Press Release-95-020, Note to Editors on Reports Received from ACRS

Press Release-95-020, Note to Editors on Reports Received from ACRS

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Issue date:	02/23/1995
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Press Release-95-020
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No.

95-20 FOR IMMEDIATE RELEASE Tel. 301/415-8200 (Thursday, February 23, 1995)

NOTE TO EDITORS:

The Nuclear Regulatory Commission has received three reports from its independent Advisory Committee on Reactor Safeguards.

The attached reports sent to the NRC's Executive Director for Operations, in the form of letters, comment on:

1)

A postulated reactor water cleanup system line break for operating boiling water reactors.

2)

Proposed final Revision 3 to Regulatory Guide 1.118, "Periodic Testing of Electric Power and Protection Systems."

3)

A proposed final amendment to NRC's Part 50 regulation incorporating two subsections of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code.

Attachments:

As stated

February 15, 1995 Mr. James M. Taylor Executive Director for Operations U. S. Nuclear Regulatory Commission Washington, D.C.

20555-0001

SUBJECT:

REACTOR WATER CLEANUP SYSTEM LINE BREAK FOR OPERATING BWRS During the 418th meeting of the Advisory Committee on Reactor Safeguards, February 9-10, 1995, we held discussions with representatives of the NRC staff concerning Issue 3 [Reactor Water Cleanup (RWCU) Systems Safety] from our letter to you dated July 13, 1994 (Reference 1).

In our letter, we pointed out that an added RWCU isolation valve inside primary containment provides long-term post-accident isolation of the ABWR if the primary containment isolation valves fail to close fully under blowdown conditions resulting from an RWCU line break outside of primary containment.

We suggested that operating plants may not have a similar capability and recommended that this issue be investigated.

In your September 9, 1994 response (Reference 2), you stated that the staff will perform a study to determine whether the environmental conditions in secondary containment resulting from an RWCU line break would create an environment bounded by the current analyses for operating plants.

We discussed this response with the NRC staff members.

They assured us that the environmental conditions would include those associated with the postulated event described below.

For this event, a pipe break is postulated in the safety or non-safety portion of the RWCU system outside of primary containment.

A blowdown of reactor coolant and steam to the break occurs until the break is isolated by containment isolation valves.

If these valves are unable to close completely due to the severity of simultaneous mechanical and electrical demands on both valves under blowdown flow conditions, the reactor will continue to discharge a portion of its coolant and steam inventory to the break indefinitely.

It is likely that several remotely operated relief valves on the reactor steam lines will be opened to divert a portion of the steam directly to the suppression pool.

However, for a typical BWR-4 (and perhaps for many other BWRs) these relief valves will close at about 50 psig even if they are externally actuated to open.

The valves will not reopen until the reactor repressurizes to about 85 psig.

If the ECCS pumps are operating, the water flowing into the reactor vessel may increase the vessel pressure sufficiently to lift and hold open the remotely operated relief valves.

This should ensure adequate core cooling while the pumps are running, but a significant portion of the ECCS flow will be diverted to the unisolated break thereby depleting the water inventory needed to ensure proper pump operation during long-term core cooling.

In addition, the diverted water will be released inside of secondary containment where it can gravitate to the lowest level where the ECCS pumps and drivers are located.

The resulting water cascading and flooding may jeopardize the continued availability of the ECCS pumps and equipment during long-term core cooling.

If adequate ECCS flow is not maintained, core uncovery to below the level of the jet pump throat (2/3 core level) is a certainty.

(The reactor coolant loss will be greater if the reactor vessel bottom drain line is open and cannot be closed.)

If the ECCS pumps are not operating, the relief valves will cycle in the 50-85 psig range.

Still, a portion of the reactor coolant will be diverted to the break.

Eventually, the fuel decay heat will be insufficient to repressurize the reactor to 85 psig.

Thereafter, the relief valves will remain closed and any ECCS flow and resulting steam will be directed to the break.

Various corrective actions or features might be considered to mitigate this event, but most have shortcomings.

For example, one could provide remotely operated relief valves which can be kept open during the event.

Since the relief valves exhaust to the suppression pool, the reactor pressure must be sufficient to overcome the drywell pressure and the pressure equivalent of the relief valve sparger submersion depth.

Although dependent on the piping arrangement to the break, the reactor pressure may be sufficient to direct most ECCS water and steam from the core to the break.

Provisions for relieving directly to the containment atmosphere could overcome this problem only if the containment is maintained at essentially the same pressure as at the break location and if the piping arrangement to the break is not conducive to siphoning.

Opening the main steam lines to a functional main condenser (if operating at partial vacuum) might be a solution if it were possible to arrange when subject to the human and equipment limitations created by the break and harsh environment in primary and secondary containment.

Other solutions may be proposed.

We believe that the primary containment isolation valves for the RWCU system must be able to perform their safety function while subjected to the conditions present when the valves are required to operate.

We agree that the ability of these valves to perform

their design function was considered in the resolution of Generic Issue 87, "HPCI Steam Line Break Without Isolation."

We also agree that the implementation of Generic Letter No. 89-10, "Safety-Related Motor-Operated Valve Testing and Surveillance,"

should improve the likelihood of proper valve functioning under design-basis conditions.

We are concerned, however, that sufficient test data under actual blowdown flow conditions and realistic geometries are not available to validate the valve reliability used in current probabilistic risk assessments.

We are concerned that the risk associated with an RWCU pipe break outside of primary containment has been underestimated and that a need may exist for additional isolation capability in the RWCU line inside of primary containment.

We look forward to seeing the results of the current investigations.

We recommend that similar studies be undertaken of the risk significance of failure to isolate high energy line breaks outside primary containment in the High Pressure Coolant Injection and Reactor Core Isolation Cooling Systems.

Sincerely, T. S. Kress Chairman, ACRS

References:

1.

Letter dated July 13, 1994, from T. S. Kress, ACRS Chairman, to James M. Taylor, NRC Executive Director for Operations,

Subject:

Some Areas for Potential Staff Consideration for Operating Nuclear Power Plants and the Review of Future Plant Designs Resulting from the ACRS Review of the Evolutionary Light Water Reactors 2.

Memorandum dated September 9, 1994, from James M. Taylor, NRC Executive Director for Operations, to T. S. Kress, ACRS Chairman,

Subject:

Some Areas for Potential Staff Consideration for Operating Nuclear Power Plants and the Review of Future Plant Designs Resulting from the ACRS Review of the Evolutionary Light Water Reactors Mr. James M. Taylor Executive Director for Operations U.S. Nuclear Regulatory Commission Washington, D.C.

20555-0001

SUBJECT:

PROPOSED FINAL REVISION 3 TO REGULATORY GUIDE 1.118, "PERIODIC TESTING OF ELECTRIC POWER AND PROTECTION SYSTEMS"

During the 418th meeting of the Advisory Committee on Reactor Safeguards, February 9-10, 1995, we reviewed the subject proposed revision to Regulatory Guide 1.118 that provides guidance for implementing some of the requirements of 10 CFR 50.55a(h),

"Protection Systems"; 10 CFR Part 50 Appendix A, General Design Criterion (GDC) 18, "Inspection and Testing of Electric Power Systems," and GDC 21, "Protection System Reliability and Testability"; and 10 CFR Part 50 Appendix B, Criterion XI, "Test Control."

During our review, we had the benefit of discussions with representatives of the NRC staff and the Institute of Electrical and Electronics Engineers (IEEE).

We also had the benefit of the documents referenced.

The proposed revision of Regulatory Guide 1.118 provides updated NRC staff guidance for complying with the Commission's regulations regarding the periodic testing of the electric power and protection systems, and endorses ANSI/IEEE Standard 338-1987, "IEEE Standard Criteria for the Periodic Surveillance Testing of Nuclear Power Generating Station Safety Systems," with certain exceptions.

The staff intends to use this revision during its evaluation of future applications for construction permits, operating licenses, and licensee modifications to existing nuclear plants that require staff approval.

During the public comment period, the responsible IEEE subcommittee raised the concern that, when a safety system test is initiated by removal of fuses or the opening of a breaker, it may result in undesirable actuation of equipment during plant operations.

At our meeting, the NRC staff stated that they were close to a resolution of this concern with IEEE.

Subject to a resolution of the above concern that is acceptable to the staff, we have no objection to the issuance of Regulatory Guide 1.118, Revision 3.

Sincerely, T. S. Kress Chairman, ACRS

References:

1.

Memorandum dated February 1, 1995, from E. Beckjord, Office of Nuclear Regulatory Research, to J. Larkins, ACRS Executive Director, transmitting Proposed Revision 3 to Regulatory Guide 1.118, "Periodic Testing of Electric Power and Protection Systems" 2.

ANSI/IEEE Std 338-1987, "IEEE Standard Criteria for the Periodic Surveillance Testing of Nuclear Power Generating Station Safety Systems" Mr. James M. Taylor

Executive Director for Operations U. S. Nuclear Regulatory Commission Washington, DC 20555ÿ0001

SUBJECT:

PROPOSED FINAL AMENDMENT TO 10 CFR 50.55a TO INCORPORATE BY REFERENCE SUBSECTIONS IWE AND IWL, SECTION XI, DIVISION 1, OF THE ASME BOILER AND PRESSURE VESSEL CODE During the 418th meeting of the Advisory Committee on Reactor Safeguards, February 9ÿ10, 1995, we discussed the subject final amendment.

At this meeting, we had discussions with representa-tives of the NRC staff and the Nuclear Energy Institute (NEI).

We also had the benefit of the document referenced.

This proposed final amendment incorporates by reference the 1992 Edition with the 1992 Addenda of Subsection IWE (Requirements for Class MC and Metallic Liners of Class CC Components of LightÿWater Cooled Power Plants) and Subsection IWL (Requirements for Class CC Concrete Components of Light ÿWater Cooled Power Plants),Section XI, Division 1, of the ASME Boiler and Pressure Vessel Code with specified modifications and a limitation.

It also expedites the schedule for performing the containment examinations.

We concur with this staff position.

A number of utilities and NEI, which have commented on a draft version of this amendment, argue that it is overly prescriptive and contrary to the trend towards performance ÿbased regulation.

However, a suitable "metric," which could be used as the basis for a performanceÿbased inspection for the assurance of the structural integrity of the containment, seems difficult to identify.

Riskÿbased inspection appears to be a more promising approach to rationalizing in-service inspection of passive structural components.

The Office of Nuclear Regulatory Research is actively pursuing this approach, and we hope to see risk ÿbased concepts being used to develop requirements for in ÿservice inspections in the notÿtooÿdistant future.

Sincerely, T. S. Kress Chairman, ACRS

Reference:

Memorandum dated December 12, 1994, from E. Beckjord, Director, Office of Nuclear Regulatory Research, to J. Larkins, Executive Director, ACRS,

Subject:

Final Amendment to 10 CFR 50.55a to Incorporate by Reference Subsection IWE and Subsection IWL,Section XI, Division 1, of the ASME Boiler and Pressure Vessel Code