ML20043H016

Requests Exemption from App J to 10CFR50 for 12 Valves in Reactor Bldg Closed Cooling Water Sys.Valves Not within Definition of Containment Isolation Valves in App J & Not Required to Be Tested

ML20043H016
Person / Time
Site:	Millstone
Issue date:	06/08/1990
From:	Mroczka E NORTHEAST NUCLEAR ENERGY CO., NORTHEAST UTILITIES
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
B13516, NUDOCS 9006210448
Download: ML20043H016 (9)

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OMO General Offices e Seloon Street, Bethn, Connecticut 9 5 nI EcE5 P.O. BOX 270 U,I[NuUcU., H ARTFORD. CONNECTICUT 061410270 L k d inimm e mmc (203) 665-5000 June 8, 1990 Docket No. 50-336 B13516

, Re: 10CFR50, Appendix J 10CFR50.12 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 Gentlemen:

Millstone Nuclear Power Station, Unit No. 2 10CFR50, Appendix J Reauest for Exemotion Pursuant to 10CFR50.12, Northeast Nuclear Energy Company (NNECO) hereby l requests an exemption from the requirements of Sections III. A and III.C of i 10CFR50, Appendix J (primary reactor containment leakage testing) for twive valves in the Reactor Building Closed Cooling Water (RBCCW) System of Onit No. 2 of the Millstone Nuclear Power Station. This request is the culmination ,

of an exchange of correspondence with the NRC Staff, in which NNECO has main-tained its position that the valves should not be subject to those require-unents. The Staff has taken the contrary position. To resolve this issue, NNECO submits this request on the recommendation of the Staff.

For the reasons discussed below, we believe that an exemption from the requirements of Sections III.A and III.C for the twelve valves is authorized '

by law, will not present an undue risk to the public health and safety, is l consistent with the common defense and security, and is justified under various special circumstances specified in 10CFR50.12.

I. Backaround By letter dated July 14, 1987, NNEC0 voluntarily. advised the Staff that it had removed the twelve valves from its Type C Test Program. This letter expressed the view that the valves are not within the definition of containment isola-tion valves in Appendix J and therefore are not required to be tested. It also summarized the conclusions of a safety evaluation, conducted in accor-dance with 10CFR50.59, that justified the removal 'of the valves from the Type C Test Program.

The Staff responded with a November 3,1987 Request for Additional Informa-tion, to which NNEC0 responde.d on January 7, 1988. The request concerned-the pipe and support design basis for the RBCCW System at Millstone Unit No. 2.

The NNECO response observed that this design basis is consistent with the RBCCW System design basis described in the Final Safety Analysis Report (FSAR) 9006210448 900600 l!

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U.S. Nuclear Regulatory Commission j B13516/Page 2 June 8, 1990 .

l for the plant and approved in the May 10, 1974 NRC Safety Evaluation Report (SER) prepared in connection with the operating license; specifically, it indicated the section of the RBCCW System inside containment was fabricated equivalent to at least Safety Class 3 requirements of the American Society of  ;

Mechanical Engineers (ASME) Boiler and Pressure Vessel Code.

In a February 10, 1988 letter, the Staff advised NNECO that it disagreed with the decision to remove the valves from the Type C Test Program. The Staff expressed the view that "such systems should be fabricated to at least Safety Class 2 requirements (of the ASME Code] in order to not require post accident leak-tightness of the associated containment isolation." However, in a September 9, 1988 letter to the NRC Staff, NNECO clarified that the RBCCW System is a closed-loop system; that characterization is documented in the Millstone Unit No. 2 FSAR and confirmed in the NRC SER for the plant. NNECO also explained that the fabrication of the RBCCW System to Safety Class 3 requirements was in accordance with the acceptance criteria for those systems in effect when it was designed; we further observed that the Staff expectation that it should be fabricated to Safety Class 2 requirements appeared to be based on acceptance criteria in the NRC Standard Review Plan (SRP) that was adopted after Millstone Unit No. 2 was licensed. Therefore, NNECO expressed the position that the change in the design basis of the plant necessitated by fabrication of the RBCCW system to Safety Class 2 requirements is not justifi-able.

On November 10,1988, NNEC0 responded to several questions on the Inservice leection Program requirements for the twelve RBCCW System valves. There-

.er, the Staff reiterated its position on the inclusion of the valves in the fype C Test Program:

We have reviewed your submittals dated September 9 and November 10, 1988 and conclude that our position has not changed from the posi-tion stated in our letter dated February 10, 1988. Accordingly, Type C testing of containment isolation valves referenced in your letter dated July 14, 1987 should continue or such systems should be

, fabricated to at least Safety Class 2 requirements in order to not require post-accident leak-tightness of the associated containment isolation valves.

Letter to Edward J. Mroczka from John F. Stolz, Millstone Unit No. 2 Reactor Building Closed Cooling Water (RBCCW) System Containment Isolation Valves (May 4, 1989).

NNECO subsequently explained, in a June 20, 1989 letter, that its decision to remove the valves from its Type C Test Program was based on a safety evalua-tion conducted in accordance with 10CFR50.59. That regulation authorized NNECO to remove those valves without prior Staff approval if the removal involved no unreviewed safety question. NNEC0 expressed the position that the Section 50.59 safety evaluation was appropria9 and that it supported the revision to our Type C Test Program. NNECO a;- otterated that the RBCCW I

y U.S. Nuclear Regulatory Commission B13516/Page 3 June 8, 1990 System is a closed-loop system based on acceptance criteria in effect when Millstone Unit No. 2 was licensed. NNECO consequently expressed the position that the " application of SitP criterion in the case, therefore, would represent the imposition of a revised Staff position and would therefore have to be addressed under 10CFR50.109." The Staff acknowledged its receipt of a "backfit" position on July 11, 1989 and indicated it.would be considered in accordance with Chapter 0514 of the NRC Manual.

The Staff completed its consideration of our "backfit" position and advised '

NNECO in an August 14, 1989 letter that "[s]ince Appendix J (Type "C") testing of the subject containment isolation valves was part of the original design basis for Millstone Unit No. 2, denial of relief from such testing cannot be considered as a "Backfit" within the definition of 10 CFR Part 50, Section 50.109(a)(1)." That conclusion was based on Table 5.2-11 of the plant FSAR, which originally listed the twelve RBCCW System valves as required to be tested. The Staff inferred from this table that "these valves have been established as containment isolation valves requiring Type "C" testing as part of the current design basis of the facility." It also stated that NNEC0 had -

failed to justify the removal of the valves from its Type C Test Program.

Finally, on January 31, 1990, NNEC0 requested a clarification of the Staff position on this issue. Specifically, we questioned (i) whether the Staff believes the twelve valves are within the definition of containment isolation valves under Section II.H of Appendix J; (ii) its conclusion that the applica-tion of acceptance criteria in the current SRP to the RBCCW System is not a backfit under Section 50.109; (iii) whether the Staff believes the safety evaluation prepared by NNECO in accordance with Section 50.59 fails to justify removal of the valves from the Type C Test Program or is otherwise inadequate; and (iv) why the revised Type C Test Program and design basis for the RBCCW System are not the baseline for purposes of backfit analysis under Sec-tion 50.109. These questions precipitated the discussions between NNEC0 and the Staff that resulted in the Staff recommendation that NNECO resolve the issue by means of an exemption request.

II. Discussion The legal standard for exemptions from NRC regulations is established in 10CFR50.12. The Commission is authorized to grant an exemption upon a demon-stration that (i) the exemption is authorized by law, will not present an undue risk to the public health and safety, and is consistent with the common defense and security; and that (ii) certain specified special circumstances are present. The special circumstances specified in the regulation, specifi-cally 10CFR50.12(a)(2), include two that are present in this instance:

(ii) Application of the regulation in the particular circumstances

! would not serve the underlying purpose of the rule or is not neces-L sary to achieve the underlying purpose of the rule; [and]

v o U.S. Nuclear Reoulatory Commission Bi3516/Page 4 June 8, 1990 i (iii) Compliance would result in undue hardship or other costs that-are significantly in excess of those contemplated when the regula- l tion was adopted, or that are significantly in excess of those incurred by others similarly situated. I We discuss the standards of 10CFR50.12 below.

A. The Exemption is Authorized By Law, Will Not Present An Undue Risk To The Public Health And Safety, And Is Consistent With The Common Defense And Security The Commission is authorized to grant exemptions from the provisions of its regulations. In addition, because the common defense and security are not implicated by this exemption, it is consistent therewith.

Moreover, for several reasons, the exemption will not present an undue risk to the public health and safety. First, the twelve RBCCW System valves, for which the exemption is requested, do not serve a containment isolation func-tion. They are designed to be open in the event of an accident because the RBCCW System is intended to cool the Containment Air Recirculation (CAR)

System. This safety-related function requires the circulation of water in the RBCCW System (at a minimum pressure of 60 psig) in the event of an accident and consequently requires the valves to be open. As a result, the valves do not receive a containment isolation signal in the event of an accident--the remote manual actuation switches for some valves are locked in the open position in the control room; other valves will open on a Safety Injection Actuation System signal. Moreover, on a failure of DC power or instrument air, the valves would fail in the open position. Because the twelve valves do not serve a containment isolation function and are designed to be open in the event of an accident, it is unnecessary to test their leak-tight integrity in the closed position.

Second, the maximum pressure in the containment structure in the event of a design bases accident would be 54 psig. Because the minimum pressure in the L RBCCW System would be 60 psig, the only leakage through the valve seals would I be into the containment structure from the RBCCW System. For this reason, and I because the valves are designed to be open in the event of an accident, they.

- are not required to isolate the containment in that event. Again, therefore, 1 it is unnecessary to test their leak-tight integrity in the closed position.

l The valves would be closed only if a RBCCW System line or CAR System cooler ruptured inside the containment structure. However, it is unnecessary to test the valves because the possibility of a rupture in connection with a design basis accident is quite small. Specifically, the RBCCW System is a Seismic Category 1 system; it is functionally protected from dynamic effects by physical separation of redundant equipment--for exampl e, the CAR System coolers are separated; and it is protected from missiles projected through

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U.S. Nuclear Regulatory Commission B13516/Page 5 June 8, 1990 failures of components that are not Seismic Category I by virtue of its location and configuration.

Third, in the event of an accident with no RBCCW System operation, the surge tank that feeds the RBCCW System and through which it is vented would, as a result of its elevation, maintain a minimum pressure therein of 42 psig. l Therefore, the only leakage through the valve seals into the RBCCW System  !

would be that forced by containment structure pressure in excess of 42 psig.

Although the maximum pressure in the containment structure in the event of a design basis accident could be 54 psig, it is unlikely to exceed 42 psig after ,

the initiation of containment spray. Moreover, even if the containment I atmosphere in an accident leaks into the RBCCW System and into its surge tank, l that atmosphere would escape only into the enclosure building, where it would '

be collected and processed by the Enclosure Building Filtration System; a i spill from the surge tank would be retained in the enclosure building.

Consequently, the leak-tight integrity of the valves with no RBCCW System operation raises no significant safety issue.

For these reasons, an exemption from the test requirements of Appendix J will not present an undue risk to the public health and safety.

B. The Application Of Sections Ill. A Or III.C of Appendix J To The Twelve RBCCW Valves In These Particular Circumstances Would Not Serve The Underlying Purpose Of Appendix J And Is Not Necessary To Achieve That Underlyina Purpose Section III.C of Appendix J requires a Type C Test Program "to measure con-tainment isolation valve leakage rates." However, it is required only for certain valves defined in Section II.H of Appendix J:

The containment isolation valves included [in a Type C Test Program]

are those that:

1. Provide a direct connection between the inside and outside atmospheres of the primary reactor containment under normal opera-tion, such as purge and ventilation, vacuum relief, and instrument valves; l 2. Are required to close automatically upon receipt of a contain-ment isolation signal in response to controls intended to effect containment isolation; 1

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U.S. Nuclear Regulatory Commission B13516/Page 6 June 8, 1990 i

3. Are required to operate intermittently under post-accident l conditions; and i I

4. Are in the main steam and feedwater piping and other systems which penetrate containment of direct-cycle boiling water power i reactors.  ;

Section III.A, which establishes the requirements for Type A lest Programs, requires that certain fluid systems and other closed systems be vented to the containment atmosphere for purposes of Type A testing, and requires testing of  ;

containment isolation valves in such systems. Section llI.A.l.(d) provides:  ;

Portions of closed systems inside containment that penetrate con-tainment and rupture as a result of a loss of coolant accident shall  ;

be vented to the containment atmosphere... Systems that are normally i filled with water and operating under post-accident conditions, such as the containment heat removal system, need not be vented. How-ever, tlic containment isolation valves in the systems defined in III.A.I.(d) shall be tested in accordance with III.C.  ;

i The Staff has indicated that the RBCCW System should be considered to rupture '

in the event of a LOCA, and thus should be vented in containment for purposes  ;

of a Type A test as required by Section III.A.I.(d). The Staff believes the System could rupture because it was fabricated to Safety Class 3 requirements  ;

and not Safety Class 2 requirements. However, for reasons previously dis-cussed, the probability that the RBCCW System could rupture is quite small._  !

The System is protected against a failure induced by a design basis accident because it is a Seismic Category I system; it is functionally protected from dynamic effects by physical separation of redundant equipment; and it is protected from missiles projected through failures of components that are not >

l Seismic Category I by virtue of its location and configuration. Moreover, the fabrication of the RBCCW System to Safety Class 3 requirements was in accor-dance with the acceptance criteria for those systems in effect when it was designed; thus, consistent with the licensing basis of the plant, the proba- i bility of rupture should be assumed to be extremely small. For those reasons, an exemption from the requirements of Section III.A.l.(d) should be granted,  ;

since this testing is not necessary to achieve the underlying purpose of the '

rule.

Regarding Section II.H of Appendix J, the valves provide no direct connection between-the~inside and outside atmospheres of the primary reactor containment under normal operation. The RBCCW System is a closed-loop system; that '

l characterization is documented in the Millstone Unit No. 2 FSAR and confirmed L in the NRC SER for the plant. It is not open to the outside atmosphere. The Staff has concluded that it is not a closed-loop system under current SRP criteria because it was fabricated to Safety Class 3 requirements; nonethe-less, the RBCCW System was designed and fabricated with no direct connection

between the inside and outside atmospheres. Moreover, the surge tank through

! which the RBCCW System is vented provides a connection only into the enclosure l

I U.S. Nuclear Regulatory Commission B13516/Page 7 June 8, 1990 building, where an Enclosure Building Filtration System would collect and process any inside containment atmosphere that leaked into the RBCCW System in the event of an accident. Therefore, the twelve valves for which the exemp- i tion is requested are not within the intent of the first provision of the Section II.H definition of containment isolation valves.

Second, the valves serve no containment isolation function and are not required to close automatically on a containment isolation signal. The RBCCW System is designed to cool the CAR System in the event of an accident. These valves, therefore, are required to be open in that event. The remote manual actuation switches for some valves are locked in the open position and other valves would open on a Safety Injection Actuation System signal; the valves would not receive a containment isolation signal. They would be closed only if a RBCCW System line or CAR System cooler ruptured inside the containment t structure; however, the possibility of a rupture in connection with a design basis accident is quite small. In any event, the twelve RBCCW System valves are neither designed nor required to isolate the containment structure in the event of an accident. Consequently, they are not within the intent of the definition of containment isolation valves under Section II.H.2.

Third, and similarly, the valves are not required to operate intermittently in an accident. This provision of the Section II.H definition apparently refers to valves that have a safety function which requires them to be open in an accident for an interim period and thereafter are required to be closed for containment isolation purposes; for example, in the proposed regulation published by the Atomic Energy Commission in 1971, this provision referred to

" valves in engineered safety systems penetrating containment which under post-accident conditions are required to close following the termination of the safety function." 36 Fed. Reg. 17,053, 17,054 (Aug. 27, 1971). However, the safety-related function performed by the RBCCW System is not an intermit-tent function; it requires the circulation of water, and the valves to be open, throughout an accident. Therefore, this provision of the Section II.H definition also is in%qlicable to the twelve RBCCW valves for which the exemption is requested."'

For the foregoing reasons, the twelve RBCCW valves for which the exemption is requested are not within the intent of the regulation as reflected in the definition of containment isolation valves that are required to be tested; I

therefore, the inclusion of those valves in the NNEC0 Type C Test Program for Millstone Unit No. 2 would not serve the purpose of either Section Ill.A.I.(d) or III.C and is not necessary to achieve it.

i l (1) Millstone Unit No. 2 is a pressurized water reactor. The fourth provision of the Section II.H definition is applicable only to boiling water reactors.

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U.S. Nuclear Regula. tory Commission 1 B13516/Page 8 June 8, 1990 C. Compliance With Appendix J Would Result in Undue Hardship And Other Costs That Are Significantly In Excess Of Those Contemplated When the Reculation Was Adopted The inclusion of the RBCCW System and valves in the NNECO Appendix J Test Program for Millstone Unit No. 2 essentially would require the replacement of the valves to ensure their leak-tight integrity in the closed position within the acceptance criteria specified in Appendix J. NNEC0 estimates that this replacement would involve a considerable expense--approximately $2,000,000.

For the reasons discussed in Section ll.A 'of this exemption request, NNEC0 believes that the replacement of the valves and their inclusion in the Type C Test Program would provide a negligible benefit to the public health and safety. The application of the regulation, moreover, would necessitate the expenditure of considerable funds that otherwise could be put to better use in L improving the operation of the plant. On balance, therefore, NNEC0 believes that its compliance with Appendix J with respect to the RBCCW System and valves would result in an undue hardship and other costs that are signifi-cantly in excess of those contemplated when the regulation was adopted.

III. Conclusion Because the requested exemption is authorized by law, will not present an undue risk to the public health and. safety, is consistent with the common defense and security, and is justified under various special circumstances specified in 10CFR50.12, the NRC should grant the exemption. In the event the NRC Staff chooses not to grant this exemption or places conditions on it not previously agreed to, NNEC0 reserves the right to submit a formal backfitting application pursuant to 10CFR50.109. Kindly contact us if you have any questions or require any additional information.

Very truly yours, NORTHEAST NUCLEAR ENERGY COMPANY ,

E. J. MroczkjV ~ (/

Senior Vice President l

cc: T. T. Martin, Region I Administrator G. S. Vissing, NRC Project Manager, Millstone Unit No. 2 P. Habighorst, Resident Inspector, Millstone Unit No. 2 W. J. Raymond, Senior Resident Inspector, Millstone Unit Nos. 1, 2, and 3 T. E. Murley, Director, Nuclear Reactor Regulation E. L. Jordan, Director, Office for Analysis and Evaluation of Operational Data

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U.S. Nuclear Regulatory Commission B13516/Page 9 June 8, 1990 STATE OF CONNECTICUT) ss. Berlin COUNTY OF HARTFORD Then personally appeared before me, E. J. Mroczka, who being duly sworn, did state that he is Senior Vice President of Northeast Nuclear Energy Company, a Licensee herein, that he is authorized to execute .and file the foregoing information ia the name and on behalf of the Licensee herein, and that the statements contained in said information are true and correct to the best of his knowledge and belief, dlAM/th  : hl 4'

JNotary P(/ublic -

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