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, Modes M, Sheikh A Office of Nuclear Reactor Regulation, NRC Region 1
References
FOIA/PA-2012-0119
Download: ML12174A052 (23)

Text

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

Subject:

PREDECISIONAL - OFFICIAL USE ONLY: Preliminary Draft Response for Comment reg.

Alkali-Silica Reaction (ASR) Issue at Seabrook Attachments: Seabrook ASR Prelim DRAFT Response (OUO) for Comment 12-20-11.docx (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 Inthis record was deleted 1 (7 1//

the Freedom of INformation Act, exeniptiofls FOIA- -O2- \. .

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 od 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 A4I 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 ML11108A131). 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 ce e 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. *11 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 Accession NL*i 1 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 0ýPRYD CI 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 Tunnel Exterior Walls , ctrical 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 noted al I fo . o rati into the planned March 2012 Engineering Evaluation as 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 ifthose 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 INS 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 ;n;Iv.-i_*

[It should be noted that NextEra's schedule indicatedin 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 GeneratorFuel Oil Tank Room R.

001).]

Recommended Actions bv Reaion I In order for Region I to independently determine operability the co onýlilding or any other important-to-safety structure affected by the ASR probli  ; inry case, we need a "asa review for adequacy of the control building prompt o era atidp and any related open issues as identified by NextEra. This inform ou e applied to he final operability determination for the control building and any oth a rtant-to-safety structures. The important-to-safety structures affected by the, lem r within the scope of the maintenance rule and are also consist nt ,r* i s I cense renewal. More specifically we need to independently develop a c re ensi e of s ues to be applied to any final operability determination as a part of o oversig o e lic ensee's process and any new insights gained from NextEra's te41ic searcrA 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 .sThe 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 ni 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.

E LY 4`ýPý 0ToL EC 1Q0A L- TO CIALG 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,

[Discussionby Region 1: NRC staff identified questions as listed in the Accession No. ML11178A338) dated June 29, 2011. The questions rel *dt e sof NextEra's comprehensive plan for assessing the ASR problem for the S=u Monitoring Program, including that for the Fuel Handling Building and Conta i, RAI B2.1.31-1, B2.1.31-4, and B2.1.28-3)]. If the issues are initially consideg c* e, please give considerationto the below additionalviews produced b the ical staff. If those issues are not considered comprehensive, then identity thoadditi I ues to be included with considerationto those listed below along ith re r othe asis for the concern. An, example would be the need for Poisson ratio calculation ,les because there are assumed numbers in the UFSAR or the need for s age st,,s because of applicable ACI standardrequiresit in the currentlicensingb*s sY

Response

(b)(5)

--A 1,

,REDISION OFFICI'SE 0Y 0L N N K,,

SISIOiA-\OFFICIA .,ý p EIZPREp, ONL;-

(FO KERN RAIN OR &OM T)

(b)(5)

Question 2:

Because the original design basis assumes no ASR design life of the ed by the presence of structure, what, if any, are the specific original de*

ASR that are not clearly evident in the UFSAR d,, 'I

[Discussion by Region 1: For example Wethods such as the relationship between compressive strength and me shearcapacity and shear force are used in the seismic analysis. These as may not be valid with ASR present in the structure.]

4 Response: A (b)(5) 13b RED SION OFFICA USE Y INT IN lN&OM )

K

(b)(5)

A A Question 3:

What is the appropriate ACl standard to be Lused foreg core sampling assessing in-situ ASR degradation for the cc 0ntrol gi ons, numbers, frequency of sampling in the future, etc)?

[Discussion by Region 1: While this is ar e on staff questioning, we need to

, know the regulatory or other basis for tlj ýe of e o two applicable standardsor other more appropriatestandard.One stand* A C2 sed by NextEra for correlation to penetration resistanceprobe datd ther iACI 214 (version 1965 is referenced in the UFSAR section 3.8.2.4). It sho be O that a later revision of ACI 214 (ACI-214.R-

03) provides for additionalsam 0 to achieve a 95% confidence level. The ACI 228 appears to be met by Ne llq !less sampling. These standards were developed for general design and struc fete structures for non-nuclearapplications.

Technical research y need in o to determine their relevance for nuclearapplication in which the stru " nfc I with rebar.]

Response

(b)(5)

ORED ISIONAL *FICIAL ON Y

(,NT LB MI& N

i p*r P CS' Aý-*cOFF 'LUSE T EN B RA

• MIG 0 E*

(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 appropriateparameters 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 rebarshould be exposed in order to assess the effect on rebarfrom the ASR issue.

PR EI S N LSOF AL FO IT 'L B INS RI & 0 N

4, 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 parameterof tensile strength of concrete may not be sufficiently accurate and therefore relevant in a constrainedstructure. As the pressure load from the ASR gel increases,that loadmay be transferredto the rebar.Available researchin 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 representthe forces contained in the, this test, in particular,elastic rebound is not considered. For split tensile the frictionalinfluences in the test itself are not accommodated. The fNI exacerbatedby the standardlaboratorypractice 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)

(*'OAPD(DV1O N'l SIOI9AN"SO N'ENNRAII OF RMhAL/" *C(**E EhO*

(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 sb ga 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 and

• ro ents 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 1 IN & 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 is*o0 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 5Condition 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 petrographicanalysis 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 Concrete

/ ýECSOn -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 -goryI 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) (ML11178A3380)

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 SBK-L-1*1063 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 Structures

• PR rCISI'- OFF IAL US NL 0 TE BRAS RNG T

I..

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 of 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 cof existing structures, The Institution of Structural on, UK, July 19912 and Addendum, April 2010 _

31. Report on the Diagnosis, Prognosis, vo "Alkali-Silica Reaction (AS R) in Transportation Structures, US De of )rtation, Federal Highway Administration, January 2010

32. PCA R&D SN2892b, Eval Oont kali-Ilica Reaction (ASR) Mortar Bar Testing (ASTM C1260 and AST 15* , rability Subcommittee - Concrete Te chnolog, Portland Cement Assocd j 09.

33. Popovics, S., ti a Properties of Concrete - A Quantitative App roach, John Wiley & o Ic,1 8.

34. Nilson, a ., Design of Concrete Structures, Eleventh Edition, MIcGraw-Hill In

35. ark T. av, Reinforced Concrete Structures, John Wiley & Sons, 19175.

1e-I N PRED (SIONAL0 ICIAL U^ ONLYA (F IN RNA B NST MIN & OM T)