ML12174A052

E-mail from G. Thomas, NRR to A. Sheikh, NRR, Et Al; Subject: Predecisional - Official Use Only: Preliminary Draft Response for Comment Reg. Alkali-Silica Reaction (ASR) Issue at Seabrook

ML12174A052
Person / Time
Site:	Seabrook
Issue date:	12/20/2011
From:	George Thomas Office of Nuclear Reactor Regulation
To:	Chaudhary S K, Modes M C, Sheikh A H Office of Nuclear Reactor Regulation, NRC Region 1
References
FOIA/PA-2012-0119
Download: ML12174A052 (23)

Text

1>'*1I~Turilin, Andrey From: Sent: To: Cc:

Subject:

Thomas, George -- oZAK-Tuesday, December 20, 2011 12:05 PM Sheikh, Abdul; Modes, Michael; Chaudhary, Suresh; Raymond, William; Graves, Herman;Pires, Jose; Lehman, Bryce Conte, Richard; Auluck, Rajender; Hogan, Rosemary; Murphy, Martin; Manoly, Kamal;Farzam, Farhad; Burritt, Arthur PREDECISIONAL

-OFFICIAL USE ONLY: Preliminary Draft Response for Comment reg.Alkali-Silica Reaction (ASR) Issue at Seabrook Seabrook ASR Prelim DRAFT Response (OUO) for Comment 12-20-11.docx Attachments: (JI it Thanks.George Thomas Structural Engineer NRR/DE/EMCB 301-415-6181 From: Hiland, Patrick Sent: Tuesday, December 20, 2011 7:47 AM To: Miller, Chris Cc: Conte, Richard; Thomas, George

Subject:

FW: Briefing Material for Tomorrow's Call at 200pm on Seabrook ASR Importance:

High (b)(5)From: Conte, Richard Sent: Monday, December 19, 2011 1:30 PM To: Auluck, Rajender; Ayres, David; Bahadur, Sher; Burritt, Arthur; Chaudhary, Suresh; Chernoff, Harold; Clifford, James;Cline, Leonard; Conte, Richard; Cruz, Holly; Delligatti, Mark; Evans, Michele; Ferrer, Nathaniel; Galloway, Melanie; Hiland, Patrick; Howe, Allen; Khanna, Meena; Lamb, John; Lehman, Bryce; Lund, Louise; Manoly, Kamal; Miller, Chris; Miller, Ed;Modes, Michael; Morey, Dennis; Murphy, Martin; Plasse, Richard; Raymond, William; Roberts, Darrell; Sakai, Stacie;Sheikh, Abdul; Thomas, George; Wilson, Peter

Subject:

Briefing Material for Tomorrow's Call at 200pm on Seabrook ASR Information In this record was deleted the Freedom of INformation Act, exeniptiofls FOIA- -O2- \. .(7 1//1 See attached agenda and talking points along with attachments for more details.2

~f PRELIMINARY DRAFT RESPONSE TO REQUEST FOR TECHNICAL ASSISTANCE FOR SEABROOK STATION ALKALI-SILICA REACTION DEGRADATION OF CONCRETE (PREDECISIONAL

-OFFICIAL USE ONLY FOR INTERNAL BRAINSTORMING

& COMMENT)

1.0 INTRODUCTION

By letter dated September 12, 2011 (Agencywide Documents Access and Manag nt Sys em (ADAMS) Accession No. ML1116105300), the U.S. Nuclear Regulatory Co sion M Region I Office requested technical assistance from the Office of Nuclear Ron (NRR) to evaluate the potential consequence of alkali-silica reaction ( ) a of a safety-related concrete structure at Seabrook Station. More specifi ally, d NRR review for adequacy of a NextEra prompt operability determination (POD)' Ited open issues, NRC staff should be-able to identify what additional i tib- ,ne d in order to fully evaluate the impact of the degradation on the current I nsin si n basis in the final operability determination for structures important-to-safety the pla s h primary case for review, NextEra evaluated the Seabrook Control Buildi lectri Tunnel and Penetration Room) in light of the recently discovered degradati e3 i er structures important-to-safety within the scope of the maintenance rule h e aso 1: affected by the ASR problem.Region I requested NRR assistance to-addres e a ove ncerns by providing answers to the five Task Interface Agreement (TIA) que w ed in Section 3.0 "Evaluation" of this response.

2.0 BACKGROUND

NextEra Energy (the licensee d co rete core samples.from the interior surface of exterior walls o the Cont n of their assessment to support renewal of their license. In AugL st 20101 sts u as a part of the core sample analysis reported a change in mate ial e T e nalysis reported the presence of ASR-degradation in core samples taken r chr ic walls below grade, with reductions reported in the concrete compressive st h mo of elasticity from that expected.

NextEra evaluated these parametric r i0 o d in the impact on the design basis of the Control Building.

By their pro th lice erforr d an immediate and prompt operability determination (POD)and conc .rily, that the Control Building (CB) was operable but with reduced stre s to esign capacity.Ne ntinue to evaluate the extent of this condition for five other safety related concrete buildin The other five buildings for which concrete core samples were taken were: Equipment Vault (ho ing ECCS equipment including that for Residual Heat Removal (RHR)], Radiological Controls Area (RCA) Walkway, Emergency Feedwater Building (EFW), Emergency Diesel Generator (EDG) Building, and the Containment Enclosure Building (CEB). As of June 30, 2011 there are two open prompt operability determinations, one for the Control Building and one for the other five buildings collectively.

The licensee found additional evidence of ASR in four of the five other buildings and they evaluated that information in a separate immediate and prompt R1 S 4I A OFFI L US CN/(FOR4 E'Rli BRA RM Cý ýET) operability determination using the same evaluation techniques as for the Control Building.

This evaluation is also considered preliminary or open. Based on NRC internal discussions, it appears that the calculation methods and correlations that NextEra used in their prompt operability determination may not be fully appropriate in light of the ASR problem.NextEra's planned actions are two-fold:

1) to follow their operability determination process; and, 2) to follow the guidance in NEI 95-10, "Industry Guideline for Implementing the Requirements of 10 CFR Part 54 -The License Renewal Rule," to develop an aging manage t program to support the license renewal application.

Possible outcomes to the PODs are: 1) r red conditions (which may not be possible);

2) resolved conditions (use "as is" ce c ge incorporated or Action Request (AR) disposition approved);

or 3) current liie i i LB)revised (e.g., 10 CFR 50.59 evaluation).

The licensee has posted on th rt e I website their operability determination process for reference (EN-A-203-1 05, No. 1 on Certrec Document Tab List).NextEra's proposal related t9 license renewal was describen a I April 14, 2011, under the response to NRC request for additional informati

.2.1. 1 g ncywide Documents Access and Management System (ADAMSc ion ML1 1108A1 31). This letter describes periodic reviews for operability as i a I oped to support the aging management review. At the time, the proposal in de a o analysis (termed "final" by NextEra) of the impact of ASR on the current e ce s n basis, including the extent of the condition, to be completed by June 20 1. ce at le r nd as noted above, the control building POD was kept open; a new im ere completed for the other five building core sample results that were i oJ ed in a xtent of conditions review. The subject NextEra letter also commits to an i r ,iluaion to be completed in March 2012. On June 29, 2011, the NRR Division ice e al issued another "Request for Additional Information" (ADAMS Accessi.o. 38) related to key aspects of NextEra's comprehensive plan for assessi AS blem for the Structures Monitoring Program including that for the F i d Containment

("Followup RAI B2.1.31-1, B2.1.31-4, and B2.1.28. The r o to thi letter dated August 11,2011, (ADAMS Accession No. ML1 2 230) es not reflec a comprehensive plan for determining operability/

ncti ity a uildings along with plans for the development of aging manageme re With res to rt irements, Region I reviewed the NextEra current Structures Monitorin od a violation of the maintenance rule for the control building.

The find n et il i NRC Inspection Report 05000443/2011002 (ADAMS Accession1 3 ). More details related to the newly discovered ASR issue were also doc d in t NRC Inspection Report 05000443/2011007 (ADAMS Accession No.ML1 1 0432) s part of a license renewal inspection.

The cover letter for the latter report notes th e agi g management review for the ASR issue is not complete and that there is a need for a continuing review in the Part 50 and 54 areas. The staff of Region I and NRR (Division of Engineering and Division of License Renewal) have been discussing actions since January 2011 to ensure that the Part 50 and 54 reviews are coordinated.

The documents listed below were made available for review on the licensee's "Certrec" internal website (Certrec Document Library Tab List). These documents reflect current NextEra view of operability' for the Control Building and the associated tunnel and penetration room. The 0ý PRYD C I SI L -OF IAýLU 0TEQ=B1 OR &C "Certrec" system was set up in order to facilitate NRC staff access to NextEra's internal documents.

The NRC staff was requested to inform Region I and NextEra if the document is to be printed, for review purposes, prior to doing so.1. C-S-1 -10159 CALC_000, Rev. 0, 'B' Electrical Tunnel Transverse Shear Evaluation Supplement to Calculation CD-20 2. C-S-1-10150 CALC_000, Rev. 0, Effects of Reduce Modulus of Elasticity

-'B ctrical Tunnel Exterior Walls , I 3. CD-20-CALC, UE Control and Diesel Generator Building Design of ten elow grade for Electrical Tunnel and the Control Building (Original Design Ca i 4. Action Request (AR) 581434 Prompt Operability Determination rete Properties Below Grade in 'B3' Electrical Tunnel Exterior Walls.On April 27, 2011, NRR Division of Engineering provid s oirt y rmi g an initial review of NextEra's basis for acceptability of the reduction in lu city in light of concrete core testing which supported 10 CFR 50.59 scre c wi ut prior NRC staff review and approval.

This evaluation and its related des c ment accept the reduced parameters of compressive strength and mods ti for the Control Building and the Containment Enclosure Building as a pot I os' the operability determination (Certrec Document Library Tab List, En s e BI n ontrol Bldg MSP -Design Change Package Description No. EC-272057, .000, o t Modulus of Elasticity Evaluation).

The staff questioned the adequacy of t sc ning ion.The licensee is also plannin a nt ause review for e maintenance rule violation noted above. Corrective actions e om ens've walkdown of all structures important-to safety with suspected c n c rda ce with a revised structures monitoring program procedure t t et th te AC st ndard in the area (ACIl349.3R-02).

This has been completed fo co b ing, cont ment enclosure building, and the containment.

Completio ft as s or the other buildings is tentatively December 2011. Further, the licensel s nduc oot cause evaluation of the ASR issue which should be completed I fo o rati into the planned March 2012 Engineering Evaluation as noted al .To ithi the limitations of their testing and analysis, NextEra determined that none of the seismi teg I structures tested.have been found to be outside their design basis and were, therefore, per ble with extent of conditions questions needing be addressed.

The Seabrook design and licen ing basis to which the licensee made these determinations was documented in UFSAR Section 3.8. NextEra is willing to address the additional questions from the NRC staff;but, it is Uncertain if those questions will be addressed in the final operability determination tentatively scheduled for September 30, 2011. It also remains uncertain what NextEra's comprehensive plan is based on review of their August 11, 2011, response to NRC letter of June 29, 2011.RE SIONA 0 ICI SE 0O JO R RN I NS T ýN G MMý In light of the newly discovered ASR issue, it appears that NextEra technical personnel are developing new insights for what key aspects must be addressed in the final operability determination for any building with evidence of ASR. NextEra is considering NRC staff questions to date and has hired consultants in this area. These consultants also will be develoDina a new model for th. Cnntainment Enclosure Buildina lond

[It should be noted that NextEra's schedule indicated in the above paragraph the prompt operability determinations were revised in mid October 201,,1 for0 Tunnel (AR 581434581434 Revision 001) and Containment Enclosure Buildinhv, Rj Vaults, EFW Pump House, and Diesel Generator Fuel Oil Tank Room R.001).]Recommended Actions bv Reaion I In order for Region I to independently determine operability important-to-safety structure affected by the ASR probli ;review for adequacy of the control building prompt o era open issues as identified by NextEra. This inform ou determination for the control building and any oth a important-to-safety structures affected by the, lem maintenance rule and are also consist nt ,r* i s we need to independently develop a c re ensi e of operability determination as a part of o oversig o e lic insights gained from NextEra's te41ic searcrA the co onýlilding or any other"asa inry case, we need a atidp and any related e applied to he final operability rtant-to-safety structures.

The r within the scope of the I cense renewal. More specifically s ues to be applied to any final ensee's process and any new Accordingly, Region I requests evaluate the adequacy of NextEra's control building prompt operability determ* its ed open issues with particular focus, but not limited to, the below listed e nic .s The licensee has provided a set of documents as noted on the "Certee " ite re nce ab've, but the NRR review should not be limited to those doc iments. gi n 'I fa ate s ring that additional documents, as needed, are available n the bsi r, a essary, b an onsite inspection.

NRR's determination should enable the st o t there is rea onable assurance of continued operability given the concrete a io d due to ASR for the control building once the final operability determi n i e extEra for this or any other important structure affected by the ASR prob \D t ur of this review, Region I requests that NRR specifically identify any concerns with ss ¶rptions, methodologies, or calculations, etc., along with the regulatory or other basis o c oncern; and, notify Region I immediately if NRR finds that any of the reviewed document the control building do not provide reasonable assurance of continued operability of that buildin .As a minimum, the response to this TIA should include an independently developed comprehensive set of issues to be addressed in the final operability determination for the Control Building in order for us to further assess the licensee's process and their new insights gained for all important-to-safety structures with evidence of ASR.4`ýP ý EC 1Q0A L- CIAL E LY 0ToL TO G 0 ýN 3.0 EVALUATION Question 1: Working with Region I staff in an inspection forum, NRR staff should identify a comprehensive list of issues that need to be addressed in the final operability determination for the Control Building, given the current view of operability by NextEra as reflected in the prontperability determination.

po,[Discussion by Region 1: NRC staff identified questions as listed in the Accession No. ML 11178A338) dated June 29, 2011. The questions rel NextEra's comprehensive plan for assessing the ASR problem for the Program, including that for the Fuel Handling Building and Conta i, 1, B2.1.31-4, and B2.1.28-3)].

If the issues are initially consideg consideration to the below additional views produced b the issues are not considered comprehensive, then identity thoadditi with consideration to those listed below along ith re r othe example would be the need for Poisson ratio calculation assumed numbers in the UFSAR or the need for s age st,, ACI standard requires it in the current sYe sof S=u Monitoring RAI B2.1.31-e, please give ical staff. If those I ues to be included asis for the concern. An, ,les because there are s because of applicable Response: (b)(5)1,--A ,REDISION OFFICI'SE 0Y 0L N N K,, p EIZPREp, SISIOiA-\OFFICIA

.,ý ONL;-(FO KERN RAIN OR &OM T) (b)(5)Question 2: Because the original design basis assumes no ASR structure, what, if any, are the specific original ASR that are not clearly evident in the UFSAR d,, design life of the ed by the presence of'I[Discussion by Region 1: For example between compressive strength and me used in the seismic analysis.

These as in the structure.]

Wethods such as the relationship shear capacity and shear force are may not be valid with ASR present 4 Response: A (b)(5)13b RED SION OFFICA USE Y INT IN lN&OM ) K (b)(5)Question 3: What is the appropriate ACl standard to be L assessing in-situ ASR degradation for the cc sampling in the future, etc)?A A used foreg core sampling 0ntrol gi ons, numbers, frequency of e on staff questioning, we need toýe of e o two applicable standards or other A C2 sed by NextEra for correlation to ther iACI 214 (version 1965 is referenced in the O that a later revision of ACI 214 (ACI-214.R-to achieve a 95% confidence level. The ACI 228!less sampling.

These standards were developed fete structures for non-nuclear applications.

to determine their relevance for nuclear application I with rebar.][Discussion by Region 1: While this is ar , know the regulatory or other basis for tlj more appropriate standard.

One penetration resistance probe datd UFSAR section 3.8.2.4).

It sho be 03) provides for additional sam 0 appears to be met by Ne llq for general design and struc Technical research y need in o in which the stru " nfc Response: (b)(5)ORED ISIONAL ON Y (,NT LB MI& N i P CS' 'L USE T EN B RA MIG 0 (b)(5)Di e p'e n adequate laboratory tests for core sampling, including appropriate para rs obtained along with laboratory test conditions?

[Discussi by Region 1: Also, during the course of this review, please identify the need for any in situ testing of control building conditions including appropriate parameters to be obtained such as temperature and humidity along with test conditions for now and in the future. Also, provide guidance on where and how much rebar should be exposed in order to assess the effect on rebar from the ASR issue.PR EI S N LSOF AL FO IT 'L B INS RI & 0 N 4,-14-No tensile strength testing is being performed on the concrete core samples and this question was raised in the RAI in terms of how shear capacity is being determined.

However, the Region I staff believe that the specific parameter of tensile strength of concrete may not be sufficiently accurate and therefore relevant in a constrained structure.

As the pressure load from the ASR gel increases, that load may be transferred to the rebar. Available research in this area appears to be conflicting.

The UFSAR for containment assume concrete in reinforced systems provide no tensile strength.

, A core sample with ASR does not represent the forces contained in the, this test, in particular, elastic rebound is not considered.

For split tensile the frictional influences in the test itself are not accommodated.

The fNI exacerbated by the standard laboratory practice of placing plywood on c tensile specimen to stop it from rolling off the test stand, thus restr'sample.]t2j K PR C1S L-O A E Y (QIkpE N B NO GC

• Question 5: Is the current NextEra structural monitoring program sufficient to discover or predict additional ASR damage to structures prior to the damage negatively impacting the design basis of the structure?

[Discussion by Region 1: To date three building assessments have been comple d: control building, the containment, and the containment enclosure building.

These assess nts were initiated as a consequence of discoveries made preparing for a renewed lich e app tio*These discoveries should be reflected in enhancements to the programs re lPl thi Maintenance Rule. The Region requests NRR assistance in evaluatin e c tabilit'of NextEra's programs to maintain the integrity of the safety related stru a.Kesponse: (b)(5)(b)(5)APD(DV1O SIOI9AN" OF hAL/" N'l N'E NNRAII SO RM

(b)(5)4.0 REGULATORY REQUIREMENTS The regulatory requirements pursuant to 10 CFR Part 50 and guidance applicable to addressing the ASR-degradation of concrete in Other Seismic Category 1 Structures at Seabrook, which includes the B Electrical Tunnel, can be found in the fol ng regulations and regulatory documents.(a) 10 CFR 50.65, Maintenance Rule, as it relates to monitorin the and condition of structures, systems, or components (SSCs) in a n ufficient to provide reasonable assurance that these SSCs are ca f i their intended functions.

When the performance or condition of an S stablished goals, appropriate corrective action shall be take (b) 10 CFR Part 50, Appendix B, as it relates to e ality as rance criteria for nuclear power plants.(c) Criterion XVI "Corrective Action" of 10 F dix B as it relates to implementing a corrective action r as e that significant conditions adverse to quality, such as fail s, u " deficiencies, deviations, defective material and equipment, an n- n an es are promptly identified, cause addressed, and corrected.(d) 10 CFR Part 50, Ap ix esign Criterion (GDC) 1 as it relates to structures, systems, d ponents being designed, fabricated, erected, and tested to quali d co surate with the importance of the safety function to be perfo5 (e) 10CF rt 5 pe ix A, GDC 2, as it relates to the design of the safety-related struc s b g a withstand the most severe natural phenomena such as wind, to d ood nd earthquakes and the appropriate combination of all loads.10 Par , Appendix A GDC 4, as it relates to safety-related structures being p protected against dynamic effects, including the effects of missiles, i vhipping, and discharging fluids, that may result from equipment failures andents and conditions outside the nuclear power unit.NUREG-0800, Standard Review Plan, Section 3.8.4 -Other Seismic Category 1"Structures (h) Regulatory Guide 1.160, Revision 2 (March 1997), Monitoring the Effectiveness of Maintenance at Nuclear Power Plants SP EC S AL- O IAL 0 Y (FO R IN 1 & T)

5.0 CONCLUSION

Based on its review of TIA 2011-013 request, available documents, literature, information obtained at the NRC inspection during the period 9/26/11 -9/30/11, and within the limitations of information available, the EMCB staff has provided reasonable technical guidance in this TIA response with regard to the issues related to the ASR degradation of concrete at Seabrook raised by Region 1 in the five questions in the TIA request. Specific technical gui ance to the issues is provided in the responses to the questions.

In order to enable staff to re a fully objective assessment, the licensee should make available to the NRC in th ' me fut its firm Action Plan and Test Plan (which should be Appendix B quality tech cu that it is implementing to comprehensively address the ASR-degradati nt-to-,safety concrete structures at Seabrook Station.

6.0 REFERENCES

Note: References 1 thru 12 are licensee documents made ailable i see's Certrec website.1. Calculation C-S-1-10159, Rev. 0, 'B' Elect verse Shear Evaluation Supplement to Calculation CD-20 2. Calculation C-S-1-10150, Rev. 0, R odulus of Elasticity-

'B' Electrical

3. Calculation CD-20-CALC, C r Ian esel Generator Building Design of Material and Walls below grade le i I and the Control Building (Original Design Calculation)

4. Drawings fop IBu rete (Electrical tunnel) 9763-F-1 11342, 9763-F-111343 and13,113 5. Action R 4, Revision 000, Prompt Operability Determination Reduced Concr ies low Grade in 'B' Electrical Tunnel Exterior Walls.6. A n et ) 581434, Revision 001, Prompt Operability Determination Reduced ies Below Grade in 'B' Electrical Tunnel Exterior Walls 5 Condition Assessment of Control Building Concrete 8. I574120 Preliminary Test Results of Control Building Concrete 9. AR 581434581434Test Results from Control Building Concrete Modulus Testing (Results of petrographic analysis of 4 of the 12 CB cores identified the presence of moderate to severe ASR in the concrete)10. EC250348, Revision 002, Condition Assessment of Building Concreteý / ECSOn -OFF,.LUSE

'Y^(F INT N "T MINCY mw 11. AR 01625775, Revision 000, Petrographic Analysis of Concrete Cores from Seabrook Station 12. System Description No. SD-66, Revision 2, System Description for Structural Design Criteria for Public Service Company of New Hampshire, Seabrook Station, Unit Nos. 1 and 2, 3/02/84.13. Structural Engineering Standard Technical Procedure 36180, Revision 0, tructural Monitoring Program," NextEra Engineering Department Standard, -201 14. Seabrook UFSAR, Revision 12, Section 3.8.4, Other Seismic C ego I 15. NUREG-0800, Standard Review Plan, Section 3.8.4- Oth m -gory I Structures

16. Regulatory Guide 1.160, Revision 2 (March 1997), nitorin erectiveness of Maintenance at Nuclear Power Plants 17. RIS 2005-20, Revision 1 dated April 16, 2 NRC Inspection Manual, Part 9900: Technical Guidance, OperabiliI De and Functionality Assessments for Resolution of Degra co rming Conditions Adverse to Quality or Safety." 18. Letter dated 6-29-2011 from Ri ard Pla e, NRC, to Mr. Paul Freeman, NextEra Energy Seabrook, LLC -Fue or Ad onal Information for the Review of Seabrook Station License Renew plitti ifically Followup to RAI B2.1.31-1 on pages 2-c 3) (ML1 1178A3380)

19. NextEra Energ er 4 to USNRC dated 8-11-2011, Docket No. 50-443, Seabrook St 'o spon to equest for Additional Information

-NextEra Energy Seabrook ' nse ew Application Request for Additional Information

-Set 15 (Specific Reon, ollow-up to RAI B.2.1.31-1 on pages 5-8)20. Ne e L r to USNRC dated 4-14-2011, Docket No. 50-443, ro tati esponse to Request for Additional Information

-NextEra Energy e Renewal Application Request for Additional Information

-Set 13 c Ily Responses to Follow-up to RAI B.2.1.31-1 and -2 on pages 4-7)S111 1310)l .Era Energy Letter SBK-L-10204 to USNRC dated 12-17-2010, Docket No. 50-443, Se brook Station Response to Request for Additional Information

-NextEra Energy Seabrook License Renewal Application Aging Management Programs (Specifically Responses to RAI B.2.1.31-1, -2 and -3 on pages 36-39) (ML1035405340)

22. ACI 318-71, Building Code Requirements for Reinforced Concrete (with Commentary)

23. ACI 349.3R-02, Evaluation of Existing Nuclear Safety-Related Concrete Structuresr CISI'- OFF IAL US NL 0 TE BRAS RNG T I..-21-24. ASTM C 823/C 823M -07, Standard Practice for Examination and Sampling of Hardened Concrete in Constructions.

25. ACI 228.IR-03, In-Place Methods to Estimate Concrete Strength 26. ACl 214R-02, Evaluation of Strength Test Results of Concrete 27. ACI 214.4R-03, Guide for Obtaining Cores and Interpreting Compre ve Sv Results 28. ACI 228.2R-98 (Reapproved 2004), Nondestructive Test Metho r luatior Concrete in Structures

29. ACI 437R-03, Strength Evaluation of Existing BuildiIpr 30. Structural effects of alkali-silica reaction -Tech*aI idance n the appraisal c existing structures, The Institution of Structural on, UK, July 199 Addendum, April 2010 _31. Report on the Diagnosis, Prognosis, vo "Alkali-Silica Reaction (AS Transportation Structures, US De of )rtation, Federal Highway Administration, January 2010 32. PCA R&D SN2892b, Eval Oont kali-Ilica Reaction (ASR) Mortar Bar Testi (ASTM C1260 and AST , rability Subcommittee

-Concrete Te Portland Cement Assocd j 09.33. Popovics, S., a ti Properties of Concrete -A Quantitative App John Wiley & o Ic,1 8.34. Nilson, a ., Design of Concrete Structures, Eleventh Edition, M Hill In 35. ark T. av, Reinforced Concrete Structures, John Wiley & Sons, 191 of of 12 and R) in ng chnolog, roach, IcGraw-75.1e-I N PRED (SIONAL 0 ICIAL U^ ONLYA (F IN RNA B NST MIN & OM T)