ML040430023

License Amendment No. 199, Extend Appendix J, Type a, Containment Integrated Leak Rate Test, Option B

ML040430023
Person / Time
Site:	Robinson
Issue date:	02/11/2004
From:	Chandu Patel NRC/NRR/DLPM/LPD2
To:	Moyer J Carolina Power & Light Co
Patel C P, NRR/DLPM, 415-3025
References
TAC MB9662
Download: ML040430023 (16)

Text

February 11, 2004 Mr. J. W. Moyer, Vice President Carolina Power & Light Company H. B. Robinson Steam Electric Plant Unit No. 2 3581 West Entrance Road Hartsville, South Carolina 29550

SUBJECT:

H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 - ISSUANCE OF AN AMENDMENT RE: CONTAINMENT INTEGRATED LEAK RATE TEST (TAC NO. MB9662)

Dear Mr. Moyer:

The Nuclear Regulatory Commission has issued the enclosed Amendment No. 199 to Facility Operating License No. DPR-23 for the H. B. Robinson Steam Electric Plant, Unit No. 2 (HBRSEP2). This amendment changes the HBRSEP2 Technical Specifications in response to your request dated June 11, 2003, as supplemented by letters dated August 20 and October 13, 2003.

The amendment allows the licensee to extend its Appendix J, Type A, Containment Integrated Leak Rate Test, Option B, for HBRSEP2 from the scheduled May 2004 timeframe to no later than April 9, 2007.

A copy of the related Safety Evaluation is enclosed. Notice of Issuance will be included in the Commissions bi-weekly Federal Register notice.

Sincerely,

/RA/

Chandu P. Patel, Project Manager, Section 2 Project Directorate II Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-261

Enclosures:

1. Amendment No. 199 to DPR-23

2. Safety Evaluation cc w/enclosures:

See next page

ML040430023 *See previous concurrence OFFICE PDII-2/PM PDII-2/LA PDII-2/SC OGC*

NAME CPatel EDunnington MLMarshall for: KKannler AHowe DATE 02/10/04 02/10/04 02/11/04 01/29/04 AMENDMENT NO. 199 TO FACILITY OPERATING LICENSE NO. DPR-23 H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 DISTRIBUTION:

PUBLIC EHackett PDII-2 Rdg. CPatel OGC EDunnington (Hard Copy)

G. Hill (2) AHowe J. Pulsipher TBoyce R. Palla ACRS P. Fredrickson, RII DLPM DPR cc: Robinson Service List

CAROLINA POWER & LIGHT COMPANY DOCKET NO. 50-261 H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 199 License No. DPR-23

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

A. The application for amendment by Carolina Power & Light Company (the licensee), dated June 11, 2003, as supplemented by letters dated August 20, 2003, and October 13, 2003, 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, 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 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.

2. Accordingly, the license is amended by changes to the Technical Specifications, as indicated in the attachment to this license amendment; and paragraph 3.B. of Facility Operating License No. DPR-23 is hereby amended to read as follows:

B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 199, are hereby incorporated in the license. Carolina Power &

Light Company shall operate the facility in accordance with the Technical Specifications.

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

FOR THE NUCLEAR REGULATORY COMMISSION

/RA by L Marshall for/

Allen G. Howe, Chief, Section 2 Project Directorate II Division of Licensing Project Management Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: February 11, 2004

ATTACHMENT TO LICENSE AMENDMENT NO. 199 FACILITY OPERATING LICENSE NO. DPR-23 DOCKET NO. 50-261 Replace the following page of the Appendix A Technical Specifications with the attached revised page. The revised page is identified by amendment number and contains a marginal line indicating the area of change.

Remove Page Insert Page 5.0-24 5.0-24

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 199 TO FACILITY OPERATING LICENSE NO. DPR-23 CAROLINA POWER & LIGHT COMPANY H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 DOCKET NO. 50-261

1.0 INTRODUCTION

By letter dated June 11, 2003, as supplemented by letters dated August 20, 2003, and October 13, 2003, the Carolina Power & Light Company (licensee) submitted a request for a change to H. B. Robinson Steam Electric Plant, Unit No. 2 (HBRSEP2), Technical Specification (TS) 5.5.16, Containment Leakage Rate Testing Program. The requested change would allow the licensee to extend its Appendix J, Type A, Containment Integrated Leak Rate Test (ILRT), Option B, for HBRSEP2 from the scheduled May 2004 timeframe to no later than April 9, 2007.

This Safety Evaluation addresses age-related degradation of the containment pressure boundary as it relates to the amendment request, which would provide a one-time extension of the containment ILRT interval from 12.1 years to 15 years. The proposed change is supported by a plant-specific risk assessment.

2.0 REGULATORY EVALUATION

Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix J, Option B, requires that a Type A test be conducted at a periodic interval based on historical performance of the overall containment system. HBRSEP2 TS 5.5.16 requires that containment leakage rate testing be performed as required by 10 CFR Part 50, Appendix J, Option B as modified by approved exemptions and in accordance with the guidelines contained in Regulatory Guide (RG) 1.163, Performance-Based Containment Leak-Test Program, dated September 1995, with one exception (discussed below). This RG endorses, with certain exceptions, NEI 94-01, Revision 0, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, dated July 26, 1995.

A Type A test is an overall (integrated) leakage rate test of the containment structure.

NEI 94-01 specifies an initial test interval of 48 months, but allows an extended interval of 10 years, based upon two consecutive successful tests. There is also a provision for extending the test interval an additional 15 months in certain circumstances.

The most recent two Type A tests at HBRSEP2 have been successful, so the current interval requirement would be 10 years, if not for an existing exception from the guidelines of RG 1.163.

License Amendment No. 193, issued on September 16, 2002, added this exception to TS 5.5.16, allowing a one-time extension of the Type A test interval to 12.1 years, which would require testing by May 9, 2004. The NRC staff accepted the exception primarily on the basis of probabilistic risk assessment (PRA). The licensee has reevaluated the risk basis for the previously approved license amendment and submits that a one-time interval extension to 15 years is justified.

The licensee is requesting a change to TS 5.5.16, Containment Leakage Rate Testing Program, that would modify the existing exception from the guidelines of RG 1.163 regarding the Type A test interval. Specifically, the proposed TS states that the first Type A test performed after the April 9, 1992, Type A test (the date of the latest test) shall be performed no later than April 9, 2007.

3.0 TECHNICAL EVALUATION

3.1 Inservice Inspection Program HBRSEP2 is a Westinghouse pressurized-water reactor. The containment pressure boundary structure is dry, post-tension, steel-lined concrete vertical cylinder with a hemispherical dome, containment access penetrations, process piping, and electrical penetrations. The integrity of the penetrations is verified through Type B and Type C local leak rate tests (LLRTs) as required by 10 CFR Part 50, Appendix J. The overall integrity of the containment structure is verified through an ILRT. These tests are performed to verify the essentially leaktight characteristics of the containment structure at the design-basis accident pressure. As stated in the request, HBRSEP2's last two successful Type A tests were completed on April 8, 1987, and April 9, 1992. Based on the last two successful Type A tests at HBRSEP2 and the requirements of 10 CFR Part 50, Appendix J, Option B, the current testing interval is 10 years. With the requested 5-year extension (from 10 to 15 years) of the ILRT time interval, the licensee proposed that the next overall verification of the containment leaktight integrity will be performed no later than April 9, 2007. Because the leak rate testing requirements (ILRT and LLRT) of 10 CFR Part 50, Appendix J, Option B and the containment inservice inspection (ISI) requirements mandated by 10 CFR 50.55a complement each other in ensuring the leaktightness of the pressure boundary and the structural integrity of the containment, the licensee, in its request, provided information related to the ISI of the containment and potential areas of weakness in the containment that may not be apparent in the risk assessment. The licensee also provided information to explicitly address staff questions raised during its review.

The NRC staffs evaluation of the licensees responses to these questions is discussed in the following paragraphs.

Regarding the ISI performed on the containment, CP&L stated that a Containment Inspection Program has been implemented in conformance with 10 CFR 50.55a(g)(6)(ii)(B). The Containment Inspection Program has been established, in accordance with Subsections IWE and IWL of the American Society of Mechanical Engineers (ASME) Code,Section XI, 1992 Edition through the 1992 Addenda, including the NRC-approved request for relief from certain Code requirements, to assure detection of deterioration affecting containment integrity. The first interval of the HBRSEP2 Containment Inspection Program began September 1998, and ends in September 2008. 10 CFR 50.55a(g)(6)(ii)(B) required that expedited examinations of the containment be completed by September 9, 2001. Visual examinations of the containment

structure were conducted on 100 percent of the accessible surfaces between 1998 and 2001, and were performed to meet the requirements of ASME Section XI, Subsections IWE and IWL.

These examinations consisted of a general visual examination of the accessible areas of the containment vessel liner (Pressure Boundary) for IWE and the reinforced concrete exterior (Structural Integrity) for IWL. Although the containment vessel liner between the floor and the containment vessel dome is insulated and not typically accessible, numerous sections of the insulation were removed over the last three refueling outages, which allowed VT-3 examinations of portions of the containment vessel liner. RG 1.163, Regulatory Position C.3, specifies that examinations of the accessible surfaces of the containment for detection of structural problems should be conducted prior to initializing a Type A test and during two other outages before the next Type A test if the interval for the Type A test has been extended to 10 years, in order to allow for early detection of evidence of structural deterioration. The visual examinations have been completed with no significant defects noted.

In accordance with IWE-1240, an engineering evaluation was developed to determine which containment surface areas required augmented examinations. The only component categorized as Category E-C (augmented examination) at HBRSEP2 was the equipment hatch cylinder. In 1999, during Refueling Outage (RO)-19, corrosion was observed on a small area of the bottom of the interior portion of the cylinder. The bottom portion of this equipment hatch was categorized as augmented because a majority of this portion of the cylinder is insulated and not visible. The insulation was removed to allow examination of the bottom portion of the cylinder interior in 2000, during RO-20. The evaluation of the identified areas indicated the following:

a. The maximum amount of corrosion observed on the cylinder that has a 1" nominal thickness was 1/16".

b. The maximum amount of corrosion observed on the cylinder that has a 3-1/2" nominal thickness was 5/16".

This degradation was determined by the licensee to be acceptable without repair. Based on IWE-3122.3 of Subsection IWE (the 1998 Edition), the reduction in base metal thickness of 10 percent of the nominal plate thickness in local area is acceptable without repair. Both the maximum observed values are within this threshold. Moreover, the licensee has identified the equipment hatch cylinder as area requiring augmented inspection as per IWE-1240, and it will be monitored during the subsequent inspections. Therefore, the NRC staff finds the existing local corrosion acceptable.

From the discussion above, the NRC staff finds that the licensees ISI program, including areas of augmented inspections, will provide adequate assurance that the containment structural integrity will be maintained during the extended ILRT period.

With regard to the issue related to the ISI of seals, gaskets, and examination and testing of bolts associated with the primary containment pressure boundary (Examination categories E-D and E-G), the licensee stated that the Type B and Type C tests concerning the seals, gaskets, and bolts are currently, and will be, performed in accordance with 10 CFR Part 50, Appendix J, Option A at each refueling outage as required by TS 5.5.16. Thus, the one-time extension requested for Type A testing does not affect the frequency at which the Type B and Type C tests will be performed. The NRC staff finds that the licensees ISI program for seals, gaskets, and bolted connections provides reasonable assurance that the integrity of the containment pressure boundary will be maintained.

With regard to the integrity of the two-ply stainless steel bellows, the licensee stated that the evaluation of this issue, as supported by NRC Information Notice 92-10, Inadequate Local Rate Testing, for HBRSEP2 dated May 6, 1993, determined that the Type B testing issue for two-ply stainless steel bellows was not applicable to HBRSEP2. The routine penetration sleeve testing is performed from outside of containment via test connections that are installed into the sleeve end plate such that the entire sleeve, including bellows, is tested as one unit. There have been no known occurrences of excessive leakage identified during the Type A test that had not been identified by the Type B test. In summary, this issue does not apply to the HBRSEP2 containment.

The ILRT helps to identify areas of through-wall degradation when the containment vessel is pressurized. The NRC staff requested that the licensee address how the potential leakage due to age-related degradation in the uninspectable areas (areas that cannot be visually examined) were considered in the risk assessment of the extended ILRT. This assessment, as stated by the licensee, provides a conservative evaluation of the change in likelihood of detecting liner corrosion due to extending the ILRT. The likelihood was then used to determine the resulting change in risk. The following issues were addressed:

1. Differences between the containment basemat and the containment cylinder and dome.

2. The historical line flaw likelihood due to concealed corrosion.

3. The impact on aging.

4. The liner corrosion leakage dependency on containment pressure.

5. The likelihood that visual inspections will be effective at detecting a flaw.

The licensee incorporated the potential for liner corrosion from the uninspectable side of the liner in its risk assessment using the following assumptions:

1. A half failure is assumed for basemat concealed liner corrosion due to the lack of identified failures.

2. The success data was limited to 5.5 years. Although it has been over 7 years since September 1996, when 10 CFR 50.55a started requiring visual inspection, the use of 5.5 years is considered to be a conservative assumption. Additional success data was not used to limit the aging impact of this corrosion issue, even though inspections were being performed prior to this date, and there is no evidence that liner corrosion issues were identified.

3. The liner flaw likelihood is assumed to double every 5 years. This is based solely on judgment and is included in this assessment to address the increased likelihood of corrosion as the liner ages.

4. The likelihood of the containment atmosphere reaching the outside atmosphere given a liner flaw is a function of the pressure inside the containment. Even without the liner, the containment is an excellent barrier. However, as pressure increases inside containment, cracks will form. If a crack occurs in the same region as a liner flaw, the containment atmosphere can communicate to the outside atmosphere. At

low pressures, this crack formation is extremely unlikely. Near the point of containment failure, crack formation is virtually guaranteed. Anchored points of 0.1 percent at 20 psia and 100 percent at 145 psia were selected. Intermediate failure likelihoods are determined through interpolation.

5. The likelihood of leakage (due to crack formation) in the basemat region is considered to be 10 times less likely than the containment cylinder and dome regions.

6. Nondetectible containment overpressurization failures are assumed to be large early releases. This approach avoids a detailed analysis of containment failure timing and operator recovery actions.

The assessment results show that the risk of extending the ILRT from 10 years to 15 years is small.

Based on the NRC staffs review of the information provided in the TS change request, and on the licensees responses to the five questions, the NRC staff finds that:

1. The structural integrity of the containment vessel is verified through the periodic ISI conducted as required by Subsections IWE and IWL of the ASME Code,Section XI.

2. The integrity of the penetrations and containment isolation valves are periodically verified through Type B and Type C tests as required by 10 CFR Part 50, Appendix J, Option A, and HBRSEP2 TS.

3. The potential for large leakage from the areas that cannot be examined by ISI has been explicitly modeled in performing the risk assessment.

In addition, the system pressure tests for containment pressure boundary (Appendix J tests) are required to be performed following repair and replacement activities, if any, in accordance with Article IWE-5000 of the ASME Code,Section XI. Also, significant degradation of the primary containment pressure boundary is required to be reported under 10 CFR 50.72 and 10 CFR 50.73.

3.2 Risk Assessment The licensee has performed a risk impact assessment of extending the Type A test interval to 15 years. A risk assessment was originally submitted by letter dated March 26, 2002. That assessment was used to support a one-cycle Type A test extension approved in License Amendment No. 193. The licensee re-evaluated the risk basis for the previously approved license amendment and determined that a one-time interval extension of 15 years was justified.

An updated risk assessment based on a methodology from WCAP-15691 was provided in a June 11, 2003, license amendment application for the one-time interval extension of 15 years.

In view of outstanding NRC staff concerns related to this WCAP report, the updated risk assessment was subsequently replaced in its entirety by a revised evaluation provided in a supplemental letter dated August 20, 2003. The revised risk evaluation is based on a method that has been used to support approval of a similar amendment at several other plants.

Additional analysis and information was provided by the licensee in its letter dated October 13, 2003. In performing the risk assessment, the licensee considered the guidelines of NEI 94-01, the methodology used in Electric Power Research Institute (EPRI) TR-104285, Risk Impact

Assessment of Revised Containment Leak Rate Testing, and RG 1.174, An Approach For Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis.

The basis for the current 10-year test interval is provided in Section 11.0 of NEI 94-01, Revision 0, and was established in 1995 during the development of the performance-based Option B to Appendix J. Section 11.0 of NEI 94-01 states that NUREG-1493, Performance-Based Containment Leak-Test Program, provided the technical basis to revise leakage rate testing requirements contained in Option B to Appendix J. The basis consisted of qualitative and quantitative assessments of the risk impact (in terms of increased public dose) associated with a range of extended leakage rate test intervals. To supplement this basis, industry undertook a similar study. The results of that study are documented in EPRI Research Project Report TR-104285.

The EPRI study used an analytical approach similar to that presented in NUREG-1493 for evaluating the incremental risk associated with increasing the interval for Type A tests. The Appendix J, Option A, requirements that were in effect for HBRSEP2 early in the plants life required a Type A test frequency of three tests in 10 years. The EPRI study estimated that relaxing the test frequency from three tests in 10 years to one test in 10 years would increase the average time that a leak that was detectable only by a Type A test goes undetected from 18 to 60 months. Since Type A tests only detect about 3 percent of the leaks (the rest are identified during local leak rate tests based on industry leakage rate data gathered from 1987 to 1993), this results in a 10-percent increase in the overall probability of leakage. The risk contribution of pre-existing leakage for the pressurized water reactor and boiling water reactor representative plants in the EPRI study confirmed the NUREG-1493 conclusion that a reduction in the frequency of Type A tests from three tests in 10 years to one test in 20 years leads to an imperceptible increase in risk that is on the order of 0.2 percent and a fraction of one person-rem per year in increased public dose.

Building upon the methodology of the EPRI study, the licensee assessed the change in the predicted person-rem per year frequency. The licensee quantified the risk from sequences that have the potential to result in large releases if a pre-existing leak were present. Since the Option B rulemaking was completed in 1995, the NRC staff has issued RG 1.174 on the use of PRA in evaluating risk-informed changes to a plants licensing basis. The licensee has proposed using RG 1.174 guidance to assess the acceptability of extending the Type A test interval beyond that established during the Option B rulemaking.

RG 1.174 defines very small changes in the risk-acceptance guidelines as increases in core damage frequency (CDF) less than 10-6 per year and increases in large early release frequency (LERF) less than 10-7 per year. Since the Type A test does not impact CDF, the relevant criterion is the change in LERF. The licensee has estimated the change in LERF for the proposed change and the cumulative change from the original frequency of three tests in a 10-year interval. RG 1.174 also discusses defense-in-depth and encourages the use of risk analysis techniques to help ensure and show that key principles, such as the defense-in-depth philosophy, are met. The licensee estimated the change in the conditional containment failure probability for the proposed change to demonstrate that the defense-in-depth philosophy is met.

The licensee provided analyses as discussed below. The following comparisons of risk from a change in test frequency from three tests in 10 years to one test in 15 years are considered to

be bounding for HBRSEP2 comparative frequencies of one test in 10 years to one test in 15 years. The following conclusions can be drawn from the analysis associated with extending the Type A test frequency:

1. Given the change from a three in 10-year test frequency to a one in 15-year test frequency, the increase in the total integrated plant risk is estimated to be about 0.01 person-rem per year. This increase is comparable to that estimated in NUREG-1493, where it was concluded that a reduction in the frequency of tests from three in 10 years to one in 20 years leads to an imperceptible increase in risk.

Therefore, the NRC staff finds that the increase in the total integrated plant risk for the proposed change is considered small and supportive of the proposed change.

2. The increase in LERF resulting from a change in the Type A test frequency from the original three in 10 years to one in 15 years is estimated to be 1.3 x 10-7 per year based on the internal events PRA. However, there is some likelihood that the flaws in the containment estimated as part of the Class 3b frequency would be detected as part of the IWE/IWL visual examination of the containment surfaces (as identified in ASME Boiler and Pressure Vessel Code,Section XI, Subsections IWE/IWL).

Visual inspections are expected to be effective in detecting large flaws in the visible regions of containment, and this would reduce the impact of the extended test interval on LERF. The licensees risk analysis considered the potential impact of age-related corrosion/degradation in inaccessible areas of the containment liner on the proposed change. The increase in LERF associated with corrosion events is estimated to be about 2 x 10-8 per year. The NRC staff concludes that increasing the Type A test interval to 15 years results in only a small change in LERF and is consistent with the acceptance guidelines of RG 1.174.

3. RG 1.174 also encourages the use of risk analysis techniques to help ensure and show that the proposed change is consistent with the defense-in-depth philosophy.

Consistency with the defense-in-depth philosophy is maintained if a reasonable balance is preserved between prevention of core damage, prevention of containment failure, and consequence mitigation. The licensee estimates the change in the conditional containment failure probability to be an increase of 0.6 percentage points for the cumulative change of going from a test frequency of three in 10 years to one in 15 years. The NRC staff finds that the defense-in-depth philosophy is maintained based on the small magnitude of the change in the conditional containment failure probability for the proposed amendment.

Based on these conclusions, the NRC staff finds that the increase in predicted risk due to the proposed change is consistent with the acceptance guidelines while maintaining the defense-in-depth philosophy of RG 1.174 and, therefore, is acceptable.

4.0 STATE CONSULTATION

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

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a 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 (68 FR 74264). Accordingly, the amendment meets 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.

6.0 CONCLUSION

Based on the foregoing evaluation, the NRC staff finds that the interval until the next Type A test at the H. B. Robinson Steam Electric Plant, Unit 2, may be extended to 15 years, and that the proposed change to TS Section 5.5.16 is acceptable.

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.

Principal Contributors: J. Pulsipher R. Palla Date: February 11, 2004

Mr. J. W. Moyer H. B. Robinson Steam Electric Plant, Carolina Power & Light Company Unit No. 2 cc:

Steven R. Carr Mr. C. T. Baucom Associate General Counsel - Legal Supervisor, Licensing/Regulatory Programs Department H. B. Robinson Steam Electric Plant, Progress Energy Service Company, LLC Unit No. 2 Post Office Box 1551 Carolina Power & Light Company Raleigh, North Carolina 27602-1551 3581 West Entrance Road Hartsville, South Carolina 29550 Ms. Margaret A. Force Assistant Attorney General Ms. Beverly Hall, Section Chief State of North Carolina N.C. Department of Environment Post Office Box 629 and Natural Resources Raleigh, North Carolina 27602 Division of Radiation Protection 3825 Barrett Dr.

U. S. Nuclear Regulatory Commission Raleigh, North Carolina 27609-7721 Resident Inspectors Office H. B. Robinson Steam Electric Plant Mr. Robert P. Gruber 2112 Old Camden Road Executive Director Hartsville, South Carolina 29550 Public Staff - NCUC 4326 Mail Service Center Mr. T. P. Cleary Raleigh, North Carolina 27699-4326 Plant General Manager H. B. Robinson Steam Electric Plant, Mr. Henry H. Porter, Assistant Director Unit No. 2 South Carolina Department of Health Carolina Power & Light Company Bureau of Land & Waste Management 3581 West Entrance Road 2600 Bull Street Hartsville, South Carolina 29550 Columbia, South Carolina 29201 Mr. Chris L. Burton Mr. James W. Holt Director of Site Operations Manager H. B. Robinson Steam Electric Plant, Performance Evaluation and Unit No. 2 Regulatory Affairs PEB 7 Carolina Power & Light Company Progress Energy 3581 West Entrance Road Post Office Box 1551 Hartsville, South Carolina 29550 Raleigh, North Carolina 27602-1551 Public Service Commission Mr. John H. ONeill, Jr.

State of South Carolina Shaw, Pittman, Potts, & Trowbridge Post Office Drawer 11649 2300 N Street NW.

Columbia, South Carolina 29211 Washington, DC 20037-1128 J. F. Lucas Manager - Support Services - Nuclear H. B. Robinson Steam Electric Plant, Unit No. 2 Carolina Power & Light Company 3581 West Entrance Road Hartsville, South Carolina 29550