above certain yield and tensile strengths are calculated for the rebar used in the walls of the structures.
2.0 

==SUMMARY==
OF RESULTS Table 2-1 summarizes the means, standard deviations, and sample sizes of the yield strength of the rebar in the structures at Salem Nuclear Generating Station. The percentages of rebar specimens that are stronger than a given yield strength are also summarized in the table below. For example, ninety percent (90%) of the rebar specimens have a yield strength of 62,200 psi or greater.Table 2-1 Rebar Yield Strength Analysis Rebar Mean Yield Yield Strength Sample 80% Lower 85% Lower 90% Lower 95% Lower Type Strength Std. Dev Size Bound Bound Bound Bound Total 69,840 6,370 394 64,100 63,300 62,200 61,300 No. 6 67,092 5,027 13 64,100 63,850 62,550 62,550 No. 8 69,410 7,376 123 63,500 63,000 61,500 60,750 No. 9 70,815 6,591 95 63,800 63,000 62,500 61,600 No. 10 70,934 5,024 47 66,150 65,400 62,200 62,100 No. 11 69,363 5,490 116 64,950 163,300 62,065 60,900 Table 2-2 summarizes the means, standard deviations, and sample sizes of the tensile strength of the rebar in the structures at Salem Nuclear Generating Station. The percentages of rebar specimens that are stronger than a given tensile strength are also summarized in the table below.For example, ninety percent (90%) of the rebar specimens have a tensile strength of 99,050 psi or greater.MPR QA Form: OA-3.1-3, Rev. 0 MVPR Associates, Inc.320 King Street Alexandria, VA 22314 Calculation No. Prepared By Checked By Page: 5 108-275-02 AA.~ Revision:
0 Table 2-2 Rebar Tensile Strength Analysis Rebar Mean Tensile Tensile Strength Sample 80% Lower 85% Lower 90% Lower 95% Lower Type Strength Std. Dev Size Bound Bound Bound Bound Total 105,850 5,116 394 101,600 100,300 99,050 97,400 No. 6 102,681 2,786 13 100,000 100,000 99,500 99,500 No. 8 105,751 5,164 123 101,300 100,000 99,700 98,000 No. 9 108,531 3,392 95 105,875 105,550 104,000 102,550 No. 10 107,336 4,773 47 104,350 103,600 102,400 98,800 No. 11 103,512 5,323 116 99,000 97,750 96,750 96,100 3.0 CALCULATION The data used in this analysis, summarized in Appendix A, was obtained from PSEG Nuclear records (Reference 
: 1) documenting chemical and physical tests of the reinforced bars at Salem.The tests were performed during the original plant construction.
Copies of the original records are provided in Appendix B. The data provided by PSEG Nuclear for this analysis is a sample of test records for the sizes and grades of rebar used in the SFP Building (Grade 60 of Size Nos. 6, 8, 9, 10, and 11 -see MPR Calc in our previous report). The method of including the test data in the analysis is listed below.1. In cases where the same test results were listed more than once within the records, only one entry is used in the statistical analysis of this calculation.
: 2. In some cases a tested sample resulted in a yield strength lower than the minimum required 60,000 psi.* In this case, if the retests of the heat resulted in values that satisfied the requirement, the retest data was included in the analysis, but not the failing original test value. The first failing test value was assumed to be invalid, but a passing original test was assumed acceptable.
* If however, the retest yield strengths still did not satisfy the minimum requirement, the entire set of data for that rebar heat was not included in the analysis.
Any rebar that did not have a yield strength of at least 60,000 psi was assumed to have not been used in the construction of the Salem Nuclear Generating Station, where Grade 60 was specified.
* In the cases where there were multiple retests of a given sample, the values were averaged in such a way to maintain no more than three (3) sets of data per heat.This was done in an effort to prevent certain heats from being more heavily weighted in the averages solely because more tests were performed.
MPR QA Form: QA-3.1-3, Rev. 0 MPR Associates, Inc..1 UM PIR 320 King Street Alexandria, VA 22314 Calculation No. Prepared By Checked By Page: 6 108-275-02 Revision:
0* The heats that were averaged, and their method of averaging, are noted in Appendix A with a footnote next to the heat number. For an even number of data values in a heat, the data was averaged in sets of two (2) or three (3) data values in order to obtain no more than 3 values per heat. If a heat included an odd number of data values, one value was maintained, while the others were averaged in groups of two (2)." If a yield strength value of less than 60,000 psi was obtained for a heat, and no retests were performed, the entire heat was discarded from this analysis.3. Note also that the tensile strength values were treated in the same manner as the yield strength data. That is, where the yield strength heats were averaged, the corresponding tensile strengths were averaged in the same way. If a set of data or an entire heat were discarded, the corresponding tensile strengths were also discarded.
The mean yield and tensile strengths were calculated for all samples, as well as for each different size rebar. The mean yield and tensile strengths were calculated using the following equation:-*--(1)n Where: X = Mean x = yield or tensile strength value n = total number of values The standard deviation of the samples was calculated using the equation below, where n and x are the same variables as mentioned above, and a is the standard deviation.
InEx' -(IxV (2)V n(n -1)Histograms of the yield and tensile strengths are presented in Figures 3-1 and 3-2 below in order to provide a visual representation of the analyzed data. Note that for the yield strength histogram, the distribution is a truncated distribution.
This is due to the fact that, as stated above, rebar that had a yield strength less than 60,000 psi was assumed not included in the construction of the plant because it did not meet minimum specifications.
MPR QA Form: QA-3.1-3, Rev. 0 
.1 JSMPR'-1~~MPR Associates, Inc.320 King Street , -'Alexandja, VA 22314~1:2)* Calculation No.- ..:Prepared By Checked By Page: 7-108-275 bevision:
0 S. .. .-..: I _ _ _ __"_ _.... ..U-tS" ItI 60-95%90% 88% 80%401 30 20- <10-5; '%' <R4Nb11 s 5 Q Al` Al AqA %'Yield Strength (psi)Figure 3-1. Yield Strength Distribution 60-95% 90% 85% 80%so 40 &#xfd;U.~30 4 U._, 20-20 Tensile Strength (psl)Figure 3-2. Tensile Strength Distribution The four vertical lines on the two figures above demark the values for which a given percentage of the data has at least that yield or tensile strength.
These values are given in Tables 2-1 and 2-2.Given the yield strength distribution, 85% of the rebar specimens had yield strengths equal to or greater than 63,300 psi, and 90% equal to or above 62,200 psi. Looking at Figure 3-2, 85% of A MPR QA Form: QA-3.1-3, Rev. 0 MPR Associates, Inc.30 King Street,1wo:JV )T...exandria, VA 22314 Calci-ation No. Prepared By Checked By Page: 8..-108-275-02 7 Revision:
0 the rebar specimens result in a tensile strength equal to or greater than 100,300 psi, and 90%equal to or above 99,050 psi.

==4.0 REFERENCES==
: 1. PSEG Nuclear records documenting reinforcing bar tests for Salem Nuclear Generating Station, performed during original plant construction (Provided in Appendix B).(Documents provided to MPR by express package from K. Fisher (PSEG Nuclear) to J.Simons (MPR) dated May 23, 2003).S MPR QA Form: OA-3.1-3, Rev. 0 0 U MPR MPR Associates, Inc.320 King Street Alexandria, VA 22314 F Calculation No. Prepared By Checked By Page: A-1 108-275-02 Revision:
0 A Appendix A: Rebar Strength Data for Salem Nuclear Generating Station MPR GA Form: QA-3.1-3, Rv. 0 Table A-1 Rebar Yieid Strength Data Yield Tensile Company Record Heat Bar Size Strength Strength PSEG 1825 64821a 8 66,850 108,750 PSEG 1825 64821a 8 65,850 108,100 Milton 24 134682 9 66,350 107,650 Milton 24 135386 9 69,400 108,650 PSEG 65 135386 9 65,800 107,400 Milton 24 135849 9 66,150 109,100 PSEG 64 135849 9 66,300 108,800 Milton 1678 135911 8 74,700 97,400 Milton 1678 135911 8 72,750 97,400 Milton 745 135984 11 64,750 107,050 Milton 746 135984 11 73,850 110,650 Milton 746 135984 11 75,500 110,650 PSEG 1089 136068 8 63,000 102,500 PSEG 1089 136068 8 60,800 102,300 PSEG 71 136069 9 63,000 111,800 PSEG 71 136069 9 80,600 112,000 PSEG 72 136099 9 62,000 104,000 PSEG 72 136099 9 61,600 104,600 PSEG 1088 136111 8 62,800 103,000 PSEG 1088 136111 8 61,500 103,000 PSEG 314 136155 11 78,900 117,000 PSEG 314 136155 11 82,000 114,100 Milton 24 136156 9 68,200 107,550 PSEG 60 136156 9 68,500 107,550 PSEG 60 136156 9 66,700 106,800 PSEG 303 136173 11 61,000 104,400 PSEG 303 136173 11 61,400 104,100 PSEG 302 136182 11 65,400 109,000 PSEG 302 136182 11 66,900 109,900 PSEG 70 136388 9 62,000 104,400 PSEG 70 136388 9 62,600 105,300 Milton 782 136663 11 61,688 101,948 Milton 781 136663 11 61,364 101,623 Milton 781 136663 11 67,834 109,873 Milton 513 136787 10 66,150 107,250 a: average of 2 data from same heat b: oriinal data in a heat where other data was averaged c: yield strength data <60kslin this heat not included A-2 MPR Calculation 108-275-02 Appendix A d: average of 3 data from same heat 1 i Table A-i Rebar Yi=.cl Strength Data Company Milton Milton Milton Milton Milton Milton Milton Milton PSEG PSEG Milton Milton Milton Milton Milton PSEG PSEG Milton Milton Milton Milton Milton Milton Milton Milton Milton Milton Milton Milton Milton PSEG Milton Milton PSEG PSEG Record 515 515 764 780 780 514 515 515 296 296 764 1642 515 515 20 46 46 746 745 746 10 10 10 10 13 10 10 13 10 12 28 10 10 150 150 Heat 136787 136787 136907 136907 136907 136958 136958 136958 137010 137010 137010 137023 137023 137023 137102 137102 137102 137248 137248 137248 137801 137801 137951 137951 137951 137958 137958 137958 138010 138010 138010 138011 138011 138333 138333 Bar Size 10 10 11 11 11 10 10 10 11 11 11 10 10 10 9 9 9 11 11 11 9 9 9 9 9 9 9 9 9 9 9 9 9 11 11 Yield Strength 73,000 70,000 72,900 76,751 78,662 76,400 82,250 83,750 71,794 67,948 65,050 62,200 72,750 72,750 69,400 63,800 64,000 71,400 66,650 66,100 72,700 74,000 76,000 77,800 72,200 80,600 83,650 74,500 76,000 65,150 87,250 74,750 72,700 65,800 68,750 Tensile Strength 105,650 106,000 115,150 112,420 112,738 115,850 116,150 116,650 108,012 108,653 109,950 104,900 104,450 104,050 109,700 108,300 108,200 99,350 102,650 99,650 108,250 107,150 110,200 108,600 109,100 113,250 113,250 113,250 102,550 102,550 102,550 105,650 106,200 100,300 100,300 a: average of2 data from same heat b: original data in a heat where other data was averaged c: yield svrength data <60ksiin this heat not included d: average of 3 data from same heat A-3 MPR Calculation 108-275-02 Appendix A Table A-1 Rebar Yield Strength Data 0 Company PSEG PSEG PSEG PSEG Milton Milton Milton Milton Milton Milton PSEG PSEG PSEG PSEG PSEG PSEG PSEG PSEG PSEG PSEG PSEG PSEG Milton Milton Milton Milton Milton Milton Milton Milton Milton Milton Milton Milton Milton Record 148 148 136 136 6 6 1700 1700 1718 1718 59 59 56 56 58 58 61 61 65 65 16 16 1785 1785 1186 1186 1186 1186 1805 1805 1805 1805 1805 1805 1242 Heat 138344 138344 138434 138434 138566 138566 138624 138624 138958 138958 138989 138989 138990 138990 139000 139000 139001 139001 139039 139039 139639 139639 139736 139736 139900 139900 140843 140843 140909 140909 140913 140913 140935 140935 140935 Bar Size 11 i1 11 11 9 9 8 8 8 8 11 11 11 11 11 11 11 11 11 11 11 11 9 9 8 8 8 8 8 8 8 8 8 8 8 Yield Strength 82,250 82,050 65,050 63,600 71,650 70,100 63,000 66,250 74,050 75,950 74,500 76,300 69,200 70,450 75,150 76,450 82,800 82,450 71,350 73,250 67,950 68,900 71,700 70,700 63,300 60,750 70,150 70,150 72,400 71,400 68,600 68,600 71,500 71,200 63,000 Tensile Strength 111,500 111,100 94,450 95,700 111,850 111,850 101,300 101,950 107,800 108,450 104,900 103,900 99,050 99,050 102,300 102,950 102,250 103,250 97,750 98,400 100,300 100,950 109,100 109,100 100,650 100,650 104,550 104,550 106,400 107,800 104,500 104,500 106,300 107,700 103,250 a: average of 2 data from same heat b: oiginal data in a heat where other data was averaged c: yield strength data <60ksi ki this heat not included d: average of 3 data from same heat A-4 MPR Calculation 108-275-02 Appendix A It~,~tJ'Table A-1 Rebar Yield Strength Data;4&#xfd;'Company Milton Milton Milton Milton Milton Milton Milton Milton Milton Milton Milton Milton Milton Milton Milton Milton Milton Milton Milton Milton Milton Milton Milton Milton PSEG PSEG PSEG PSEG Milton Milton Milton Milton Milton Milton Milton Record 1192 1192 1192 1192 1785 1785 1793 1793 1219 1219 1219 1219 1219 1219 1477 1477 1815 1815 1815 1815 1819 1819 1819 1819 1875 1875 1877 1877 1051 1051 1921 1921 1921 1921 1914 Heat 141392 141392 141470 141470 141473 141473 141515 141515 141842 141842 141843 141843 141865 141865 141898 141898 141901 141901 141902 141902 141953 141953 141954 141954 142548 142548 143440 143440 143668 143668 144032 144041 144042 144043 144076 Bar Size 6 6 8 8 9 9 10 10 11 11 11 11 6 6 10 10 9 9 9 9 8 8 8 8 8 8 8 8 6 6 9 9 9 9 9 Yield Strength 64,900 62,550 72,750 76,600 73,200 73,200 71,750 73,150 67,500 65,300 66,000 65,100 64,100 64,300 72,000 70,400 70,700 70,900 71,950 70,100 77,250 71,450 66,250 64,950 65,150 66,450 64,450 63,650 73,000 80,450 73,750 79,000 67,350 65,300 69,700 Tensile Strength 100,250 100,000 106,500 106,500 105,650 106,200 105,250 106,100 97,400 97,400 96,800 96,500 99,500 100,000 108,000 108,000 105,550 106,100 108,150 108,750 107,150 107,150 101,950 101,300 103,300 103,300 103,950 103,250 106,050 106,050 115,650 116,850 107,150 107,650 106,050* ?~'*'*i'.~'l I j.r 4~a: average of 2 data from same heat b: original data ii a heat where other data was averaged c: yield strength data <60ksi in this heat not included d: average of 3 data from same heat A-5 MPR Calculation 108-275-02 Appendix A Table A-I Rebar Yield Strength Data Company Milton Milton PSEG PSEG Milton Milton Milton Milton PSEG PSEG Milton Milton Milton Milton Milton Milton Milton Milton Milton PSEG Record 1914 1914 1915 1915 1916 1914 1914 1921 1920 1920 1923 1916 1919 1064 1570 1570 1678 1678 24 61 Heat 144597 144597 144663 144663 144663 144691 144691 144691 144692 144692 144692 144708 144724 144854 146213 148742 235532 235532 235656 235656 Bar Size 6 6 8 8 8 8 8 8 8 8 8 8 8 8 11 11 8 8 9 9 Yield Strength 64,650 64,400 63,500 64,000 70,100 64,450 65,800 63,150 62,000 61,300 63,150 66,250 66,250 67,100 71,850 63,450 71,450 75,350 77,150 65,500 Tensile Strength 103,250 102,800 105,600 106,600 109,300 103,300 103,950 103,950 100,000 99,500 104,600 105,850 105,200 107,250 103,300 102,550 107,150 107,800 109,250 107,200 108,165 108,735 103,415 103,225 104,150 105,000 104,900 104,850 103,350 115,200 104,400 104,000 109,600 106,500 14..4 PSEG 1084, 1631 235722a PSEG PSEG PSEG PSEG PSEG PSEG PSEG PSEG PSEG PSEG Milton PSEG Milton 1084, 1631 235722a 311 235766Wc 311 311 300 300 1021 1021 54 54 24 63 24 235766:c 235766"'235783 235783 235985a 235985a 236029 236029 236033 236033 236137 a: average of 2 data fro 8 65,760 8 67,280 11 61,470 11 62,065 11 60,880 11 63,200 11 60,500 10 62,100 10 62,100 9 77,000 9 74,460 9 62,500 9 70,800 9 62,500*m same heat I.:...b: original data in a heat where other data was averaged c: yiel strength data <60ksi In this heat not Included d: average of 3 data from same heat A-6 MPR Calculation 108-275-02 Appendix A Table A-1 Rebar Yield Strength Data Yield Tensile Company Record Heat Bar Size Strength Strength PSEG 59 236137 9 66,000 108,000 Milton 24 236180 9 64,800 109,200 PSEG 62 236180 9 63,000 105,800.Milton 22 236429a.c 9 66,239 106,601 Milton 22 236429"c 9 69.445 106,060 Milton 23 2 3 6 4 2 9 b'c 9 76,531 107,150 Milton 23 236431 9 69,697 105,555 Milton 21 236431 9 62.121 106,060 Milton 21 236436 9 63,265 106,133 Milton 23 236436 9 64,796 106,632 Milton 519 236458a. 10 71,143 107,521 Milton 518, 519 236458ac 10 70,326 107,317 Milton 519 236458b'c 10 70,732 107,724 Milton 514 236501 10 65,850 109,900 Milton 515 236501 10 67,850 108,350 Milton 515 236501 10 75,400 107,950 PSEG 43 236636"-c 9 61,450 103,950 PSEG 43 236636&- 9 61,250 103,200 PSEG 20, 43 236636ac 9 63,325 105,875 PSEG 249 236776 11 73,550 100,650 PSEG 249 236776 11 76,150 101,000 Milton 1363 236776 11 66,650 101,650 PSEG 247 236778 11 65,900 96,450 PSEG 247 236778 11 64,950 96,100 Milton 1363 236778 11 65,250 98,700 Milton 1672 236844 8 62,800 107,050 PSEG 1056 236884 8 73,700 108,350 PSEG 1056 236884 8 73,400 106,950 Milton 16 237124 9 65,300 109,700 PSEG 42 237124 9 80,800 109,100 PSEG 42 237124 9 71,950 109,050 Milton 10 237569 10 79,100 111,750 Milton 12 237569 9 76,000 113,250 a: average of 2 data from same heat b: original data k; a heat where other data was averaged c: yield strength data <60ksl in this heat not included A-7 MPR Calculation 108-275-02 Appendix A d: average of 3 data from same heat Table A-1 Rebar Yieia Strength Data Yield Tensile Company Record Heat Bar Size Strength Strength Milton 10 237569 9 84,200 112,250 Milton 10, 12 237588a 9 75,000 108,950 PSEG 31 237588a 9 77,800 109,700 Milton 1682 237615 8 68,850 103,900 Milton 1682 237615 8 68,850 103,900 Milton 13 237640 9 72,950 112,250 Milton 10 237640 9 77,050 110,700 PSEG 32 237640 9 81,100 111,200 PSEG 147 237722 11 72,200 102,650 PSEG 147 237722 11 73,500 102,650 PSEG 132 237839 11 70,500 102,500 PSEG 132 237839 11 70,950 102,600 Milton 1718 238100 8 91,150 112,200 Milton 1718 238100 8 86,550 112,800 Milton 504 238190 10 73,400 112,100 Milton 504 238190 10 73,600 112,600 Milton 504 238202 10 77,400 118,550 Milton 504 238202 10 79,600 118,000 PSEG 64 238283 11 70,700 98,700 PSEG 64 238283 11 70,300 98,400 PSEG 63 238288 11 71,650 102,550 PSEG 63 238288 11 72,300 101,900 PSEG 57 238291 11 70,250 99,000 PSEG 57 238291 11 75.150 99,000 PSEG 60 238309 11 70,600 97,050 PSEG 60 238309 11 71,100 96,750 Milton 1785 238721 11 68,500 101,600 Milton 1785 238721 11 71,450 101,600 PSEG 18 238745 11 68,050 101,600 PSEG 18 238745 11 67,650 101,300 PSEG 19 238916 11 67,400 100.300 PSEG 19 238916 11 68,200 101,000 Milton 1186 240118 8 64,300 105,200 Milton 1186 240118 8 67,550 105,200 Milton 1793 240235 9 73,500 108,500 a: average of 2 data from same heat b: orignal data in a heat where other data was averaged c: yield strength data <6Oksi in this heat not inckided A-8 MPR Calculation 108-275-02 Appendix A d: average of 3 data from same heat Table A-1 Rebar Yield Strength Data Yield Tensile Company Record Heat Bar Size Strength Strength Milton 1793 240235 9 74,750 110.100 Milton 1785 240684 9 85,600 117,000 Milton 1785 240684 9 81,450 117,000 Milton 1785 240714 8 74,350 107,700 Milton 1785 240714 8 74,700 109,100 Milton 1192 240731 8 81,800 103,900 PSEG 1783 240731 8 86,200 105,250 Milton 1793 240734 9 78,800 114,650 Milton 1793 240734 9 83,350 115,150 Milton 1785 240736 8 86,200 105,250 Milton 1785 240760 10 70,750 105,300 Milton 1785 240760 10 73,600 105,300 Milton 862 240960 8 65,400 101,300 Milton 862 240960 8 64,100 101,900 Milton 1815 241147 8 76,650 110,400 Milton 1815 241147 8 77,900 110,400 Milton 1477 241204 10 71,100 113,800 Milton 1477 241204 10 76,400 113,800 Milton 1815 241213 9 68,900 106,650 Milton 1815 241213 9 67,850 106,100 Milton 1823 241236 8 68,400 95,400 Milton 1823 241236 8 65,100 95,400 Milton 1823 241238 8 75,300 112,300 Milton 1823 241238 8 77,300 110,400 PSEG 1820 241245 8 67,100 98,000 Milton 1823 241245 8 67,100 98,000 Milton 1819 241250 8 75,000 107,900 Milton 1819 241250 8 69,750 107,900 PSEG 1874 241968 8 76,300 111,850 PSEG 1874 241968 8 75,000 112,500 Milton 1911 243627 8 70,500 105,750 Milton 1911 243627 8 66,650 105,750 PSEG 1917 243908 8 61,000 99,700 PSEG 1917 243908 8 61,000 99,000 Milton 1919 243908 8 70,150 110,400 a: average of 2 data from same heat b: original data in a heat where other data was averaged A-9 c: yied strength data <60ksl in this heat not included MPR Calculation 108-275-02 Appendix A d: average of 3 data from same heat Table A-I Rebar Yield Strength Data Yield Tensile Company Record Heat Bar Size Strength Strength PSEG 1922 243912 8 60,500 97,700 PSEG 1922 243912 8 60,500 99,000 Milton 1923 243912 8 65,800 107,250 Milton 1923 243942 8 65,400 100,000 Milton 1064 244136 8 61,050 99,350 Milton 1570 247569 11 73,050 106,600 Milton 1570 247720 11 77,300 117,100 Milton 1570 247723 11 63,300 104,200 Milton 23 321698 9 67,857 107,653 Milton 21 321698 9 63,776 107,653 Milton 509 321737 10 62,000 103,600 Milton 509 321737 10 64,500 105,650 PSEG 312 321880 11 81,400 120,000 PSEG 312 321880 11 76,700 116,700 PSEG 301 321934 11 62,200 104,500 PSEG 301 321934 11 60,900 105,000 PSEG 305 322047 11 62,200 105,600 PSEG 305 322047 11 60,800 105,600 PSEG 1019 322064a 10 64,600 105,800 PSEG 1019 322064a 10 65,400 104,450 PSEG 55 322183 9 62.600 107,600 PSEG 55 322183 9 63,000 108,200 PSEG 307 322207 11 60,500 105,900 PSEG 307 322207 11 62,800 106,500 PSEG 56 322309 9 68,200 108,600 PSEG 56 322309 9 68,000 109,400 PSEG 50 322310 9 63,200 103,200 PSEG 50 322310 9 60,400 102,400 Milton 1678 322425 6 70,000 101,150 Milton 1678 322425 6 69,550 101,600 Milton 1634 322692 8 67,550 108,450 Milton 1637 322692 8 65,584 105,194 Milton 1637 322692 8 66,233 105,844 Milton 1637 322705 8 68,831 110,389 Milton 1637 322705 8 69,480 111,038 a: average of 2 data from same heat b: original data in a heat where other data was averaged c: yield strength data <60ksi k7 this heat not included A1 0 MPR Calculation 108-275-02 Appendix A d: average of 3 data from same heat ii'1,~~v *~2~ I''P.~1J1 p'If~ ~ii Table A-1 Rebar Yield Strength Data Company PSEG Milton Milton PSEG PSEG Milton PSEG PSEG Milton Milton Milton Milton Milton Milton Milton Milton Milton Milton Milton Milton PSEG PSEG Milton PSEG PSEG Milton Milton Milton PSEG PSEG PSEG PSEG PSEG PSEG Record 291 780 764 299 299 1642 250 250 764 513 515 515 513 515 515 509 509 513 515 515 248 248 1363 1057 1057 1672 242, 746 745, 746 33 33 133 133 149 149 Heat 322764 322764 322764 322803 322803 322803 322820a 322820a 322820b 322836 322836 322836 322837 322837 322837 322844 322844 322845 322845 322845 323068 323068 323068 323151 323151 323151 323632a 323632a 323827 323827 324227 324227 324235 324235 Bar Size 11 11 11 11 11 11 11 11 11 10 10 10 10 10 10 10 10 10 10 10 11 11 11 8 8 8 11 11 9 9 11 11 11 11 Yield Strength 68,831 69,300 68,850 70,192 69,871 69,350 70,266 62,180 63,450 72,200 72,800 73,200 70,950 70,800 70,800 67,050 66,950 66,550 66,550 69,350 79,600 75,650 68,550 79,500 73,400 63,450 71 300 72,275 74,000 82,150 73,700 74,350 68,500 66,150 Tensile Strength 105,844 100,980 109,100 108,012 108,012 111,600 107,235 107,500 108,950 104,350 105,200 105,200 105,650 104,400 104,400 102,850 102,400 99,200 98,000 98,800 103,600 103,600 103,300 104,500 103,800 102,550 105,810 107,100 110,700 111,750 105,900 106,250 96,100 96,100 ri IT~4'J~j ~, .*:"~4 a: average of 2 data from same heat b: onighial data ki a heat where other data was averaged c: yield strength data <60ksi in this heat not included d: average of 3 data from same heat A-11 MPR Calculation 108-275-02 Appendix A I.Table A-1 Rebar Yield Strength Data A'.Company PSEG PSEG PSEG PSEG PSEG PSEG PSEG PSEG PSEG PSEG Milton Milton Milton Milton Milton Milton Milton Milton Milton Milton Milton Milton Milton* .,Milton Milton Milton YP Milton PSEG PSEG PSEG MtPSEG Milton Milton Milton Milton Record 134 134 137 137 135 135 20 20 17 17 1242 1186 1186 1192 1192 1186 1186 1219 1219 1793 1793 1793 1793 981 981 1510 1873 1876 1876 1918 1918 1919 1916 1923 1570 Heat 324320 324320 324324 324324 324325 324325 325500 325500 325501 325501 325780 326413 326413 326580 326580 326742 326742 326810 326810 327609 327609 327610 327610 327677 327677 327885 327885 329345 329345 330619 330619 330619 330646 330660 332081 Bar Size 11 11 11 11 11 11 11 11 11 11 6 6 6 8 8 8 8 10 10 8 8 8 8 9 9 8 8 8 8 8 8 8 8 8 11 Yield Stredgth 67, 00 67,t00 64,700 65,150 65,450 65.900 67,400 67,300 67,650 67,750 66,350 64,100 63,850 77,250 76,600 63,900 64,750 72,000 71,200 88,450 92,300 90,250 94,800 77,600 75,500 65,600 65,600 63,800 66,450 60,000 60,100 64,1 0 61,520 63,800 75,000 Tensile Strength 99,650 99,350 95,750 96,700 96,400 97,350 99,350 99,050 98,400 99,050 104,200 101,600 108,400 109,100 109,100 101,900 103,850 109,200 109,200 122,450 122,450 118,200 118,200 112,800 112,800 107,150 107,150 100,000 101,300 100,000 100.300 100,000 100,000 103,300 109,200 J: '1 a: average of 2 data from same heat b: origi&#xfd;al data in a heat where other data was averaged c: yield strength data <60ksi i this heat not included d: average of 3 data from same heat A-12 MPR Calculation 108-275-02 Appendix A Table A-1 Rebar Yield Strength Data: 1 Company Milton Mitton US Steel US Steel PSEG PSEG US Steel PSEG PSEG Bethlehem Bethlehem PSEG PSEG Record 1570 755 1889 1849 1848 1848 1826 1850 1850 1852 1852 1851 1851 Heat 334582 336776 03M585 04M326b 0 4 M 3 2 6d 04M 3 2 6d 06L821 209C680a 209C6802 209C680b 2 0 9 C 6 8 1 b 209C681a 209C681 a Bar Size 11 8 8 8 8 8 8 8 8 8 8 8 Yield Strength 71,300 66,650 60,510 69,620 65,833 65,533 80,510 69,000 69,500 76,200 66,800 66,450 66.400 Tensile Strength 105,150 101,650 102,780 107,850 104,467 104,733 118,730 113,200 113,300 122,500 107,800 108,650 108,800 a: average of 2 data from same heat b: original data in a heat where other data was averaged c: yield strength data <60ksi in this heat not included d: average of 3 data from same heat A-13 MPR Calculation 108-275-02 Appendix A 
.MPR Associates, Inc.*MPR 320 King Street SM M... .Alexandria, VA 22314 Calculation No. Prepared By Checked By Page: B-I 108-275-02 
,- , Revision:
0 B Appendix B: PSEG Nuclear Records of Reinforcing Steel Test Results PSEG Nuclear records documenting reinforcing bar tests for Salem Nuclear Generating Station, performed during original plant construction (Provided in Appendix B). (Documents provided to MPR by express package from K. Fisher (PSEG Nuclear) to J. Simons (MPR) dated May 23, 2003).MPR QA Form: QA-3.1-3.
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K.-'.7 t'o PUBLIC SERVICE ELECTRIC AND GAS COMPANY'.4 TE =VUCJ-DKPA7T" VR T TESTING LABORATORY REPORT-701 REIKFORCEMENr BAR TEST R___ CT CLJ UORDER NO. [DATE IDENTIFICAT(ON NO. MATERIAL J HEAT NO. SIZ --I1A-RA ./_1,: ,- / __ rz. /11',&#xfd;2I ~ j gZ9~/3 e? .;l 11'~/DATle T E N S I ON T E S T R E 5 U L T S YIELD -NOTE iI..uLTIMATE 
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PUBLIC SERVICE ELECTRIC AND GAS COMPANY TESTING LABORATORY REPORTW 0 0 REIXFORCEMENT BAR TEST PROJEC SjL U ~-CLEAB U.N. STA. JOORNO. DATE ..(, IDNTFCAIN o MATE IAL _T_ ttQ,___ &#xfd;si-4 --AREA -,43___ I -ra ___DATE IDENT. NO.T ENS IO N YIELD (NOTE 1)TIE S T RESULTS IUL- TLI MAT iE I, 3~.LBS. PSIf7 i1,-L S.85 -PSI~'C /ELONGATION IN_B__INCNLS. 
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.7 0 PUBI.IC SERVICE ELECTRIC AND GAS COMPANY Testing Laboratory REINFORCEMENT BAR TES.T REPORT-5&#xfd;10011 jPT.OJECT S ~ TA.kDOER NO.InFNTIPICATION NO. MATERIAL HEAT NO. SIZE AprA-) Z'#i 216 C2!2 9~ '/ -2.23 6.J 2 .29 el T FN 1,1O0N TEFS T R E&#xfd;S kLT 5 YIE:LD INOIE I UI.7 IMAIE. ELONCATION IN DATE I ENT. NO. IS.PSI LObS PSI ft IN(M(&#xb6;. %_____ 74'&#xfd; :!J'c Moe) 4' /C)4'4p-4 104,4 NorFJ I. CrIl6!0Fl OVFR a INCHES OR DROP OF BEAM TESTI D fly ~*S t.. ';&#xfd;j Cev. Jr., -t?%Sg 4 B ' 4M .J"' 'ABORATORY ENGINrrR" \MAATfP1ALS nPVlSl'AJ CH~IEF PUBLIC SERVICE ELECTRIC AND GAS COMPANY Testing Laboratory REINFORCEMENT BAR TEST REPORT 00"%0 U PROJECT f.AIjli 1';0.-.B Ci.... ETA, AOER NO.ICENTIFICATIUNj NO. MATERtAL NHAT NO. SIZE AREA 2 -...... A 163 32 A /,oo ."..TFNS tI N TEST RESULTS VIELP (VIOTE 1) UL T IMATF ELONGATION IN DATE IDENT. NO. 1L.1S. PSI LnS. PSI A tNCHES. %1A2//A -Q " /0 6, 20 8. / -7.-5 NOTE D. OIVIOER OVER 0 INCHES OR DROP OF REAM TESTED Y ._ r b. 9 c' ..... W  7' ' /1b -: '. .-A00RATORY ENGINEER,,..-
-F'- 'MATCRIALS DIVISION CHIET 
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.flEINFORCENIENT BAHi TEST REPOHL'.-.`
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..*REINFORI PROJECT C ~' ULP .SA : tbfmTlrICATION NO:. .4TRA 4..-~T.~~j, ((I I:V ~'.DATE A.iif A A IV ,_ .* .: .,_"_..7 -., ., .'4.;, ...-1 -.....T ' 1. 3 T.-AESULTI iI ULTIMATL i im 1 J w , i...,-..;.:,, .--. LOS.*ELONGATION INI a INCHES %PST I'~/5 ~Jf/V 2 2 -fel Ao/0* .4 ~ 4 I.4 4 4 mi4m. .... .-4 4 .~. ______________
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NO- .IDENTItFICATICN NO. MATERIAL HEAT NO. SIZE AREA)IJA- S'- --v, S , TENS I ON T E ST R E 5 U L T 5 YIELO (NOTE 1) ULTIMATE ELONGATION IN DATE IDENT. NO. LRS. PSi LBS. PSI 8 INCHES. %____
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R 13 ST4'FROJECr sLint NucL,.0 C~L, -'TL 0 La'1aS'IDENTIFICATION NO.' "MTERAL~ K 3iT& .'q I ARIEA-~~ ~~ 4 1:..;j3 A.~_ _ _ _ _ _ _ _ _ _ _ _~ /. 7 ~4_ _ _ _ _ _ _ _ _ _ 90.. -3 F_ _! 6 2 I 4....................
a I **T ENS I ON T E 3 T N f SU LT S"E T1 YIELD (NOTE I). ULTIMATE IELONGATION IN 0ATIE l IDIENT , W , LB S. IA PS I", ." LS .'. 1S, 6 IW0CES S 1370.39- 11/2,c0 o 4 // -5 I52a510*. a.t~aJ.~ .. I -I ~, ~ I-. aaari~r~:.a Z.~ r-~~.aa~-- -a,.-a--'MOTE 1. DIVIDER OVER a INCHES 09 DROP OP sI Tr~TrO BY Af A40Ar'.,.6C I -..ti..), &~4~)u,~~daVr.q7~S
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A HEAT NwO. SIZE AREA____ ___ ___ __ ~ z< ~ J /?J%~ _ ___ /00 T E N S I ON T ST R E S U L T S YIELD (NOTE 1) ULTIMATE ELONGATION IN DATE: " IDENT. NO. LBS. PSI LOS. PSI a INCHES, v ______ o_ e- 167, , 0  /O7,/01 NOTE* 1. DIVIDER OVER 6 INCHErS OZ-9*flZ-0r VtAm * -TESTEB ., ., ,. e, /..LABORATOR\'
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HEAT NO. I SI:E *13_______
63 K34p' 2 41 I Z/"-32! 1It I. I 1 .1 "__ _ 1 _ __ __7 E N S C' TN 7 Z R R E S U L TS Y:ELO. (;CdTZ I ULT71I L. ;IO .I. L2S. I PSI I L1S. zPSI I C Zb .I//7 _ _ _ _ __ _ _ _ __ __i e I ' -c3c"P. 'oC2_0__2______eleq.
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REPORT* Xo.-6.PROJECT Sm CE o OT*.- REINFOCENENT._ORET I IDENTIFICATION NO. " 
.AT N ..AR"A.32, 3 , ,, -/ .7 ..,.; '2 :rI e4':.... -""? 4I TEN S I aN TEST R ESULTS*YIELD (1407C 1)ULTIMATE ELONGATION IN ,nrNr. ~u, w L13 PS I NOTE I. DIVIDER OVER A I .CHES OR DOPO 6E;I4.5k.Irn y //bL roH L-,4!= 7 OiV ;6 &* ~ j~M TERIALS DIVI!7 ;'I CHIEF~LABORATORY ENGINEE .-.V,~t; ~.6/3~:~.10Si
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-C Margin Reduction from Rebar Corrosion This appendix contains the following MPR Calculation.
.MPR Calculation 0108-0275-35, "Salem Spent Fuel Pool Reinforcing Steel Load Capacity at Degraded Conditions," Revision 0.0 MPR-2613 Revision 3 C-1 MPR Associates, Inc.320 King Street Alexandria, VA 22314 CALCULATION TITLE PAGE Client: PSEG Nuclear Page I of 9 Project: Task No.Salem Spent Fuel Pool Leakage 0108-0303-0275 Title: Calculation No.Salem Spent Fuel Pool Reinforcing Steel Load Capacities at Degraded Conditions 0108-0275-35 Preparer / Date Checker / Date Reviewer & Approver / Date Rev. No.Michelle Heinz Lisa Lichtenauer Robert Keating 0 QUALITY ASSURANCE DOCUMENT This document has been prepared.
checked. and reviewed/approved in accordance with the Quality Assurance requirements of 10CFR50 Appendix 13. as specified in the MPR Quality Assurance Manual.MPR-QA Form QA-3.1-1.
Rev. 0 
*MPR MPR Associates, Inc.320 King Street Alexandria, VA 22314 RECORD OF REVISIONS Calculation No. Prepared By Checked By Page: 2 0108-0275-35 
.,4., '. " Z,- Revision Affected Pages Description 0 All Initial Issue-- I__ --__Note: The revision number found on each individual page of the calculation carries the revision Note: The revision number found on each individual page of the calculation carries the revision level of the calculation in effect at the time that page was last revised.MPH QA Form OA-3.1-2, Rev. 0 MPR Associates, Inc.U MPR 320 King Street W M P RAlexandria, VA 22314 Calculation No. Prepared By Checked By Page: 3 0108-0275-35 Revision:
0 Table of Contents Table of Contents............................................................................
3 1. 0 Purpose.................................................................................
4 2.0 Results .................................................................................
4 3.0 Calculation......................................
I..........................
I.............
5 4.0 References 
.............................................................................
9 MPR QA Form: QA-3.1-3, Rev. 0 MPR Associates, Inc.WM PR 320 King Street Alexandria, VA 22314 Calculation No. Prepared By Checked By Page: 4 0108-0275-35 )A.
V Revision:
0 1.0 PURPOSE The purpose of this calculation is to determine the percent reductions in the Salem spent fuel pool reinforcing steel load capacities due to various levels of reinforcing steel degradation.

===2.0 RESULTS===
Table 2-1 shows the percent reductions in the load capacities of the reinforcing steel sizes present in the spent fuel pool structure due to various levels of reinforcing steel degradation.
Table 2-1. Percent Reductions in Reinforcing Steel Load Capacity at Various Degradation Levels 0 Size #8 #9 #10 #11 Diameter (In.) 1.000 1 1.128 1.270 1.410 0.000 0.00 0.00 0.00 0.00 0.010 1.99 1.77 1.57 1.41 0.020 3.96 3.51 3.12 2.82 0.030 5.91 5.25 4.67 4.21 Degradation 0.040 7.84 6.97 6.20 5.59 (Inches) 0.050 9.75 8.67 7.72 6.97 0.060 11.64 10.36 9.23 8.33 0.070 13.51 12.03 10.72 9.68 0.080 15.36 13.68 12.20 11.03 The expressions relating the percent reduction in reinforcing steel load capacity and the reinforcing steel degradation level are provided below. In each equation, 'x' represents the reinforcing steel degradation level, and 'y' represents the percent reduction in load capacity.#8: y = 192x +0.0933#9: y = 171.02x + 0.0734#10: y = 152.52x + 0.0579#l 1: y = 137.82x + 0.0469 MPR QA Form: QA-3.1 3, Rev. 0 
*MPR MPR Associates, Inc.320 King Street Alexandria, VA 22314 Calculation No. Prepared By Checked By Page: 5 0108-0275-35 Revision:
0 3.0 CALCULATION The yield strength of reinforcing steel, fy, is defined as the force that the steel can Withstand at a certain cross-sectional area: fy = F/As (1)where: (2)4 d = reinforcing steel diameter F = reinforcing steel load capacity Substituting Equation (2) into Equation (1) and solving for the load capacity yields: F = f d 4 (3)The percent reduction in the original load capacity due to degradation of the reinforcing steel is calculated as:%Reduction=
1- Vith-Degradalion 
*100= 1- 100 Fo-Depradationd where: dd= reinforcing steel degradation level (4)Note that dd represents the degradation off the reinforcing steel diameter, and not the radius.The reinforcing steel sizes present in the Salem spent fuel pool structure are provided in References 1-9. This information is summarized in Table 3-1. Note that only the reinforcing steel around the wetted perimeter of the structure is included in the table (e.g., reinforcing steel in the portion of the North wall above the water is not included).
MPR QA Forni: OA-3.1-3, Rev. 0 MPR Associates, Inc.OIM M PR 320 King Street Alexandria, VA 22314 Calculation No. Prepared By Checked By Page: 6 0108-0275-35 
)'{A -'.&#xfd; V .,-&#xfd; Revision:
0 Table 3-1. Spent Fuel Pool Steel Reinforcement Sizes Horiz, Tension Horiz, Tension Vert, Tension Vert, Tension Wall Inside Outside Inside Outside North #8 #8 #8 #8 West #9 #8 #9 #8 South #11 #11 #11 #11 East #10 #8 #10 #8 N-S, Tens. Top N-S, Tens Bottom E-W, Tens. Top E-W, Tens. Bottom Slab #9 #8 #8 #8 Notes: 1. 'Inside' refers to the water side of the wall.from the water.'Outside' refers to the side of the wall remote The diameters of the reinforcing steel under consideration in this calculation are provided in Table 3-2. This information is taken from Reference 10.Table 3-2. Reinforcing Steel Diameters Size Diameter#8 1.000#9 1.128#10 1.270#11 1.410 MPR OA Form: QA-3.1-3, Rev. 0 MPR Associates, Inc.320 King Street UM PR Alexandria, VA 22314 Calculation No. Prepared By Checked By Page: 7 0108-0275-35 ,AA.t 13. "CKC' L"&#xfd;- Revision:
0 Degradation, dd in Equation (4), is applied in increments of 0.01 inches to determine the effect on the capacity of each size of reinforcing steel. Results are presented in Table 3-3.Table 3-3. Percent Reductions in Reinforcing Steel Load Capacity at Various Degradation Levels Size #8 1 #9 #10 #11 Diameter (in.)1.000 1.128 1.270 1.410 0.000 0.00 0.00 0.00 0.00 0.010 1.99 1.77 1.57 1.41 0.020 3.96 3.51 3.12 2.82 0.030 5.91 5.25 4.67 4.21 Degradation 0.040 7.84 6.97 6.20 5.59 (inches) 0.050 9.75 8.67 7.72 6.97 0.060 11.64 10.36 9.23 8.33 0.070 13.51 12-03 10.72 9.68 0.080 15.36 13.68 12.20 11.03 Figure 3-1 shows a plot of the percent reduction in reinforcing steel load capacities versus degradation for the various reinforcing steel sizes present in the spent fuel pool. Included on the plot are equations relating the percent load capacity reduction of each reinforcing steel size to the degradation level.MPR OA Form: OA-3.1-3, Rev. 0 W FMPR MPR Associates, Inc.320 King Street Alexandria, VA 22314 Calculation No. Prepared By Checked By Page: 8 0108-0275-35 *A. Revision:
0 Figure 3-1. Reductions in Reinforcing Steel Load Capacities at Various Diameter Degradation Levels 14 12.L 0 to C36 U 2 0 r-.0.000 0.010 0.020 0.030 0.040 0.050 0.060 Reinforcing Steel Degradation (inches off diameter)0.070 MPR QA Form: QA-3.1-3, Rev. 0 MPR Associates, Inc.320 King Street Aloxanclra, VA 2314 Calculation No. Prepared By Checked By Page: 9 0108-0275-35 Revision:
0 

==4.0 REFERENCES==
: 1. PSEG Nuclear Drawing No. 201075 A 8706-2, "Salem Nuclear Generating Station, No. I Unit -Fuel Handling Area, Plan at El. 78'-0"." 2. PSEG Nuclear Drawing No. 201076 A 8706-4, "Salem Nuclear Generating Station, No. I Unit -Fuel Handling Area, Plan at El. 84'-0"." 3. PSEG Nuclear Drawing No. 201077 A 8706-8, "Salem Nuclear Generating Station, No. I Unit -Fuel Handling Area, Plan at El. 100'-0" and 116'-0"." 4. PSEG Nuclear Drawing No. 201078 A 8706-4, "Salem Nuclear Generating Station, No. I Unit -Fuel Handling Area, Plan at El. 130'-0"." 5. PSEG Nuclear Drawing No. 201079 A 8706-3, "Salem Nuclear Generating Station, No. I Unit -Fuel Handling Area, Roof Plan." 6. PSEG Nuclear Drawing No. 201080 A 8706-7, "Salem Nuclear Generating Station, No. I Unit -Fuel Handling Area, Sections A-A & B-B." 7. PSEG Nuclear Drawing No. 201081 A 8706-6, "Salem Nuclear Generating Station, No. I Unit -Fuel Handling Area, Sections C-C, D-D & E-E." 8. PSEG Nuclear Drawing No. 201082 A 8706-5, "Salem Nuclear Generating Station, No. I Unit -Fuel Handling Area, Sections F-F & G-G." 9. PSEG Nuclear Drawing No. 201085 A 8706-5, "Salem Nuclear Generating Station, No. I Unit -Fuel Handling Area, Elevation P-P & Str. Bar Schedule." 10. Spiegel, Leonard, and George F. Limbrunner, "Reinforced Concrete Design," Prentice-Hall, Inc.: Englewood Cliffs, NJ, 1980.MPR QA Form: QA-3.1-3.
Rev. 0 D Margin Reduction from Concrete Degradation This appendix contains the following MPR Calculation.
.MPR Calculation 0108-0275-34, "Salem Spent Fuel Pool Structure Capacities Based on Degraded Concrete Conditions," Revision 0.MPR-2613 Revision 3 D-1 MPR Associates, Inc.*M PR 320 King Street Alexandria, VA 22314 CALCULATION TITLE PAGE Client: PSEG Nuclear Page 1 of 15+ Appendices Project: Task No.Salem Spent Fuel Pool Leakage 0108-0303-0275 Title: Calculation No.Salem Spent Fuel Pool Structure Capacities Based on Degraded Concrete Conditions 0108-0275-34 Preparer / Date Checker / Date Reviewer & Approver Date Rev. No.Michelle Heinz Lisa Lichtenauer Robert Keating 0 QUALIII'Y ASSURANCE DOCUJME.NT This document has been prepared.
checked, and reviewed/approved in accordance with the Quality Assurance requirements of 10CFR50 Appendix B. as specified in the MPR Quality Assurance Manual.MPR-OA Form QA 3 1 1, Rev. 0 MPR Associates, Inc.UPMPR 320 King Street Alexandria, VA 22314 0 RECORD OF REVISIONS 0 All Initial Issue Note: The revision number found on each individual page of the calculation carries the revision level of the calculation in effect at the time that page was last revised.MPR QA Form OA-3.1-2, Rev. 0 MVPR Associates, Inc, FAIMPR320 King Street Alexandria, VA 22314 Calculation No. Prepared By Checked By Page: 3 0108-0275-34 Revision:
0 Table of Contents Table of Contents............................................................................
3 1.0 Purpose.................................................................................
4 2.0 Results .. ..................................................
i............................
4 2.1 Moment Capacities...........................................................................
4 2.2 Shear Capacities..............................................................................
5 3.0 Methodology...........................................................................
5 4.0 Calculation 
.............................................................................
6 4.1 Shear ..........................................................................................
6 4.2 Moment.......................................................................................
7 4.2.1 Limiting Spent Fuel Pool Sections ..................................................
7 4.2.2 Working Stress Design Method...............0.......................................
8 4.2.3 Capacities with Concrete Degradation.............................................
11 4.2.4 Transfer Pool.........................................................................
14 5.0 References 
................................................................. 
...........
14 A Reduction in Moment Cap acities Resulting from Concrete Degradation 
.....A-1* Plot of Spent Fuel Pool Moment Capacity Reductions 
........................
B-i MPR OA Form: QA-3.1-3, Rev. 0 MPR Associates, Inc.W3MPR King Street Alexandria, VA 22314 Calculation No. Prepared By Checked By Page: 4 0108-0275-34 J .4..s- -4 Revision:
0 1.0 PURPOSE The purpose of this calculation is to determine the percent reductions in the capacities of the limiting sections of the Salem spent fuel pool structure due to various levels of concrete degradation in the structure.

===2.0 RESULTS===
2.1 Moment Capacities Table 2-1 shows the percent reductions in the moment capacities of the limiting sections of each wall of the spent fuel pool at various levels of concrete degradation.
Also shown in the table are the total moments, allowable moments, and design margins of the limiting sections of each wall with no concrete degradation.
Table 2-1. Percent Reduction in the Allowable Moment of the Spent Fuel Pool Walls% Reduction In Allowable Moment at Various No Concrete Degradation Concrete Degradation Levels (in inches)Wall Total Allowable Limiting Moment Moment Design 1" 2" 3" 4" 5" (kip-ft/ft)* (kit-ft/ft)
Margin North -148 -154 1.04 1.0 2.0 3.0 4.0 5.0 South 258 3.00 1.6 3.1 4.7 6.2 7.8 East 103 1.05 1.5 3.0 4.5 6.0 7.5 West -294 -299 1.02 0.92 1.8 2.8 3.7 4.6 Slab -191 -197 1.03 0.79 1.6 2.4 3.2 3.9*Note that a negative moment indicates compression on the inside (water side) of the pool wall, and tension on the outside of the pool wall (per Appendix C, Page 13 of Reference I).The expressions relating the percent reduction in moment capacity and the concrete degradation level are provided below. In each equation, 'x' represents the concrete degradation level, in inches, and 'y' represents the percent reduction in allowable moment.North Wall: y = 1.006x + 0.0003 MPR OA Form: OA-3.1-3, Rev. 0 MVPR Associates, Inc.PAIIMPR320 King Street Alexandria, VA 22314 Calculation No. Prepared By Checked By Page: 5 0108-0275-34 A Revision:
0 South Wall: y =1.5523x +i 0.00 13 East Wall: y = 1.5042x + 0.0007 West Wall: y = 0.9 167x + 0.0002 Slab: y = 0.7894x + 0.0001 Note that Table 2-1 only addresses the walls of the spent fuel pooi. The walls of the transfer pool, located beside the spent fuel pool, are not expected to show significantly larger reductions in moment capacities.

===2.2 Shear===
Capacities The spent fuel pool wall design margins for shear are sufficiently high that concrete degradation will not have an impact. The spent fuel pool walls are not evaluated in detail for shear in this calculation.

==3.0 METHODOLOGY==
The design analysis of the Salem spent fuel pool building was performed in MPR-1863 (Reference 
: 1) based on the requirements specified in the Salem Structural Design Criteria (Reference 2). The spent fuel pool was divided into approximately seventy sections for the evaluation.
Loads and design margins were calculated for each section.This calculation evaluates only the one section of each spent fuel pool wall having the most limiting design margin for moment. These limiting sections are determined from Reference 1.The depth of concrete degradation is varied by increments of 0.25" for each limiting section, and the reduction in moment capacity based on the degraded concrete conditions is calculated.
Shear capacities at degraded concrete conditions are not specifically evaluated in this calculation, as discussed in Section 4.Although all sections of the spent fuel pool were evaluated for various load combinations, load types, and load directions in Reference 1, it is justifiable to evaluate only the most limiting section of each wall from among all the load combinations, load types, and load directions based on the following:
*Capacities arc limited by either a Multiple of'the working stress allowable or the ultimate str'ength design allowable, depending on the load combination under consideration, per Reference 2, Paragraph 7.2.1. The moment aim- of a concrete section is determined by the MPR OA Form: QA-3.1-3, Rev. 0 MPR Associates, Inc.SM PR 320 King Street Alexandria, VA 22314 Calculation No. Prepared By Checked By Page: 6 0108-0275-34 A Revision:
0 distance between the centroids of the compression block and tension reinforcement.
The compression block is triangular in the working stress design method and rectangular in the ultimate strength design method. Therefore, changes in the overall depth of a concrete section (e.g., resulting from concrete degradation) would have a similar effect on the moment arms resulting from both design methods, and the percent reductions in moment capacities would be comparable.
It is noted that all limiting design margins evaluated in this calculation result from the working stress design method. The design margins based on ultimate strength design are much larger than those based on working strength design (limiting design margins from Appendix C of Reference I are 1.02 and 1.27 based on working stress and ultimate strength capacities, respectively).
The load type (horizontal moment, vertical moment, etc) determines whether the vertical or horizontal rebar carry load. The size and spacing of the rebar in a wall of the structure, which determines the effective area of the rebar, may vary depending on whether the rebar is oriented horizontally or vertically.
However, the percent reduction in moment capacity is most significantly influenced by the depth of the concrete section and is relatively insensitive to the effective area of the steel.0 The direction of a load will determine whether the inside rebar (i.e., on the water side of the wall) or the outside rebar are in tension. Although the size and spacing of the rebar in a wall of the structure may vary depending on whether the rebar is on the inside or outside of a wall, the percent reduction in moment capacity is most significantly influenced by the depth of the concrete section and is relatively insensitive to the effective area of the steel. It is noted that all limiting design margins evaluated in this calculation result from compression of concrete on the inside of the structure, and tension of rebar on the outside of the structure.
As stated above, the wall depth has a significant impact on moment capacity.
PSEG Nuclear drawings (References 5-13) show that the South wall is the only wall not constant in depth from the bottom to the top. However, it is still justifiable to evaluate only one most limiting section from the South wall as all sections have significantly large design margins, per Appendix C of Reference 

====1.4.0 CALCULATION====

===4.1 Shear===
Reference 2, Paragraph 7.2.1, stipulates that shear capacities for the spent fuel pool wall be calculated according to ACT 318-63 (Reference 3). Reference 2 limits normal loads to working stress allowables:
operating basis earthquake loads to 1/3 above working stress allowables; and design basis earthquakc and tornado related loads to ultimate strength design allowables.
Per MPR QA Form: GA-3.1-3, Rev, 0 MPR Associates, Inc.NOMPR 320 King Street Alexandria, VA 22314 Calculation No. Prepared By Checked By Page: 7 0108-0275-34 
,. ..A Revision:
0 Reference 3, Paragraph 1201(a) and Paragraph 1701(a), working stress and ultimate strength design allowables for shear loads are calculated as V = vcbd where: V = allowable shear load for working stress or ultimate strength design v, = allowable shear stress for unreinforced concrete-1. for working stress design (Reference 3, Paragraph 1201(c))= 20 for ultimate strength design (Reference 3, Paragraph 1701(c))b = width of compression face d = distance from extreme compression fiber to centroid of tension reinforcement f,' = concrete compressive strength= capacity reduction factor The above equation shows that shear capacity is a function of the cross-sectional area of the concrete section under consideration, such that the percent reduction in shear capacity is directly proportional to the percent reduction in the depth of the section.Appendix C of Reference I shows that the design margins for shear are large; the most limiting is 1.99. Therefore, the spent fuel pool walls will still be acceptable for shear even under degraded concrete conditions, and the effect of concrete degradation on the shear capacities of the spent fuel pool walls is not specifically evaluated in this calculation.

===4.2 Moment===

====4.2.1 Limiting====
Spent Fuel Pool Sections Table 4-1 shows the design margin, load information, and location of the most limiting section of each spent fuel pool wall considering no concrete degradation.
All values in the table are from Appendix C of Reference 1.MPR QA Form: QA-3.1-3, Rev. 0 MPR Associates, Inc.*M P R 320 King Street Alexandria, VA 22314 Calculation No. Prepared By Checked By Page: 8 0108-0275-34 , .Revision:
0 Table 4-1. Limiting Design Margins of Each Spent Fuel Pool Wall Location Total Alongoa Load Type Moment Margin Wall i Wall Height Combination (kip-ftift) (klt-ft)North Middle Bottom Normal-148 
-154 1.04 Operation Moment South West Bottom Normal Horizontal  258 3.00 Operation Moment East Middle Bottom Normal-98 
-103 1.05 Operation Moment West Middle Top East-West OBE Horizontal 
-294 -299 1.02_______ _______ ________Moment 
-9 29 10 Slab Middle Middle Normal Horizonta 
-191 -197 1.03 Operation Moment I _Table 4-1 shows that the limiting design margins result from only two load combinations 
-Normal Operation and East-West Operating Basis Earthquake (OBE). Per Reference 2, Paragraph 7.2.1, the moment capacities for the spent fuel pool structure are to be calculated according to Reference 
: 3. Per Reference 3, moment capacities are limited to working stress allowables for normal loads and 1/3 above working stress allowables for OBE loads. Therefore, the working stress design method is used in this calculation to determine the percent reduction in the moment capacities of the limiting spent fuel pool sections when the concrete of those sections is degraded.4.2.2 Working Stress Design Method The working stress design method (detailed in Reference 
: 4) assumes that the area of the reinforcing steel in a structure may be thought of as replaced by an equivalent area of concrete, scaled by the ratio of the moduli, n: E,(1)where: E, = concrete modulus of elasticity E, = reinforcing steel modulus of elasticity 0 This is illustrated in Figure 4- 1.MPR OA Form: QA-3.1-3, Rev. 0 MPR Associates, Inc.OIMPR 320 King Street Alexandria, VA 22314 Calculation No. Prepared By Checked By Page: 9 0108-0275-34 
.,4 .'c. Revision:
0 kd n/s d id j nA._________________fs i~ ~ ~ ~ ~ ~~. .................. 
...i Figure 4-1. Working Stress Design Method In order to find the neutral axis, designated by 'kd' in Figure 4-1, the moments of the compression and tension areas about the neutral axis are balanced, giving b (kd) nA,(d- kd) =0 2 where: A, =equivalent area of tension reinforcement d = distance from extreme compression fiber to centroid of tension reinforcement b = width of concrete section k = ratio of the distance from the extreme compression fiber to the neutral axis and the distance 'd'If the reinforcing ratio is A, bd (2)MPR QA Form: QA-3.1-3, Rev. 0 MPR Associates, Inc.FIMPR 320 King Street Alexandria, VA 22314 Calculation No. Prepared By Checked By Page: 10 0108-0275-34 .tRevision:
0 Then b (kd-_2 -pnbd(d- kd) = 0 2 Solving for k gives k = j2pn + (pn)2-pn (3)If the concrete is cracked and is not supporting tension, the stress profile resembles that shown in Figure 4-1, and the distance between the compressive and tensile forces is kd jd=d---3 Dividing through by 'd' yields j1 k 3 (4)The moment balance requires that the compression force equals the tension force (C = T), and that the external bending moment M be equal to Tjd = Cjd. If yielding of the steel is limiting, and T = A~f., then M = Asfjd (5)where: f= allowable stress of steel for working stress design If the concrete limits, then C = '/2f,'kdb, and M = 1/22f,'k jd 2 b Where: G' = concrete compressive strength MPR QA Form: QA-3.1-3, Rev. 0 MPR Associates, Inc.PIM PR 320 King Street Alexandria, VA 22314 Calculation No. Prepared By Checked By Page: I I 0108-0275-34 ,J. A t_ ...Revision:
0 Per References 5-13, the walls of the spent fuel pool are very lightly reinforced (reinforcing steel of size #11 or less is used). Therefore, this analysis assumes that yielding of the steel will limit the failure.4.2.3 Capacities with Concrete Degradation Equations The equations used to calculate the reduced moment capacities of the limiting spent fuel pool sections are summarized below. Note that equations (1) through (5) were developed in Section 4.2.Equation (1) -Modulus Ratio n =L Equation (2) -Reinforcing Ratio A, bd Equation (3) -Neutral Axis Location k = &#xfd;2pn + (pn)2 -pn Equation (4) -Distance between Compressive and Tensile Forces k j =1---3 Equation (5) -Working Stress Design Moment (Allowable Moment)M = A~fjd Equation (6) -Distance from Extreme Compression Fiber to Centroid of Tension Reinforcement
: d. = l.- (dd + d,)MPR DA Form: OA-3.1-3, Rev. 0 UMPR MPR Associates, Inc.320 King Street Alexandria, VA 22314 Calculation No. Prepared By Checked By Page: 12 0108-0275-34 tA. Ls Lj-.. Revision:
0 where: L = full depth of concrete section dd = depth of steel from the nearest wall edge (refer to Figure 4-1)d, = depth of concrete degradation (refer to Figure 4-1)Equation (7) -Percent Reduction in Original Moment Capacity% Re duction = 100{1 -MomentCapacity with-deg nadation MomentCapacity no-deg radation The results of these calculations for each limiting spent fuel pool section are provided in Appendix A. A plot of the percent reductions in moment capacities versus the concrete degradation levels is provided in Appendix B. Equations relating the percent reductions in moment capacities to the concrete degradation level are shown on the plot.Material Properties The material properties to be used in equations (1) through (7) to calculate the reduction in the moment capacities of the limiting wall sections at degraded conditions are listed belOw, along with references for each.re' = strength of concrete fy = yield strength of steel E, = modulus of steel f, = steel allowable stress= 3,500 psi= 60,000 psi= 29 x 106 psi= 24,000 psi (Reference 5)(Reference 5)(Reference 3, Paragraph 1100)(Reference 3, Paragraph 1003(a))Note that Paragraph 1003(a) of Reference 3 limits the allowable stress of steel in working stress design, f 5 , to 24,000 psi for deformed bars with a yield strength of 60,000 psi or more and in sizes#11 and smaller reinforcing steel. Per References 5-13, all reinforcing steel in the spent fuel pool structure is smaller than #11.The modulus of concrete, E,, is calculated as E: = wI533(f'"J/ 
= 3.41 x 106 psi (Reference 3, Paragraph 1102(a))where w = 145 Ibs/ft 3 (Rcfcrcnce 3, Paragraph 1102(a))MPR OA Form: OA-3.1-3, Rev. 0 MPR Associates, Inc.*M P R 320 King Street Alexandria, VA 22314 Calculation No. Prepared By Checked By Page: 13 0108-0275-34 j ,/,cf t_- Revision:
0 Dimensions The spent fuel pool wall and reinforcing steel dimensions to be used in equations (1) through (7)to calculate the reductions in moment capacities of the limiting spent fuel pool sections are discussed in this section. Note that the limiting section location information and the reinforcing steel direction and location information provided in Table 4-2 are taken from Appendix C of Reference 1.The full depth of the limiting sections of the spent fuel pool walls, L, is determined from References 5-13. Because the values of the depths of steel from the nearest wall edges shown in References 5-13 are to the edges of the reinforcement steel, dd is calculated by summing the values of the depths from References 5-13 and 1/22 the diameter of the steel reinforcement, taken from Reference 
: 14. The dimensions used in the calculations are summarized in Table 4-2.The size and spacing of reinforcing steel within a structure determine the equivalent area, A,, of the steel. The size and spacing of the reinforcing steel in the limiting section of each spent fuel pool wall is provided in Table 4-2. This information was provided by References 5-13. Table 4-2 also shows the equivalent areas of the reinforcing steel, taken from Reference 14.Table 4-2. Wall and Reinforcing Steel Dimensions Reinforcing Full Depth Depth of Equivalent Limiting Section Location Reinforcing Reinforcing Reinforcing Steel Steel Steel Steel Size & of Wall, L Steel, dd Area. As Wall Along Wall Height Direction Location*
Spacing (in.) (in.) (in.) (in North Middle Bottom Vertical Outside #8@12" 1.00 105 4.0 0.79 South West Bottom Horizontal Outside #11 @9" 1.41 72 5.705 2.08 East Middle Bottom Vertical Outside #8012" 1.00 72 4.25 0.79 West Middle Top Horizontal Outside #8@9" 1.00 115 4.0 1.05 Slab Middle Middle North-South Outside #8@ 12" 1.00 132 3.5 0.79'Inside' refers to the water side of the wall. 'Outside' refers to the side of the wall remote from the water.The unit width of a concrete section used to calculate moment capacities in this calculation, b, is 12 inches. The depth of concrete degradation, d,,, is varied by increments of 0.25" for each limiting section to determine the result on the moment capacity of each section.MPR GA Form: GA-3.1-3, Rev. 0 MPR Associates, Inc.*M P R 320 King Street Alexandria, VA 22314 Calculation No. Prepared By Checked By Page: 14 0108-0275-34 , Revision:
0 4.2.4 Transfer Pool The transfer pool is located next to the spent fuel pool, with the two pools separated by the wall previously indicated as the 'south wall' in this calculation.
The walls of the transfer pool were not evaluated in Reference 
: 1. While the transfer pool walls are not specifically evaluated in this calculation, a short assessment regarding the effect of concrete degradation on the capacity is made in this section.The results of Sections 4.2.1 -4.2.3, in particular the figure in Appendix B, show that the reduction in moment capacity due to concrete degradation is most dependent on the depth of a section; the smaller thesection depth, the higher the reduction in moment capacity.
References 5-13 show that the transfer pool walls have the following full depths: Slab: 78" South Wall: 140" East Wall: 72" (same as the east wall of the spent fuel pool)West Wall: 115" (same as the west wall of the spent fuel pool)North Wall: varies from 48" to 72" (same as the south wall of the spent fuel pool)Note that none of the transfer pool walls have a smaller depth than any of the spent fuel pool walls (see Section 3.0 for a discussion of the varying depth of the spent fuel pool south wall/transfer pool north wall). Therefore, the percent reduction in moment capacities for the transfer pool walls are not expected to be significantly greater than those for the spent fuel pool walls.

==5.0 REFERENCES==
: 1. MPR-1863, "Salem Generating Station Spent Fuel Pool Building Structural Design Analysis," Revision 0, 2. PSEG Nuclear Department Technical Standard SC.DE-TS.ZZ-4201(Q), "Salem Structural Design Criteria," Revision 1.3. American Concrete Institute Standard ACI 318-63, "Building Code Requirements for Reinforced Concrete," June, 1963.4. Winter, George and Arthur H. Nilson, "Design of Concrete Structures," McGraw-Hill Book Company: New York., 1979.5. PSEG Nuclear Drawing No. 201075 A 8706-2, "'Salem Nuclear Generating Station, No. 1 Unit -Fuel Handling Area, Plan at El: 78'-0"." MPR QA Foirn: OA-3,1-3, Rev, 0 MPR Associates, Inc.FAMPR 320 King Street Alexandria, VA 22314 Calculation No. Prepared By Checked By Page: 15 0108-0275-34 b ,.-,OD.o&#xfd; Revision:
0 6. PSEG Nuclear Drawing No. 201076 A 8706-4, "Salem Nuclear Generating Station, No. I Unit -Fuel Handling Area, Plan at El. 84'-0"." 7. PSEG Nuclear Drawing No. 201077 A 8706-8, "Salem Nuclear Generating Station, No. I Unit -Fuel Handling Area, Plan at El. 100'-0" and 116'-0"." 8. PSEG Nuclear Drawing No. 201078 A 8706-4, "Salem Nuclear Generating Station, No. 1 Unit -Fuel Handling Area, Plan at El. 130'-0"." 9. PSEG Nuclear Drawing No. 201079 A 8706-3, "Salem Nuclear Generating Station, No. 1 Unit -Fuel Handling Area, Roof Plan." 10. PSEG Nuclear Drawing No. 201080 A 8706-7, "Salem Nuclear Generating Station, No. I Unit -Fuel Handling Area, Sections A-A & B-B." 11. PSEG Nuclear Drawing No. 201081 A 8706-6, "Salem Nuclear Generating Station, No. I Unit -Fuel Handling Area, Sections C-C, D-D & E-E." 12. PSEG Nuclear Drawing No. 201082 A 8706-5, "Salem Nuclear Generating Station, No. I Unit -Fuel Handling Area, Sections F-F & G-G." 13. PSEG Nuclear Drawing No. 201085 A 8706-5, "Salem Nuclear Generating Station, No. I Unit -Fuel Handling Area, Elevation P-P & Str. Bar Schedule." 14. Spiegel, Leonard, and George F. Limbrunner, "Reinforced Concrete Design," Prentice-Hall, Inc.: Englewood Cliffs, NJ, 1980.MPR OA Foim: OA-3.1-3, Rev. 0 MPR Associates, Inc.320 King Street*IM P R Alexandria, VA 22314 Calculation No. Prepared By Checked By Page: A-I 0108-0275-34 , Revision:
0 A Reduction in Moment Capacities Resulting from Concrete Degradation MPR QA Form: QA-3.1-3, Rev. 0 0 North Wall, Water Side Compression, #8 Rebar, 8.75' Wall Thickness Allowable Stress of Steel, Working Stress Design, 1, (psi) 24000 Strength of Concrete, f.' (psi) 3500 Modulus of Concrete, F,, (psi) 3.41 E+06 Modulus of Steel, E., (psi) 2.90E+07 Modulus ratio, n = E,/E. 8.50E+00 Equivalent Steal Area, A. (in') 0.79 Unit Width/Height of Wall, b (In.) 12 Depth of steel from nearest wall edge, dd (in.) 4 Full depth of wall, L (In.) 105 Distance from Extreme Neutral Axis Distance Depth of Compression Fiber Reinforcing Location, between Moment %Concrete to Centrold of Ratio, k=eqrt[2*p*n Compressive Capacity, M Reduction Degradation, Tension p = AJ(bd) +(p*n)2 and Tensile (ftjkip/ft)
In Moment d~(n) Tnin = ~ -Forces, Capkeplfy dw (in.) Reinforcement, d = p Fnorc Capacity[L-(dd+d,.)] (in.) 1_1_k/3 0 101.0 6.52E-04 9.99E-02 9.67E-01 154.3 0.00 0.25 100.8 6.53E-04 1.OOE-01 9.67E-01 153.9 0.25 0.50 100.5 6.55E-04 1.OOE-01 9.67E-01 153.5 0.50 0.75 100.3 6.57E-04 1.OOE-01 9.67E-01 153.1 0.75 1.00 100.0 6.58E-04 1.OOE-01 9.67E-01 152.7 1.01 1.25 99.8 6.60E-04 11.001E-01 9.67E-01 152.3 1.26 1.50 99.5 6.62E-04 1.01E-01 9.66E-01 151.9 1.51 1.75 99.3 6.63E-04 1.01E-01 9.66E-01 151.5 1.76 2.00 99.0 6.65E-04 1I01E-01 9.68E-01 151.2 2.01 2.25 98.8 6.67E-04 1.011E-01 9.66E-01 150.8 2.26 2.50 98.5 6.68E-04 1.01E-01 9.66E-01 150.4 2.52 2.75 98.3 6.70E-04 1.01E-01 9.66E-01 150.0 2.77 3.00 98.0 6.72E-04 1.01E-01 9.66E-01 149.6 3.02 3.25 97.8 6.73E-04 1.01E-01 9.66E-01 149.2 3.27 3.50 97.5 6.75E-04 1.02E-01 9.66E-01 148.8 3.52 3.75 97.3 6.77E-04 1.02E-01 9.66E-01 148.4 3.77 4.00 97.0 6.79E-04 1.02E-01 9.66E-01 148.1 4.02 4.25 96.8 6.80E-04 1.02E-01 9.66E-01 147.7 4.28 4.50 96.5 6.82E-04 1.02E-01 9.66E-01 147.3 4.53 4.75 96.3 6.84E-04 1.02E-01 9.66E-01 146.9 4.78 5.00 96.0 6.86E-04 1.02E-01 9.66E-01 146.5 5.03 5,25 95.8 6.88E-04 1.02E-01 9.66E-01 146.1 5.28 5.50. 95.5 6.89E-04 1.03E-01 9.66E-01 145.7 5.53 5.75 95.3 6.91E-04 1.03E-0l 9.66E-01 145.3 5.78 6.00 95.0 6.93E-04 1.03E-01 9.66E-01 145.0 6.04 MPR Calculation 0108-0275-34, Appendix A Page A2 0 West Wall, Water Side Compression, #8 Rebar, 9.583' Wall Thickness Allowable Stress of Steel, Working Stress Design, f. (psi) 24000 Strength of Concrete, fV' (psi) 3500 Modulus of Concrete, E., (psi) 3.41E+06 Modulus of Steel, E., (psi) 2.90E+07 Modulus ratio, n = EJE./ 8.50E+00 Equivalent Steel Area, A. (In') 1.06 Unit Wldth/lHelght of Wall, b (in.) 12 Depth of steel from nearest wall edge, dda (in.) 4 Full depth of wall, L Jin.) 115 Distance from D h Extreme Neutral Axis Distance Depth of Compression Reinforcing Location, between Moment %Concrete Fiber to Centrold Ratio, k=sqrt[2*p~n Compressive Capacity., Reduction Degradation, -and Tensile p In Moment dw(n) of Tension p = AJ(bd) +(p'n)2] Foce, M (ft~kip/h)I omn d.(n)Forces, Capacity Reinforcement, d = 1=1-k/3 I[L-(dd+d.)] (in.)0.00 1l11.0 7.88EL-04 1.09E-01 9.64E-01 ' 224.6 0.00 0.25 110.8 7.90E-04 1.09E-01 9.84E-01 224.1 0.23 0.50 110.5 7.92E-04 1.10E-01 9.63E-01 223.6 0.46 0.75 110.3 7.941-04 1.10E-01 9.63E-01 223.1 0.69 1.00 110.0 7.95E-04 1.10E-01 9.63E-01 222.5 0.92 1.25 109.8 7.97E-04 1.101-01 9.63E-01 222.0 1.15 1.50 109.5 7.99E-04 1.10E-01 9.63E-01 221.5 1.38 1.75 t09.3 8.01E-04 1.10E-01 9.63E-01 221.0 1.60 2.00 109.0 8.03E-04 1.10E-01 9.63E-01 220.5 1.83 2.25 108.8 8.05E-04 1,1E-01 9.63E-01 220.0 2.06 2.50 108.5 8.06E-04 1.10E-01 9.63E-01 219.5 2.29 2.75 108.3 8.08E-04 1.1IE-01 9.63E-01 218.9 2.52 3.00 108.0 8.1OE-04 1.11E-01 9.63E-01 218.4 2.75 3.25 107.8 8.12E-04 1.11E-01 9.63E-01 217.9 2.98 3.50 107.5 8,14E-04 1.1iE-01 9.63E-01 217.4 3.21 3.75 107.3 8.16E-04 1.1 IE-01 9.63E-01 216.9 3.44 4.00 107.0 8.18E-04 1.11E-01 9.63E-01 216.4 3.67 4.25 106.8 8.20E-04 1.11E-01 9.63E-01 215.9 3.90 4.50 106.5 8.22E-04 1.11E-01 9.63E-01 215.3 4.13 4.75 106.3 8.24E-04 1.12E-01 9.63E-01 214.8 4.35 5.00 106.0 8.25E-04 1.12E-01 9.63E.01 214.3 4.58 5.25 105.8 8.27E-04 1.12E-01 9.63E-01 213.8 4.81 5.50 105.5 8.29E-04 1.12E-01 9.63E-01 213.3 5.04 5.75 105.3 8.31E-04 1.12E-01 9.63E-01 212.8 5.27 6.00 105.0 8.33E-04 1.12E-01 9.63E-01 212.3 5.50'Per Table 4-. in the main body of the calculation, the limiting design margin for the west wall is based on the OBE allowable, which Is 1/3 above normal allowables.
Although the moment capacities in the above table are normal aflowables, the percent reductions for OBE allowabtas are the same.MPH Calculation 0108-0275-34, Appendix A Page A3 0 East Wall, Water Side Compression, #8 Rebar, 6' Wall Thickness Allowable Stress of Steel, Working Stress Design, f. (psi) 24000 Strength of Concrete, f.' (pal) 3500 Modulus of Concrete, E., (psi) 3.412E+06 Modulus of Steel, E., (psi) 2.90E+07 Modulus ratio, n = EJE, 8.50E+00 Equivalent Sleei Area, A. (in') 0.79 Unit Width/Height of Wall, b (in.) 12 Depth of steel from nearest wall edge, dd (in.) 4.25 Full depth of wall, L (in.) 72 Distance from Extreme Neutral Axis Distance Concreteh of Compression Fiber Reinforcing Location, between Moment %Cocrdti, to Centrold of Ratio, k=sqrt[2*pln Compressive Capacity, M Reduction dw (in.) Tension p = A./(bd) +(p*n)2] -and Tensile (ft-kipilt)
In Moment datin, Forces, Capacity Reinforcement, d = pn J=1-k/3 I_ [L-(dd+dw)] (in.)0.00 67.8 9.72E-04 1.21E-01 9.60E-01 102.7 0.00 0.25 67.5 9.75E-04 1.21E-01 9.60E-01 102.4 0.38 0.50 67.3 9.79E-04 1.21E-01 9.60E-01 102.0 0.75 0.75 67.0 9.83E-04 1.21E-01 9.60E-01 101.6 1.13 1.00 66.8 9.86E-04 1.21E-01 9.60E-01 101.2 1.50 1.25 66.5 9.90E-04 1.22E-01 9.59E-01 100.8 1.88 1.50 66.3 9.94E-04 1.22E-01 9.69E-01 100.4 2.26 1.75 66.0 9.97E-04 1.22E-01 9.59E-01 100.0 2.63 2.00 65.8 1.00E-03 1.22E-01 9.59E-01 99.7 3.01 2.25 65.5 1.01E-03 1.22E-01 9.59E-01 99.3 3.39 2.50 65.3 1.01E-03 1.23E-01 9.59E-01 98.9 3.76 2.75 65.0 1.01E-03 1.23E-01 9.59E-01 98.5 4.14 3.00 64.8 1.02E-03 1.23E-01 9.59E-01 98.1 4.51 3.25 64.5 1.02E-03 1.23E-01 9.59E-01 97.7 4.89 3.50 64.3 1.02E-03 1.24E-01 9.592-01 97.3 5.27 3.75 64.0 1.03E-03 1.24E-01 9.59E-01 96.9 5.64 4.00 63.8 1.03E-03 1.24E-01 9.59E-01 96.6 6.02 4.25 63.5 1.04E-03 1.24E-01 9.59E-01 96.2 6.39 4.50 63.3 1.04E-03 1.24E-01 9.59E-01 95.8 6.77 4.75 63.0 1.04E-03 1.25E-01 9.58E-01 95.4 7.15 5.00 62.8 1.05E-03 1.25E-01 9.58E-01 95.0 7.52 5.25 62.5 1.05E-03 1.25E-01 9.58E-01 94.6 7.90 5.50 62.3 1.06E-03 1.25E-01 9.58E-01 94.2 8.27 5.75 62.0 1.06E-03 1.26E-01 9.58E-01 93.9 8.65 6.00 61.8 1.07E-03 1.262-01 9.58E-01 93.5 9.03 MPR Calculation 0108-0275-34, Appendix A Page A4 0 Slab, Water Side Compression, #8 Rebar, 11' Slab Thickness Allowable Stress of Steel, Working Stress Design, 1. (poll 24000 Strength of Concrete, f,' (psi) 3500 Modulus of Concrete, E., (psi) 3.41E+06 Modulus of Steel, E,, (psi) 2.90E+07 Modulus ratio, n = EJEC 8.50E+00 Equivalent Steel Area, A, (inz) 0.79 Unit Width/Height of Wall, b (in.) 12 Depth of steel from nearest wall edge, dd (in.) 3.5 Full depth of wall. L (in.) 132 Distance fromT Extreme Neutral Axis Distance Depth of Compression Fiber Reinforcing Location, between Moment %Concrete to Centrold of Ratio, k=sqrt[2*p*n Compressive Capacity, M Reduction Degradation, ensin j and Tensile in Moment d .I n ) T n s o n p = A .~ b ) + p n 21 F o rc e s , ( Wt k ip /ft) C a p a c ity Reinforcement, d = pn J=1 -Ca3 I[L-(dd+d6)] (in.)0.00 128.5 5.12E-04 8.91E-02 9.70E-01 197.0 0.00 0.25 128.3 5.132-04 8.92E-02 9.70E-01 196.6 0.20 0.50 128.0 5.14E-04 8.93E-02 9.70E-01 196.2 0.39 0.75 127.8 5.15E-04 8.93E-02 9.70E-01 195.8 0.59 1.00 127.5 5.16E-04 8.94E-02 9.70E-01 195.4 0.79 1.25 127.3 5.17E-04 8.95E-02 9.702-01 195.1 0.99 1.50 127.0 5.18E-04 8,98E-02 9.70E-01 194.7 1.18 1.75 126.8 5.19E-04 8.97E-02 9.70E-01 194.3 1.38 2.00 126.5 5.20E-04 8.98E-02 9.70E-01 193.9 1.58 2.25 126.3 5.21E-04 8.98E-02 9.70E-01 193.5 1.78 2.50 126.0 5.22E-04 8.99E-02 9.70E-01 193.1 1.97 2.75 125.8 5.24E-04 9.00E-02 9.70E-01 192.7 2.17 3.00 125.5 5.25E-04 9.01 E-02 9.70E-01 192.3 2.37 3.25 125.3 5.26E-04 9.02E-02 9.70E-01 191.9 2.57 3.50 125.0 5.27E-04 9.03E-02 9.70E-01 191.6 2.76 3.75 124.8 5.28E-04 9.04E-02 9.70E-01 191.2 2.96 4.00 124.5 5.29E-04 9.04E-02 9.70E-01 190.8 3.16 4.25 124.3 5.30E-04 9.05E-02 9.70E-01 190.4 3.36 4.50 124.0 5.31E-04 9.061-02 9.70E-01 190.0 3.55 4.75 123.8 5.32E-04 9.07E-02 9.70E-01 189.6 3.75 5.00 123.5 5.33E-04 9.08E-02 9.70E-01 189.2 .3.95 5.25 123.3 5.34E-04 9.09E-02 9.70E-01 188.8 4.14 5.50 123.0 5.35E-04 9.10E-02 9.70E-01 188.4 4.34 5.75 122.8 5.36E-04 9.11E-02 9.70E-01 188.1 4.54 8.00 122.5 5.37E-04 9.11E-02 9.70E-01 187.7 4.74 MPR Calcualation 0108-0275-34, Appendix A Page A5 Bottom South Wall, Water Side Compression, #11 Rebar, 6' Wall Thickness Allowable Stress of Steel, Working Stress Design, f. (psi) 24000 Strength of Concrete, f' (psi) 3500 Modulus of Concrete, E., (psi) 3.41E+06 Modulus of Steel, E., (psi) 2.90E+07 Modulus ratio, n = E,/E1 8.50E+00 Equivalent Steel Area, A, (in) 2.08 Unit Width/Height of Wall, b (in.) 1 Depth of steel from nearest wall edge, dd (in.) 5.705 Full depth of wall, L (in.) 72&#xfd;Distance from Extreme Neutral Axis Distance Depth of Compression Fiber Reinforcing Location, between Moment %Concrete to Centrold of Ratio, k=sqrt[2*p-n Compressive Capacity, M Reduction Degradation, Tension p = AJ(bd) +(p*n)2] -and Tensile (ft'klplft)
In Moment dw(n)Forces, Capacity Reinforcement, d = pn j=l-kl3[L-(dd+dw)] (in.)0.00 66.3 2.61E-03 1.90E-01 9.37E-01 258.3 0.00 0.25 66.0 2.62E-03 1.90E-01 9.37E-01 257.3 0.39 0.50 65.8 2.63E-03 1.90E-01 9.37E-01 256.3 0.78 0.75 65.5 2.64E-03 1.91E-01 9.36E-01 255.3 1.17 1.00 65.3 2.65E-03 1.91E-01 9.36E-01 254.3 1.55 1.25 65.0 2.66E-03 1.91E-01 9.36E-01 253.3 1.94 1.50 64.8 2.68E-03 1.92E-01 9.36E-01 252.3 2.33 1.75 64.5 2.69E-03 1.92E-01 9.36E-01 251.3 2.72 2.00 64.3 2.70E-03 1.92E-01 9.36E-01 250.3 3.11 2.25 64.0 2.71E-03 1.93E-01 9.36E-01 249.3 3.50 2.50 63.8 2.72E-03 1.93E-01 9.36E-01 248.3 3.88 2.75 63.5 2.73E-03 1.93E-01 9.36E-01 247.3 4.27 3.00 63.3 2.74E-03 1.94E-01 9.35E-01 246.3 4.66 3.25 63.0 2.75E-03 1.94E-01 9.35E-01 245.3 5.05 3.50 62.8 2.76E-03 1.94E-01 9.35E-01 244.3 5.44 3.75 62.5 2.77E-03 1.95E-01 9.35E-01 243.3 5.82 4.00 62.3 2.78E-03 1.95E-01 9.35E-01 242.3 6.21 4.25 62.0 2.79E-03 1.96E-01 9.35E.01 241.3 6.60 4.50 61.8 2.80E-03 1.96E-01 9.35E-01 240.3 6.99 4.75 61.5 2.82E-03 1.96E-01 9.35E-01 239.3 7.38 5.00 61.3 2.83E-03 1.97E-01 9.34E-01 238.3 7.76 5.25 61.0 2.84E-03 1.97E-01 9.34E-01 237.3 8.15 5.50 60.8 2.85E-03 1.97E-01 9.34E-01 236.3 8.54 5.75 60.5 2.86E-03 1.98E-01 9.34E-01 235.3 8.93 6.00 60.3 2.872-03 1.98E-01 9.34E-01 234.3 9.32 MPR Calculation 0108-0275-34, Appendix A Page A6 MPR Associates, Inc.FIM PR 320 King Street Alexandria, VA 22314 Calculation No. Prepared By Checked By Page: B-i 0108-0275-34 Revision:
0 B Plot of Spent Fuel Pool Moment Capacity Reductions MPR OA Form: OA-3.1-3, Rev. 0 Percent Reduction in SFP Wall & Slab Allowable Moment at Various Concrete Degradation Levels 10.0 __"__----North Wall, Water Side Compression, #8 Rebar, 8.75' Wall Thickness 9.0 --- West Wall, Water Side Compression, #8 Rebar, 9.583' Wall Thickness-A- -East Wall, Water Side Compression, #8 Rebar, 6' Wall Thickness 8.0--- Slab, Water Side Compression, #8 Rebar, 1 I'Slab Thickness 7.0 --" -Bottom South Wall, Water Side Compression, #11 Rebar, 6' Wall .Thickness 6 .0 .. .... .o Y = 1.5523x + 0.0013 (south wall) _-4--< 5.0 y,. 1.5042x + 0.0607-(e-ast-7a-.
4.0 y 1.006x + 0.0003 (north wall)1y = 0.9167x + 0.0002 (west wall)3.0 -y = 0.7894x + 0.0001 (slab)2.0 --........
1.0 ... ....... ....0.0 1 0.0 1.0 2.0 3.0 4.0 5.0 6.0 Concrete Degradation (inches)MPR Calculation 0108-0275-34, Appendix B Page B2}}