ML022040265

Issuance of Amendment on Steam Generator Surveillance Requirements

ML022040265
Person / Time
Site:	South Texas
Issue date:	07/31/2002
From:	Minns J NRC/NRR/DLPM/LPD4
To:	Cottle W South Texas
Thadani M, NRR/DLPM, 415-1476
References
TAC MB3963, TS 4.4.5.3a
Download: ML022040265 (15)

Text

July 31, 2002 Mr. William T. Cottle President and Chief Executive Officer STP Nuclear Operating Company South Texas Project Electric Generating Station P. O. Box 289 Wadsworth, TX 77483

SUBJECT:

SOUTH TEXAS PROJECT, UNIT 1 - ISSUANCE OF AMENDMENT ON STEAM GENERATOR SURVEILLANCE REQUIREMENTS (TAC NO. MB3963)

Dear Mr. Cottle:

The Commission has issued the enclosed Amendment No. 140 to Facility Operating License No. NPF-76 for the South Texas Project (STP), Unit 1. The amendment consists of changes to the Technical Specifications (TSs) in response to your application dated January 28, 2002, as supplemented by letters dated June 20, July 3, and 30, 2002. In the June 20, 2002 letter, the licensee modified the request to be solely for Unit 1.

The amendment revises TS 4.4.5.3a, Steam Generator Surveillance Requirements.

Specifically, the proposed changes would revise the STP, Unit 1 TS to allow a 40-month inspection interval after its first (post-replacement) inservice inspection, rather than two consecutive inspections resulting in C-1 classification.

A copy of our Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commissions next biweekly Federal Register notice.

Sincerely,

/RA by Mohan Thadani for/

John L. Minns, Project Manager, Section 1 Project Directorate IV Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-498

Enclosures:

1. Amendment No. 140 to NPF-76

2. Safety Evaluation cc w/encls: See next page

July 31, 2002 Mr. William T. Cottle President and Chief Executive Officer STP Nuclear Operating Company South Texas Project Electric Generating Station P. O. Box 289 Wadsworth, TX 77483

SUBJECT:

SOUTH TEXAS PROJECT, UNIT 1 - ISSUANCE OF AMENDMENT ON STEAM GENERATOR SURVEILLANCE REQUIREMENTS (TAC NO. MB3963)

Dear Mr. Cottle:

The Commission has issued the enclosed Amendment No. 140 to Facility Operating License No. NPF-76 for the South Texas Project (STP), Unit 1. The amendment consists of changes to the Technical Specifications (TSs) in response to your application dated January 28, 2002, as supplemented by letters dated June 20, July 3, and 30, 2002. In the June 20, 2002 letter, the licensee modified the request to be solely for Unit 1.

The amendment revises TS 4.4.5.3a, Steam Generator Surveillance Requirements.

Specifically, the proposed changes would revise the STP, Unit 1 TS to allow a 40-month inspection interval after its first (post-replacement) inservice inspection, rather than two consecutive inspections resulting in C-1 classification.

A copy of our Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commissions next biweekly Federal Register notice.

Sincerely,

/RA by Mohan Thadani for/

John L. Minns, Project Manager, Section 1 Project Directorate IV Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-498

Enclosures:

1. Amendment No. 140 to NPF-76

2. Safety Evaluation cc w/encls: See next page DISTRIBUTION:

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ACCESSION NO: ML022040265 RidsNrrDlpmLpdiv-1(RGramm)

• SE did not change **SEE PREVIOUS CONCURRENCES OFFICE PDIV-1/PM PDIV-1/PM PDIV-1/LA DE/EMCB/SC OGC PDIV-1/SC NAME JLMinns** MThadani MMcAllister LLund* NLO** RGramm RWeisman**

DATE 07/24/02 7/31/02 7/31/02 7/23/02 07/29/02 7/31/02 OFFICIAL RECORD COPY

STP NUCLEAR OPERATING COMPANY DOCKET NO. 50-498 SOUTH TEXAS PROJECT, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 140 License No. NPF-76

1. The Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by STP Nuclear Operating Company* acting on behalf of itself and for Houston Lighting & Power Company (HL&P), the City Public Service Board of San Antonio (CPS), Central Power and Light Company (CPL), and the City of Austin, Texas (COA) (the licensees), dated January 28, 2002, as supplemented by letters dated June 20, and July 3, and 30, 2002, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, as amended, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; D. The issuance of this license amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.

• STP Nuclear Operating Company is authorized to act for Houston Lighting & Power Company (HL&P), the City Public Service Board of San Antonio, Central Power and Light Company, and the City of Austin, Texas, and has exclusive responsibility and control over the physical construction, operation, and maintenance of the facility.

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and Paragraph 2.C.(2) of Facility Operating License No. NPF-76 is hereby amended to read as follows:

2. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 140, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license. The STP Nuclear Operating Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3. The license amendment is effective as of its date of issuance and shall be implemented within 30 days from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

Robert A. Gramm, Chief, Section 1 Project Directorate IV Division of Licensing Project Management Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: July 31, 2002

ATTACHMENT TO LICENSE AMENDMENT NO. 140 FACILITY OPERATING LICENSE NO. NPF-76 DOCKET NOS. 50-498 Replace the following page of the Appendix A Technical Specifications with the attached revised page. The revised page is identified by amendment number and contain marginal lines indicating the areas of change.

REMOVE INSERT 3/4 4-14 3/4 4-14

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 140 TO FACILITY OPERATING LICENSE NO. NPF-76 STP NUCLEAR OPERATING COMPANY, ET AL.

SOUTH TEXAS PROJECT, UNIT 1 DOCKET NO. 50-498

1.0 INTRODUCTION

In a letter dated January 28, 2002, (Reference 1) as supplemented by letters dated June 20 (Reference 2), July 3, (Reference 3) and 30, (Reference 4) 2002, STP Nuclear Operating Company (the licensee), submitted a request for changes to the South Texas Project (STP),

Units 1 and 2 Technical Specifications (TSs). In the June 20, 2002 letter, the licensee modified the request to be solely applicable to Unit 1. The requested changes would revise TS 4.4.5.3a, Steam Generator Surveillance Requirements. Specifically, the proposed changes would revise the STP, Unit 1 TS to allow a 40-month inspection interval after its first (post-replacement) inservice inspection, rather than after two consecutive inspections resulting in C-1 classification.

The supplemental letters dated June 20, and July 3, and 30, 2002, provided additional information that did not change the scope of the request or the initial proposed no significant hazard consideration determination (67 FR 12607, published March 19, 2002).

2.0 REGULATORY EVALUATION

The licensee replaced the steam generators (SGs) in STP, Unit 1 during the 1RE09 refueling outage in May of 2000. The replacement SGs are Westinghouse Delta 94 design which incorporate significant design improvements, including thermally treated Alloy 690 tubing.

Since the replacement of the SGs, the licensee performed an inservice inspection during the 1RE10, fall 2001 refueling outage. No service induced degradation of the SG tubes was identified during this inspection. The licensee stated that the 1RE10 inspection results along with the improved Westinghouse replacement SG design provide the basis for the proposed TS amendment.

Currently, TS 4.4.5.3a requires that subsequent inservice inspection of SG tubes after the first post-replacement inservice inspection be performed at intervals of not less than 12 nor more than 24 calendar months after the previous inspection. In accordance with the extension criteria in TS 4.4.5.3a, If two consecutive inspections, not including the preservice inspection, result in all inspection results falling into the C-1 category or if two consecutive inspections demonstrate that previously observed degradation has not continued and no additional

degradation has occurred, the inspection interval may be extended to a maximum of once per 40 months. A C-1 category is defined as less than 5 percent of the total tubes inspected are degraded (i.e., contain defects greater than or equal to 20 percent throughwall) and none of the inspected tubes are defective (i.e., contain defects greater than or equal to 40 percent throughwall).

The proposed TS amendment requests a one time exception to the extension criteria. The licensee proposed adding the following to its TS: For Unit 1, a one-time inspection interval of a maximum of once per 40 months is allowed for the inspection performed immediately following 1RE10. This is an exception to 4.4.5.3a in that the interval extension is based on all of the results of one inspection falling into the C-1 category. Approval of this TS change request would allow the licensee to omit the inspection of the SGs during the next scheduled refueling outage, 1RE11, scheduled for spring 2003.

The licensee estimated that approval of this TS change would reduce the radiation dose by approximately 30.5 person-REM.

3.0 TECHNICAL EVALUATION

The Nuclear Regulatory Commission (NRC) staffs evaluation covers: the replacement SGs improved design features; the first inservice fall 2001 SG inspection scope; the first inservice fall 2001 SG inspection results; and related industry operating experience.

3.1 SG Design Improvements The replacement SGs incorporate both design and material improvements to address problems the industry experienced with the original SG design features. Several examples of these improvements are discussed below.

• The replacement SG tubing is made of thermally-treated Alloy 690 material which has an increased resistance to stress corrosion cracking (SCC) over the original mill annealed Alloy 600 SG tubing. The thermally-treated Alloy 690 material has a 13 percent higher chromium content and correspondingly reduced nickel content than the original mill annealed Alloy 600 SG tubing. The higher chromium content reduces the degree of sensitization of the material, thus increasing resistance to corrosion attack. In addition, heat treatment of the Alloy 690 material was optimized for SCC resistance. Extensive laboratory tests have been performed by the industry which has demonstrated that thermally-treated Alloy 690 material is superior to mill annealed Alloy 600 material in its resistance to both primary and secondary system SCC, pitting, and general corrosion.

• The replacement SGs have an advanced minimum gap U-bend support structure, wider bars, and better material (Type 405 stainless steel) for anti-vibration bars (AVBs).

These elements significantly decrease the likelihood of wear over the original SGs.

• The tube support plates (TSPs) have trefoil-shaped holes produced by broaching to reduce tube dryout and chemical concentration in the regions where the tubes pass through the TSPs. The TSP broached holes result in line contact of the tube at only three points or lands. The broached-hole, flat land TSP is designed to reduce the

tube-to-TSP crevice area, thus reducing the potential for crevice corrosion, while providing for maximum steam/water flow in the open areas adjacent to the tube.

• The replacement SG TSP material is Type 405 stainless steel which shows improved corrosion resistance over the carbon steel TSPs used in the original SGs. Corrosion resistant TSP material limits the potential for crevice corrosion product buildup, and subsequent denting and degradation of the SG tube.

• Full depth hydraulic tube expansions within the tubesheet minimize the depth of the crevice between the tubes and the top of the tubesheet. The full depth expansion minimizes the accumulation of contaminants in the tubesheet crevice and the hydraulic expansion process minimizes the residual stresses in the SG tubes. Both these improvements reduce the susceptibility of the SG tube within the tubesheet to corrosion.

• The SG is equipped with a passive sludge collector. It is fabricated as an integral part of the primary moisture separator assembly, and designed to minimize the amount of suspended particles in the secondary side recirculation flow. This reduces the accumulation of sludge at the top of the tubesheet.

• The SG design incorporated features to minimize the development of loose parts during operation and maintenance. Specific design efforts were taken to minimize corrosion potential on small thickness metal parts. In addition, the feedwater ring spray nozzle assemblies have a series of 0.29-inch outlet holes, which function to trap potential foreign objects that might otherwise be introduced into the SGs from the feedwater systems.

The NRC staff finds the replacement SGs design and material improvements should enhance the SG tubings resistance to service induced degradation of the type experienced with the original SGs, especially during the first several cycles of operation.

3.2 First Inservice Fall 2001 SG Inspection Scope The licensee stated that during the fall 2001 refueling outage (1RE10), following the first cycle of operation since the SG replacement, 100 percent of the tubing in all four SGs was inspected full-length (i.e., hot leg tube end to cold leg tube end) with an eddy current probe containing a bobbin coil. In addition, inspections were performed with a rotating pancake coil (RPC) probe containing a +Point coil of all possible indications (I-code) identified with the bobbin coil probe.

Bobbin coil results were given an I-code under the following circumstances:

• If bobbin data from manufacturing burnish marks and dents/dings exhibited change, when compared against the baseline inspection data, they were given an I-code, and

• Non-quantifiable bobbin indications were given an I-code.

The licensee utilizes the Electric Power Research Institute (EPRI) Pressurized Water Reactor (PWR) SG Examination Guidelines (Revision 5) for the purposes of implementing its SG inspection activities. The PWR SG Examination Guidelines indicate that locations for which general purpose (i.e., bobbin probes) techniques are not qualified must be examined by site-qualified techniques. Locations such as the hot leg top of the tubesheet expansion

transition region, low row U-bends, and dents and dings are examples of these types of locations. Licensees typically use rotating pancake coil (RPC) probes in these regions. STP Nuclear Operating Company elected not to use RPC probes in these regions, unless I-codes were identified, during the fall 2001 SG inspection. The licensee documented the technical justification for this decision, which was based on the improved corrosion resistance of the thermally treated alloy 690 tubing material and related industry operating experience. The licensee determined that performing RPC inspection of these regions no sooner than the end of the third operating cycle for the new SGs does not present an unacceptable challenge to tube integrity. As discussed in Section 3.4, industry operating experience supports the licensees technical justification because stress corrosion cracking is not expected to occur within the first three operations cycles. In addition, the licensees fall 2001 inspections indicate that no cracking that can be detected by the bobbin probe was identified. Accordingly, the NRC staff determined this is acceptable for the requested inspection interval extension.

In view of the above, the NRC staff concluded that the eddy current inspection scope (i.e., bobbin and RPC inspections) during the fall 2001 outage was acceptable to support the requested inspection interval extension.

3.3 First Inservice Fall 2001 SG Inspection Results The licensee stated that although a number of eddy current signals were identified, no service induced degradation of the SG tubing was identified during the fall 2001 (1RE10) refueling outage. The licensee determined that the eddy current signals were due to dings, dents, and manufacturing burnish marks. The licensee provided a discussion on the basis used to determine that these signals were acceptable to leave in service. The NRC staff did not identify any concerns.

Prior to performance of sludge lancing on the secondary side tubesheet region during the fall 2001 refueling outage, the licensee identified several foreign objects (i.e., small pieces of spiral-wound metal gasket banding). No tube wear had occurred and the sludge lancing process removed the material. The licensee also identified a possible loose part near the sixth hot leg support plate deep in the tube bundle via the bobbin coil inspection. This region could not be visually inspected due to its location within the tube bundle. Consequently, the licensee performed a bounding loose part analysis, which assumed the loose part was a spiral-wound metal gasket banding piece located at the worst SG tube location. The loose parts analysis demonstrated safe operation for the proposed extended inspection interval. In addition, the licensee inspected all surrounding tubes with an RPC probe containing a +Point coil. No wear of the surrounding tubes was identified.

The licensee performed a condition monitoring assessment to evaluate the as-found condition of the SG tubes based on eddy current inspection results. The licensee concluded that all performance criteria had been met.

The licensee also performed an operational assessment to evaluate the predicted condition of the SG tubing after the proposed extended inspection interval. They concluded that all structural and accident leakage performance criteria are predicted to be met through the end of the next two cycles of operation.

The NRC staff concluded that the inspection results, condition monitoring assessment, and operational assessment results provide assurance that unexpected degradation of SG tubing has not occurred and is not expected to occur over the proposed extended inspection interval.

3.4 Related Industry Operating Experience The licensee provided a discussion of related industry operating experience with replacement SGs in support of the proposed extended inspection interval. They cited: (1) industry experience with a similar Westinghouse Delta model replacement SG, specifically at the V.C. Summer nuclear plant, (2) experience with other Westinghouse Delta model SGs, and (3) communications with the industry on replacement SG operating experience.

V.C. Summer has Westinghouse Delta 75 SGs which have been in operation since 1994 and are a scaled down version of the STP, Unit 1 Delta 94 SGs. Two factors, which can affect the operating experience with SGs, are the average hot leg temperature and conditions of primary and secondary water chemistry. The licensee stated that the STP, Unit 1 and V.C. Summer average hot leg temperatures are only one degree apart. In addition, although each plant has its own method of chemistry control, both plants operate within the EPRI Primary and Secondary Water Chemistry Guidelines. Since initial operation with the replacement SGs, V.C. Summer has not identified any SCC. During the October 2000 inspection, V.C. Summer did identify three AVB wear signals, one at 5 percent depth and two at 9 percent depth.

However, these were also identifiable in the baseline (i.e., preservice) and previous outage inspection data and are not projected to reach plugging conditions for at least 18 additional cycles.

Two other plants have Westinghouse Delta model SGs in operation. However, they have significantly less time in-service than the V.C. Summer SGs. The licensee stated that if tube degradation or cracking were identified at one of these plants, a detailed comparison between the subject plant and STP, Unit 1 would be performed to determine susceptibility for similar degradation. If extended operation based on this evaluation could not be justified, then a SG inspection would be performed at the next scheduled refueling outage.

NRC staff is cognizant of the fact that no plant with thermally treated alloy 690 SG tubing has reported the identification of any degradation mechanism, other than mechanical wear. The licensee for STP, Unit 1 stated that its SG engineering organization maintains close communications with other utility representatives through direct participation on EPRI SG management program member committees. If degradation is discovered at another plant with thermally treated alloy 690 tubing, this information would be quickly communicated to the licensee for STP, Unit 1. They would perform an operational assessment to determine STP, Unit 1s susceptibility to this degradation and determine whether extended operation, without an inspection, was warranted.

In view of the above, the NRC staff concluded that the industry operating experience with replacement SGs supports the licensees proposed extended inspection interval.

4.0 ADDITIONAL COMMENTS In the letter dated July 30, 2002, the licensee provided four technical justifications in support of exceptions it took to the EPRI SG Examination Guidelines. The fourth exception (F.4) was

reviewed and is discussed in Section 3.2 of this safety evaluation. The NRC staff did not review the remaining justifications in detail as they were not related to its review of the licensees proposed license amendment request. However, the NRC staff did note that the first exception, F.1, dealt with the use of qualified inspection techniques for the tube within the tubesheet below the expansion transition region located at the top of the tubesheet. Although the licensee for STP, Unit 1 did not utilize this exception during its first inservice inspection, the NRC staff would like to note that during the spring 2002 refueling outage season, the NRC staff became aware of similar exceptions being taken by other licensees, and have raised concerns with this exception. NRC is considering issuing a generic communication to the industry on this topic in the near-term.

5.0

SUMMARY

The NRC staff evaluated the replacement SGs improved design features, the scope and results of the first inservice SG inspection, and related industry operating experience as part of its review of the proposal.

The NRC staff concluded that the replacement SGs incorporate both design and material improvements that are expected to improve the SGs tubing resistance to all forms of service induced degradation, especially during the first several cycles of operation. In addition, the fall 2001 inspection scope, the results of the inspection, and the conclusions of the operational assessment indicate that the tubing is not experiencing any service induced degradation and can be safely operated during the proposed extension. Lastly, the industry operating experience with both the thermally treated alloy 690 tubing and the improved Westinghouse Delta design provides added assurance that the SGs can be safely operated over the proposed period of operation without an inspection of the SG tubing. Therefore, the NRC staff has determined that the proposed technical specifications changes are acceptable.

6.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Texas State official was notified of the proposed issuance of the amendment. The State official had no comments.

7.0 ENVIRONMENTAL CONSIDERATION

The amendments change an inspection or surveillance requirement. The NRC staff has determined that the amendment involves no significant increase in the amounts and no significant change in the types of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding (67 FR 12607, published March 19, 2002). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b),

no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

8.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

9.0 REFERENCES

1. Letter from J. J. Sheppard (STPNOC) to NRC Document Control Desk, License Amendment Request - Proposed Amendment to Technical Specification 4.4.5.3a, dated January 28, 2002 (NOC-AE-02001245).

2. Letter from T.J. Jordan (STPNOC) to NRC Document Control Desk, Revised Proposed Amendment to Technical Specification 4.4.5.3a. dated June 20, 2002 (NOC-AE-02001351).

3. Letter from T. J. Jordan (STPNOC) to NRC Document Control Desk, Response to Request for Additional Information, dated July 3, 2002 (NOC-AE-02001355).

4. Letter from S. M. Head (STPNOC) to NRC Document Control Desk, Submittal of Information Previously Sent by Facsimile, dated July 30, 2002 (NOC-AE-02001375).

Principal Contributors: April Smith, Cheryl Khan Date: July 31, 2002

South Texas, Units 1 & 2 cc:

Mr. Cornelius F. OKeefe A. H. Gutterman, Esq.

Senior Resident Inspector Morgan, Lewis & Bockius U.S. Nuclear Regulatory Commission 1111 Pennsylvania Avenue, NW P. O. Box 910 Washington, DC 20004 Bay City, TX 77414 Mr. T. Jordan, Vice President A. Ramirez/C. M. Canady Engineering & Technical Services City of Austin STP Nuclear Operating Company Electric Utility Department P. O. Box 289 721 Barton Springs Road Wadsworth, TX 77483 Austin, TX 78704 S. M. Head, Manager, Licensing Mr. M. T. Hardt Nuclear Quality & Licensing Department Mr. W. C. Gunst STP Nuclear Operating Company City Public Service Board P. O. Box 289, Mail Code: N5014 P. O. Box 1771 Wadsworth, TX 77483 San Antonio, TX 78296 Environmental and Natural Resources Mr. C. A. Johnson/R. P. Powers Policy Director AEP - Central Power and Light Company P. O. Box 12428 P. O. Box 289 Austin, TX 78711-3189 Mail Code: N5022 Wadsworth, TX 77483 Jon C. Wood Matthews & Branscomb INPO 112 East Pecan, Suite 1100 Records Center San Antonio, TX 78205 700 Galleria Parkway Atlanta, GA 30339-3064 Arthur C. Tate, Director Division of Compliance & Inspection Regional Administrator, Region IV Bureau of Radiation Control U.S. Nuclear Regulatory Commission Texas Department of Health 611 Ryan Plaza Drive, Suite 400 1100 West 49th Street Arlington, TX 76011 Austin, TX 78756 D. G. Tees/R. L. Balcom Brian Almon Houston Lighting & Power Co. Public Utility Commission P. O. Box 1700 William B. Travis Building Houston, TX 77251 P. O. Box 13326 1701 North Congress Avenue Judge, Matagorda County Austin, TX 78701-3326 Matagorda County Courthouse 1700 Seventh Street Bay City, TX 77414 July 2002

Susan M. Jablonski Office of Permitting, Remediation and Registration Texas Natural Resource Conservation Commission MC-122 P.O. Box 13087 Austin, TX 78711-3087 G. R. Bynog, Program Manager/

Chief Inspector Texas Department of Licensing and Regulation Boiler Division P. O. Box 12157, Capitol Station Austin, TX 78711