ML14176A511

Notification of 900824 Meeting W/Util in Raleigh,Nc to Discuss Licensing Process & Submittals

ML14176A511
Person / Time
Site:	Harris, Brunswick, Robinson
Issue date:	08/17/1990
From:	Le N Office of Nuclear Reactor Regulation
To:	Adensam E Office of Nuclear Reactor Regulation
References
NUDOCS 9008220146
Download: ML14176A511 (12)

Text

August 17, 1990 Docket No. 50-400 and 261 50-325 and 50-324 MEMORANDUM FOR:

Elinor G. Adensam, Director Project Directorate II-1 Division of Reactor Projects I/II FROM:

Ngoc Le, Project Manager Project Directorate II-1 Division of Reactor Projects -

I/II

SUBJECT:

FORTHCOMING MEETING WITH CAROLINA POWER & LIGHT COMPANY DATE & TIME:

Friday, August 24, 1990 8:30 a.m.

LOCATION:

Carolina Power & Light Company Corporate Offices Raleigh, North Carolina PURPOSE:

To discuss Licensing process and submittals

• PARTICIPANTS:

NRC CP&L E. Adensam CP&L Licensing Staff P. Anderson N. Le Orignal signed by:

Ngoc Le, Project Manager Project Directorate II-1 Division of Reactor Projects -

I/II Office of Nuclear Reactor Regulation cc:

See next page

• Meetings between NRC technical staff and applicants or licensees are open for interested members of the public, petitioners, intervenors, or other parties to attend as observers pursuant to "Open Meeting Statement of NRC Staff Policy,"

43 Federal Register 28058, 6/28/78.

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OFFICIAL RECORD COPY Document Name:

MEETING CP&L 90082:2016 9000b PRADOCK 05--000261-

Carolina Power & Light Company cc:

Mr. Russell B. Starkey, Jr.

Mr. H. A. Cole Vice President Special Deputy Attorney General Brunswick Nuclear Project State of North Carolina P. 0. Box 10429 P. 0. Box 629 Southport, North Carolina 28461 Raleigh, North Carolina 27602 Mr. R. E. Jones, General Counsel Mr. Robert P. Gruber Carolina Power & Light Company Executive Director P. 0. Box 1551 Public Staff -

NCUC Raleigh, North Carolina 27602 P. 0. Box 29520 Raleigh, North Carolina 27626-0520 Ms. Frankie Rabon Board of Commissioners Mr. C. R. Dietz P. 0. Box 249 Manager, Robinson Nuclear Project Bolivia, North Carolina 28422 Department P.O. Box 790 Resident Inspector Hartsville, South Carolina 29550 U. S. Nuclear Regulatory Commission Star Route 1 Mr. Heyard G. Shealy, Chief P. 0. Box 208 Bureau of Radiological Health Southport, North Carolina 28461 South Carolina Department of Health and Environmental Control Regional Administrator, Region II 2600 Bull Street U. S. Nuclear Regulatory Commission Columbia, South Carolina 29201 101 Marietta Street, Suite 2900 Atlanta, Georgia 30323 Resident Inspector/Harris NPS c/o U. S. Nuclear Regulatory Commission Mr. Dayne H. Brown, Director Route 1, Box 3158 Division of Radiation Protection New Hill, North Carolina 27562 N. C. Department of Environmental, Commerce and Natural Resources Mr. R. B. Richey, Manager P. 0. Box 27687 Harris Nuclear Project Raleigh, North Carolina 27611-7687 Harris Nuclear Plant P.O. Box 165 Mr. J. L. Harness New Hill, North Carolina 27562 Plant General Manager Brunswick Steam Electric Plant Mr, C. S. Hinnant P. 0. Box 10429 Plant General Manager Southport, North Carolina 28461 Harris Nuclear Plant P.O. Box 165 U.S. Nuclear Regulatory Commission New Hill, North Carolina 27562 Resident Inspector's Office H. B. Robinson Steam Electric Plant Route 5, Box 413 Hartsville, South Carolina 29550 Mr. R. Morgan General Manager H. B. Robinson Steam Electric Plant P.O. Box 790 Hartsville, South Carolina 29550

DISTRIBUTION Facility: Carolina Power & Light Company Docket File NRC PDR Local PDR T. Murley 12-G-18 F. Miraglia 12-G-18 J. Partlow 12-G-18 S. Varga 14-E-4 G. Lainas 14-H-3 E. Adensam 14-8-20 P. Anderson 14-8-20 N. Le 14-B-20 OGC 15-8-18 E. Jordan MNBB-3302 B. Grimes 9-A-2 ACRS (10)

P-315 GPA/PA 17-F-2 V. Wilson 12-H-5 L. Thomas 12-E-4 B. Borchardt 17-D-19 E. Tana N. Green, Jr.

6;F REGq UNITED STATES NUCLEAR REGULATORY COMMISSION REGION II 101 MARIETTA STREET, N.W.

ATLANTA, GEORGIA 30323 August 10, 1990 NOTE TO:

Ellis W. Merschoff FROM:

Goutam Bagchi

SUBJECT:

NOTES ON H. B. ROBINSON 2 I attended the Maintenance Team Inspection (MTI) exit meeting at H. B. Robinson (HBR2) on July 13, 1990.

At that time, I had an opportunity to tour the plant also.

HBR2 is an older plant with relatively simple design of safety systems. Based on the findings of the MTI and the tour of the plant, I am left with the impression that the engineering support at the plant site should be strengthened and the plant may have unique vulnerabilities to both internally and externally initiated severe accident sequences.

The purpose of this note is to write down my thoughts for future followup.

Generic Letter (GL) 88-20 started the process of individual plant examination (IPE) for internal events. One of the requirements of the IPE is the involve ment of the plant staff. In response to GL 88-20, the utility (CP&L), in its letter dated October 31, 1989 (Enclosure 1), indicated that it will perform a PRA for HBR2 and that it plans to focus its internal expertise on IPE develop ment.

GL 88-20 is a tool for implementation of the Commission policy on severe accidents; therefore, its implementation process is not suitable for regional inspection.

However, as the licensee gains insights from its PRA review, improvements in plant operating procedures and hardware are likely to follow in the long term. The external events part of IPE is expected to be issued as a supplement to GL 88-20 by the end of calendar year 1990. For HBR2, the unique vulnerabilities for internal and external events may be due to (a) a failure of engineered safety features that are in close proximity to each other and susceptible to failure by a common cause, and (b) a lack of tornado missile protection to some main steam line equipment, respectively.

However, the contribution to core damage frequency from these initiating events may or may not be significant from the standpoint of total core damage frequency.

The results of an IPE for a plant like HBR2 would have to be reviewed carefully. This type of review will be conducted from headquarters.

There are several other issues that are being tracked by headquarters:

1) single failure criterion for electrical systems (Enclosure 2), 2) delay in implementation of control room habitability and inadequate outage planning, and 3) license extension for three years. The issue (1) above may be important for the electrical distribution system features inspection (EDSFI) that will be conducted by the DRS, and it may be worth while for the DRS to follow this issue.

The following items will be on my wish list to follow from headquarters:

(a) nondestructive examination of surface cracks in the concrete in the lower part of the HBR2 containment structure, and (2) long term degradation of the steel pile foundation for the containment structure.

aiNe

Ellis W. Merschoff 2

August 10, 1990 I understand that the MTI findings among other things will highlight the following: 1) a general lack of adequate engineering support, specifically in systems engineering, and 2) an inadequate maintenance facility at the site. A speedy and satisfactory resolution of these two issues may need some management level meetings between the Region and HBR2 plant staff.

Goutam Bagchi

Enclosures:

1. CP&L Response to GL 88-20

2. NRR Letter of 6/1/90 on Single Failure of Electrical System cc w/encls:

S. D. Ebneter L. A. Reyes C. A. Julian D. M. Verrelli L. Garner J. J. Blake E. H Girard.

R. E. Carroll

ENCLOSURE 1 Carona Power & Ught Company P.O. Box 1551

• Ralegh N.C. 27602 SERIAL:

NLS-89-290 OCT 3 1 1989 A. B CUTTER Vice President NUClear Services Deparrment,

/

United States Nuclear Regulatory Commission ATTENTION:

Document Control Desk Washington, DC 20555 H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 DOCKET NO. 50-261/LICENSE NO. DPR-23 SHEARON HARRIS NUCLEAR POWER PLANT DOCKET NO. 50-400/LICENSE NO. NPF-63 BRUNSWICK STEAM ELECTRIC PLANT, UNIT NOS.

1 AND 2 DOCKET NOS. 50-325 & 50-324/LICENSE NOS. DPR-71.& DPR-62 RESPONSE TO GENERIC LETTER 88-20 INDIVIDUAL PLANT EXAMINATIONS Gentlemen:

Carolina Power & Light Company supports the Individual Plant Examination (IPE) philosophy of allowing licensees to self-identify and resolve safety issues relevant to a specific plant. Accordingly, the Company plans to respond to the requirements of Generic Letter 88-20 in a manner consistent with the guidelines provided in NUREG-1335, Integrated within an existing Company framework that currently uses a risk-based approach to identify, evaluate, and resolve nuclear safety issues. PRA technology-based analyses have been conducted by the Company which have resulted in reduction in plant risk through modifications and procedure and training enhancements. The Company views the requirements of Generic Letter 88-20 and NUREG-1335 as a logical extension of internal programs that have been evolving since 1983.

The following discussion provides the information requested in Generic Letter 88-20, Supplement 1 dated August 29, 1989.

1. METHODOLOGY AND APPROACH The methodology employed will be Prbabilistio Risk Assessment PRA)'for each ot 9._ icensed Carolina Power.& Li ht Company nuclear units:

tI5rIPs ck Steam -e1trE 715it, Unit Nos. I and CSE?-i a' BstE?-25,

'geSThaiLon Harr7s Nuc.ear Power Plant _SHNPP),an& he i. B. Robinson thletric Plant,Unit No,HER-2), The IPEs will be based on PRAs that are in various stages of completion for the Company's nuclear units.

The SS5? Level 1 PRA has been submitted to the NRC for review.

Independent -4er reviews of the SHNPP I

7aMBeveFl T FRKE veV eeihil

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NLS-89-290 / Page 2 The front-end analysis phase of the IPEs will be conducted by using a Level 1 PRA consistent with the guidance provided in Generic Letter 88-20 and NUREG-1335. The back-end analysis phase will also be conducted consistent with the guidance provided in Generic Letter 88-20 and NUREG-1335. Any analyses of source term behavior will be based on appropriate codes including the Source Term Code Package (STCP) modified by Risk Management Associates (RMA) and the SWRSAR Code, Information from work performed in past PRAs (both NRC and Industry) will be used to the maximum extent possible.

2. DESCRIPTION Front-End Analysis:

The PRA methodology for the four units will use small functional event trees and large fault trees to model the plant response to postulated initiating events. The plant systems will be modelled using a modularized fault tree approach.

Functional sequences defined by the event trees will be quantified by linking the fault trees for the appropriate front-line and support systems with the postulated initiators and solving the resulting functional fault tree for accident sequence cutsets.

These analyses, including equipment modelling, common cause analysis, human reliability analysis, and plant damage state analysis, will be performed in a manner consistent with the guidance provided in NUREG-1335.

Back-End Analysist oi-ER

'ieni vevent "ees 11 beonstrupted to assess the performance of of'aInment for the '

nt dahiq-states defined in the front-end analysis. The oontainiment event trees will be developed to ensure sufficient detail to address important phenomenology and events that impact containment performance.

Plant-specific information and data from similar PRAs will be used, where feasible, to determine containment failure modes, containment challenges, and containment failure times.

Supporting analyses will be performed when existing data is not available or is not appropriate. The containment event trees will be quantified using thermal-hydraulic and source-term calculations, probabilistic ranges from similar PRAs, evaluations of the plant-specific characteristics of the containment, and from an understanding of the events under consideration.

The overall analysis will be performed in a manner consistent with the guidance provided in NUREG-1335.

Documentation of the assumptions, results and insights will be consistent with the intent of NUREG-1335 guidance.

Exceptions will be made for the BSEP-1 and BSEP-2 IPEs as discussed below.

BSEP DIFFERENCES The BSE? IPE submittal is expected to differ from those for SHNPP and HBR-2 based on the formal PRA submitted to the NRC on May 12, 1988 and because of additional efforts required for BWRs with Mark I containments.

Since the sub'mittal of the BSEP PRA to the NRC, the system models have been modified to provide consistency with the SHNPP and HBR-2 models and to increase quantification efficiency.

These modifications have not resulted in any significant change in the results. The original BSEP PRA

NLS-bi-290 / PageO5 submittal format will be retained to allow resources to be focused on analysis rather than rewrite. A cross-reference to NUREG-1335 documentation guidance will be provided to assist in your review of the BSEP IPE submittals. The Company is proceeding with development of the BSEP IPE in accorda Ui eve

• t The Company expects that the BSEP IPEs will require more extensive plant-specific containment structural analysis than the SHNPP and HER-2 IPEs. This will facilitate CP&L's assessment of the NRC staff proposals for Mark I containment moIifications.

Based on our current understanding of anticipated NRC guidance on other containment types, we expect to use available generic informaton for the SHNPP and HBR-2 containment evaluations.

3. IDENTIFY SCHEDULE AND MILESTONES a

The Company's approach to the IPE task will be to update and refine the existing Level I PRAs as discussed previously. This front-end analysis phase will be followed by transition to and completion of the back-end analysis phase, External events analysis will be addressed based on NRC guidance to be issued by a separate generic letter. Finally, extensive internal reviews are planned before final documentation and submittal to the NRC.

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2s on IE Oe toutstad Vssstact may be necessary to assist ng the requjred independent reviews. to this letter rovides the schedule for the major milestones of each IPE.

Dhe V el re eve A ov max m ng, t th

.98Vtoffg ighou-e ge of the Ka0sg The UN any VIIprovide additional information Wo 5hng these sbb dules and other features of the IPE project at your request.

Please refer any questions regarding this submittal to Mr. Leonard I. Loflin at (919) 546-6242.

Yours ery tru A. B. utter ABC/REM/1hr (516CRS) cc:

Mr. R. A. Becker Mr. S. D. Ebneter Mr. L. Garner (NRC -

HBR)

Mr. R. Lo Mr. D. Modeen (NUMARC)

Mr. W. H. Ruland Mr. J. E. Tedrow Mr. E. G. Tourigny

ATTACHMENT :

Milestone BSEP I & 2 SHNPP 2

Complete Front-End Analysis 12/90 3/91 Complete Back-End Analysis 12/91 9/91 Complete External Events Analysis Complete Internal Review 4/92 1/92 4 /9 Submittal to NRC 8/92 8/92 192

• To be determined based on anticipated NRC generic Letter.

taA RG,,,ENCLOSURE 2

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, 0. C. 20555 June 1, 1990 Docket No. 50-261 Mr. Lynn W. Eury Executive Vice President Power Supply Carolina Power & Light Company Post Office Box 1551 Raleigh, North Carolina 27602

SUBJECT:

SINGLE FAILURE CRITERION, ELECTRICAL SYSTEMS - H.B.

ROBINSON STEAM ELECTRIC PLANT, UNIT NO.

2 (TAC NO.

72969)

By letters dated April 19, May 19, and June 14, 1989, you responded to the NRC staff's request, under 10 CFR 50.54 (f), concerning your actions, taken or planned, regarding single electrical failure evaluation for ECCS and other safety systems. This subject was further discussed with the staff during a meeting on April 19, 1990. The purpose of this letter is to request your documentation of commitments and actions that would be responsive to the staff's request under 10 CFR 50.54(f).

By letters dated April 19 and May 19, 1989, you have committed to reevaluate the ECCS to assure that the acceptance criteria of 10 CFR 50.46, along with the single failure criterion of Appendix K, can be met. Through your discussions with the staff, we understand that this effort is in progress as scheduled and should be completed prior to the end of the refueling outage 13.

Within 60 days following the end of the outage, you are requested to submit the results of your evaluation to provide assurance that the single failure criterion of Appendix K is being met.

You were also requested to identify and correct single electrical. failure vulnerabilities per 10 CFR Part 50, Appendix A, General Design Criteria (GDC) for safety systems other than the ECCS.

In response to that request, by letter dated June 14, 1989 and during the meeting on April 19, 1990, you stated that you have embarked on two programs, the Design Basis Documentation (OB) program and the overall plant-based Level I Probabilistic Risk Assessment (PRA).

Those programs are not designed to ensure literal compliance with the Appendix A single failure criterion for electrical systems. However, you stated that you "feel strongly that the programs...

will assure that the plant meets the basic intent of the GDC, which is to ensure...

that the plant can be operated without undue risk to the health and the safety of the public."

In essence, your reference to "without undue risk" represents meeting the statutory require ments of Section 182 of the Atomic Energy Act for "adequate protection of the health and safety of the public," 42 USC 2232. The staff agrees with you in that the GDC are considered to be a codification of the requirements existing at the time Robinson was licensed (e.g.,

Section 182 of the Act) and could be used as a benchmark to assess the safety of the facility.

However, deviations from the GDC must be justified on the basis that deviations provide an equivalent

Mr. Lynn June 1, 1990 degree of safety, whether or not exemptions from the GDC are required. There fore, you are requested to commit to demonstrate that the plant can meet "an equivalent degree of safety" as afforded by meeting the GDC.

During the meeting on April 19, 1990, you discussed the DBD and the PRA programs. It is still not clear how these programs will demonstrate "an equivalent degree of safety."

For example, you stated that it was not the function of the DBD program to identify electrical single failure vulnerabilities. However, without such identification (i.e., without first identifying the deviations from the CDC), it is unclear to us how the demonstration of "an equivalent degree of safety" can be accomplished.

Therefore, you are requested to submit a description of your plan clearly identifying how the DB0 and PRA program will demonstrate equivalence, including sufficient detail on criteria and methodology. The plan description, including its schedule, should be submitted to us within 60 days of the receipt of this letter.

The reporting and/or recordkeeping requirements contairec ir this letter affect fewer than ten respondents; therefore, OMB clearance is not required under P. L.96-511.

Sincerely,

./

us C. Lainas, Assistant e

ctor for Region II Reactors Division of Reactor Projects -I/11 Office of Nuclear Reactor Regulation cc:

See next page

Mr. L. W. Eury H. B. Robinson S'team Electric Carolina Power & Light Company Plant, Unit No. 2 cc:

Mr. R. E. Jones, General Counsel Mr. Dayne H. Brown, Director Carolina Power & Light Company Department of Environmental, P. 0. Box 1551 Health and Natural Resources Raleigh, North Carolina 27602 Division of Radiation Protection P. 0. Box 27687 Raleigh, North Carolina 27611-7687 Mr. H. A. Cole Special Deputy Attorney General Mr. Robert P. Gruber State of North Carolina Executive Director P. 0. Box 629 Public Staff -

NCUC Raleigh, North Carolina 27602 P. 0. Box 29520 Raleigh, North Carolina 27626-0520 U.S. Nuclear Regulatory Commission Mr. C. R. Dietz Resident Inspector's Office Manager, Robinson Nuclear Project H. B. Robinson Steam Electric Plant Department Route 5, Box 413 H. B. Robinson Steam Electric Plant Hartsville, South Carolina 29550 P. 0. Box 790 Hartsville, South Carolina 29550 Regional Administrator, Region II U.S. Nuclear Regulatory Commission Mr. Heyward G. Shealy, Chief 101 Marietta Street Bureau of Radiological Health Suite 2900 South Carolina Department of Health Atlanta, Georgia 30323 and Environmental Control 2600 Bull Street Mr. R. Morgan Columbia, South Carolina 29201 General Manager H. B. Robinson Steam Electric Plant P. 0. Box 790 Hartsville, South Carolina 29550