B12193, Forwards Reviews for Listed Isap Topics,Including Topics 1.08, Seimic Mods to RCS & 1.13, Inadequate Core Cooling Instrumentation, Per 850517 Commitment

Forwards Reviews for Listed Isap Topics,Including Topics 1.08, Seimic Mods to RCS & 1.13, Inadequate Core Cooling Instrumentation, Per 850517 Commitment

ML20215J414
Person / Time
Site:	Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date:	10/16/1986
From:	Opeka J CONNECTICUT YANKEE ATOMIC POWER CO.
To:	Charemagne Grimes Office of Nuclear Reactor Regulation
References
B12193, NUDOCS 8610240433
Download: ML20215J414 (12)

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CONNECTICUT YANKEE ATOMIC POWER COMPANY B E R L I N, CONNECTICUT P o BOX 270 HARTFORD. CONNECTICUT 06141-0270 TELErwoNE 2 "'"

October 16,1986 Docket No. 50-213 B12193 Office of Nuclear Reactor Regulation Attn: Mr. Christopher I. Grimes, Director Integrated Safety Assessment Project Directorate Division of PWR Licensing - B U.S. Nuclear Regulatory Commission Washington, D.C. 20555

References:

(1) 3. F. Opeka letter to C.1. Grimes, dated May 17,1985.

(2) H. L. Thompson letter to J. F. Opeka, dated July 31,1985.

Gentlemen:

Haddam Neck Plant Integrated Safety Assessment Program In Reference (1), Connecticut Yankee Atomic Power Company (CYAPCO) provided a proposed scope for the Integrated Safety Assessment Program (ISAP) review of the Haddam Neck I-hnt. In Reference (2), the Staff formally issued the results of the ISAP screening review process, establishing the scope of ISAP for Haddam Neck and initiating issue-specific evaluations. Reference (1) also indicated that for each issue or topic included in ISAP, CYAPCO would provide a discussion of the safety objective and an evaluation of the plant design with respect to the issue being addressed to identify specific items to be considered in the integrated assessment. In accordance with this commitment, reviews for the following ISAP topics are attached:

1) ISAP Topic No. 1.08 " Seismic Modifications to Reactor Coolant System"

2) ISAP Topic No.1.13 " Inadequate Core Cooling Instrumentation"

3) ISAP Topic No.1.17 " Replacement of Motor Operated Valves"

4) ISAP Topic No.1.43 " Seismic Qualification of Equipment" If you have ar.y questions concerning the attached reviews, please contact us.

Very truly yours, CONNECTICUT YANKEE ATOMIC POWER COMPANY b u$t~

3. F.Wka L Senior Vice President 8610240433 861016 1 PDR ADOCK 05000213 \

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Docket No. 50-213 B12193 Haddam Neck ISAP Topic No.1.08 Seismic Modifications to Reactor Coolant System October,1986

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Haddam Neck ISAP Topic No.1.08 Seismic Modifications to Reactor Coolant System

1. Introduction As part of SEP Topic III-6 (Seismic Design Considerations) the reactor coolant system (RCS) at Haddam Neck was reanalyzed to determine its ability to withstand deadweight, pressure, thermal, and seismic loadings.

The complete system, including reactor vessel, piping, steam generators, reactor coolant pumps, and pressurizer, was mathematically modelled by Westinghouse. Static analyses were performed for deadweight, pressure and thermal loading cases, while dynamic analyses were performed for the seismic load case. Ground response spectra, developed under this same SEP topic, were used as input to develop the seismic loadings.

II. Review Criteria

1. SEP Topic 111-6

2. Standard Review Plan

3. USI-40 " Seismic Design Criteria"

4. ASME B&PV Code Section III, Subsection NF III. Related Projects / Interfaces

1. ISAP Topic No.1.04 " Seismic Qualification of Safety-Related Piping"

2. ISAP Topic No.1.05 " Seismic Structural Modifications"

3. ISAP Topic No.1.08 " Seismic Qualification of' Equipment" IV. Evaluation The results of the Westinghouse analyses indicate there are RCS supports that may be stressed beyond code allowables. They are:

(a) Steam generator upper hold-down bolts, (b) Reactor coolant pump spring hangers, ,

(c) Pressurizer lateral support, and (d) Surge line support.

Based upon Westinghouse reports received to date, the indications were that these four items should be further analyzed on a plant specific basis, and if necessary, modified to bring the RCS to within code allowables.

These analyses compared calculated loads to conservatively established code values and have shown acceptable margins for all equipment, except the steam generator bolts and the pressurizer lateral truss. Hence, further analytical work is necessary and will be completed in early 1937. See Reference 17.

V. Conclusions CYAPCO has proposed physical modifications to the RCS as specified above to meet the standards set forth by the seismic reevaluation plan.

This project will be further evaluated in the integrated assessment.

VI. References

1. D. G. Eisenhut letter to W. G. Counsit, "Haddam Neck Plant Systematic Evaluation Program", dated August 4,1980.

2. D. M. Crutchfield letter to W. G. Counsil, "Haddam Neck - SEP Topic 111-6, ' Seismic Design Considerations,"' dated April 8,1981.

3. W. G. Counsil letter to D. M. Crutchfield, "Haddam Neck Plant, Systematic Evaluation Program - Seismic Reevaluation," dated Aug-ust 5,1980.

4. W. G. Counsil letter to D. M. Crutchfield, "Haddam Neck Plant -

Seismic Reevaluation Program," dated September 15,1980.

5. 'W. G. Counsil letter to D. M. Crutchfield, "Haddam Neck Plant, SEP-Anchorage and Support of Safety - Related Electrical Equip-ment," dated December 30,1980.

6. W. G. Counsil letter to D. L. Ziemann, "Haddam Neck Plant, SEP Evaluation Program - Seismic Reevaluation," dated January 17, 1980.

7. W. G. Counsil letter to D. M. Crutchfield, "Haddam Neck Plant -

Seismic Reevaluation Program, SEP Topic Ill Seismic Design Considerations," dated June 11, 1981.

8. D. M. Crutchfield letter to W. G. Counsil, "Haddam - Neck - SEP Topic Ill-6, 'Scismic Design Considerations,"' dated September 28, 1981.

9. W. G. Counsit letter to D. M. Crutchfield, "Haddam Neck Plant -

Seismic Reevaluation Program, SEP Topic III Seismic Design Considerations", dated November 16,1981.

10. W.G. Counsil letter to D.M. Crutchfield, "Haddam Neck Plant - SEP Topic III-6, Seismic Design Considerations," dated January 8,1982.

11. D.M. Crutchfield letter to W.G. Counsil, "SEP Topic III-6, Seismic Design Considerations, Staff Guidelines for Seismic Evaluation Cri-teria for the SEP Group 11 Plants," dated September 21,1982.

12. -W.G. Counsil letter to D.M. Crutchfield, "Haddam Neck Plant, SEP Topic III-6, Seismic Design Considerations," dated April 29,1983.

13. W.G. Counsit letter to D.M. Crutchfield, "Haddam Neck Plant, SEP Topic III-6, Seismic Design Considerations," dated January 19,1984.

14. W.G. Counsil letter to D.M. Crutchfield, "Haddam Neck Plant, SEP Topic III-6, Seismic Design Considerations," dated May 2,1984.

15. 3.F. Opeka letter to 3.A. Zwolinski, "Haddam Neck Plant, SEP Topic III-6, Seismic Design Considerations," dated June 27,1985.

16. 3.F. Opeka letter to C.I. Grimes, "Haddam Neck Plant, SEP Topic III-6, Seismic Design Considerations, Clarification to the Criteria Document for Safety-Related Piping," dated April 29,1986.

17. 3.F. Opeka letter to C.I. Grimes, "Haddam Neck Plant - Reactor Coolant System Seismic Reevaluation," dated August 26,1986.

Docket No. 50-213 B12193 Haddam Neck ISAP Topic No.1.13 Inadequate Core Cooling Instrumentation October 1936

Haddam Neck ISAP Topic No.1.13 Inadequate Core Cooling Instrumentation I. - Introduction Instrumentation such as a reactor vessel level monitoring system (RVLMS) can serve as an additional tool for the operator to detect conditions of inadequate core cooling. In particular, a RVLMS could potentially be important in mitigating small break LOCA and steam generator tube ruptures, which are significant contributors to core melt and hence, public risk. Another potential use of the RVLMS would be to avert unnecessary reactor vessel head venting and possible containment flooding.

II. Review Criteria

1. NUREG-0737, Item II.F.2 " Inadequate Core Cooling Instrumentation"

2. IEEE No. 323-74 and No. 344-75.

IIL Related Projects / Interfaces ISAP Topic No.1.21 " Regulatory Guide. l.97 Instrumentation" IV. Evaluation A RVLMS.using heated junction thermocouples was installed at the Haddam Neck plant during the 1986 refueling outage. The core exit thermocouples and the subcooled margin monitor were upgraded to meet all of the qualification requirements (per Reference 1).

V. Conclusions In Reference 1, the Staff concluded that the inadequate core cooling system design is acceptable and is consistent with NRC criteria contained in NUREG-0737 Item ILF.2. Hence, CYAPCO considers this ISAP topic closed.

VI. Reference (1) F. M. Akstulewicz letter to 3. F. Opeka, "TMI Action Plan Item II.F.2 -

Inadequate Core Cooling Instrumentation," dated December 12,1985.

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Docket No. 50-213 B12193 Haddam Neck ISAP Topic No.1.17 Replacement of Motor Operator Valves October 1936

Haddam Neck ISAP Topic No.1.17 Replacement of Motor-Operated Valves (MOVs)

' I. Introduction In order to comply with the requirements of 10 CFR 50.49 " Environmental

. Qualifications of Safety Related Electrical Equipment for Nuclear Power Plants," the remaining fourteen motor operated valves in the residual heat removal (RHR) system, letdown isolation, auxiliary pressurizer spray, PORY block valve, core deluge and cold leg injection systems have to be replaced with qualified Limitorque operators.

IL Review Criteria 10 CFR 50.49 III. Related Topics / Interfaces None.

IV. Evaluation Proper qualification of all 14 motor operators has been achieved by replacing the existing Crane /Teledyne MOVs with qualified Limitorque operators.

V. Conclusions The modifications noted_ above have recently been completed during the 1986 refueling outage. As such, CYAPCO considers this ISAP topic closed.

VI. References

1. H. R. Denton letter to W. G. Counsil, " Extension Request for Time Granted for Final EEQ," dated April 5,1984.

2. W. G. Counsil letter to H. R. Denton, " Extension of EEQ Deadline for Electrical Equipment," dated February 28,1985.

3. H. R. Denton- letter to W. G. Counsil, " Schedular Exter.sion for Equipment Qualifications," dated March 28,1985.

4. H. L. Thompson, Jr. letter to 3. F. Opeka, "! SAP Topic No.1.17, EEQ of Motor Operated Valves," dated July 31,1985.

5. 3. F. Opeka letter to N. 3. Palladino, "Haddam Neck Plant -

Environmental Qualification of Electrical Equipment Schedular Extension Requests," dated September 30,1985.

6. 3. F. Opeka letter to 3. A. Zwolinski, " Environmental Qualification of Electrical Equipment Justifications for Continued Operation," dated November 27,1985.-

7. S. 3. Chilk memorandum to 3. F. Opeka, " Extension for the Environmental Qualification of Electrical Equipment at Connecticut Yankee," dated November 20,1985.

Docket No. 50-213 B12193 Haddam Neck ISAP Topic No.1.48 Seismic Qualification of Equipment October,1986

Haddam Neck ISAP Topic No.1.48 Seismic Qualification of Equipment (USI A-46)

1. Introduction Unresolved Safety Issue (USI) A-46 concerns the seismic qualification of safe shutdown equipment at operating nuclear power stations. The safety issue in USI A-46 is the concern that the margins of safety provided by equipment in operating nuclear power plants under seismically induced loads may vary considerably. Significant changes in design criteria and methods for seismic qualification of equipment have been adopted since these plants were reviewed for their operating licenses.

In cooperation with the NRC, the Seismic Qualification Utility Group (SQUG) has investigated and developed methods and acceptance criteria to assess the seismic capability of mechanical and clectrical equipment. The main focus of the SQUG has been in the accumulation and evaluation of seismic experience data from many strong motion earthquake sites, in order to determine the feasibility of applying experience data to nuclear plant equipment, a pilot study of eight classes of equipment was under-taken. This program concluded that experience data could be used to assess the seismic capability of equipment and that the eight selected equipment classes were seismically adequate.

Two major concerns were raised by the Senior Seismic Review and Advisory Panel (a panel of experts selected by SQUG and the NRC). First, eculpment must be properly anchored. Secondly, a large number of protective relays do not exhibit sufficient seismic ruggedness to assure operability; particularly for earthquakes which exceed 0.3g peak ground acceleration.

II. Review Criteria

1. NUREG-1030, " Seismic Qualification of Equipment in Operating Nuclear Power Plants"

2. NUREG-0933,"A Prioritization of Generic Safety Issues" 111. Related Topics / Interfaces y 1. ISAP Topic No. 1.04 " Seismic Qualification of Safety-Related i Piping"

2. ISAP Topic No.1.05 " Seismic Structural Modifications"

3. ISAP Topic No.1.08 "Scismic Modifications to Reactor Coolant System"

4. ISAP Topic No.1.09 " Design Codes, Design Criteria, Load Combina-tions"

5. ISAP Topic No.1.50 " Fracture Toughness Supports" IV. Evaluation With the pilot program complete, two positions on the status of USI A-46 developed. Within the owners group and other industry groups, many felt that the eight classes of equipment studied showed the seismic qualifi-cation of equipment not to be a significant safety concern. Further, based

on the backfit rule, cost benefit analyses would not support the seismic i qualification of existing plant equipment.

The second position which developed was endorsed by the NRC. This position was that the experience approach is an acceptable means of establishing the seismic capability of equipment and that the eight classes should be extended to include all safe shutdown equipment classes.

Required for this approach would be an equipment screening criteria for plant walkdowns, anchorage _ review and inspection criteria and systems analyses to assess the effects of relay chatter. At the present time, the latter position is prevailing.

Two primary areas of concern were in relation to anchorage and support of safety-related electrical equipment and the seismic capabilities of cable trays and conduit. To address the first concern, a list of all safety-related electrical equipment was compiled, a survey was performed to determine what anchorage existed, and the adequacy of the. existing anchorage was evaluated. Examination of these specified items revealed that most equipment had positive anchorage. Appropriate modifications were then undertaken and completed to ensure the seismic adequacy of the remaining equipment. The second concern, seismic qualification of cable trays and conduit, was evaluated generically by the SEP Owners', Group and a determin . tion was made that the seismic capabilities were adequate to ensure their design basis function under seismic conditions.

V. Conclusions Building upon the foundation summarized above, CYAPCO is in the process of completing further seismic analyses on plant equipment at rhe Haddam Neck Plant. As such, further evaluation will be undertaken in the integrated assessment.

VI. References

1. D.G. Eisenhut letter to W.G. Counsil, "SEP Topic III-6," dated January 1,1980.

2. W.G. Counsil letter to D.L. Ziemann, "Haddam Neck Plant - SEP -

Anchorage and Support of Safety-Related Electrical Equipment,"

dated February 14, 1980.

3. W.G. Counsil letter to D.L. Ziemann, "Haddam Neck Plant - SEP -

Anchorage and Support of Safety-Related Electrical Equipment,"

dated March 14,1980.

4. W.G. Counsit letter to D.M. Crutchfield, "Haddam Neck Plant - SEP -

Anchorage and Support of Safety-Related Electrical Equipment,"

dated November 21,1980.

5. R.M. Kacich letter to C.I. Grimes, "SEP Topic III-6, Seismic Design Considerations - SEP Owners Group Cable Tray / Conduit Test Pro-gram," dated October 15,1984.