ML20154J019

Responds to Request for NRR Assistance on Allegation Re Containment Integrity & Inservice Testing Program.Safety Significance of Not Performing Inservice Testing of Certain Affected Valves During First Yr of Plant Operation Minimal

ML20154J019
Person / Time
Site:	Clinton
Issue date:	09/16/1988
From:	Stevens J Office of Nuclear Reactor Regulation
To:	Miller H NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
References
TAC-67447, NUDOCS 8809220214
Download: ML20154J019 (7)

Text

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NUCLEAR REGULATORY COMMISSION 3 g ca WASHINGTON. D. C. 20666 September 16, 1988

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Docket No. 50-461 MEMORANDUM FOR: Hubert J. Miller Director Division of Reactor Safety RIII THRU: Daniel R. Muller, Director Project Directorate III-2 h lh' DivisionofReactorProjectIII,$

IV, Y and Special Projects NRR FROM: Janice A. Stevens, Project Manager Project Directorate III-2 Division of Reactor Project III.

IV, Y and Special Projects NRR

SUBJECT:

RESPONSE TO REQUEST FOR NRR ASSISTANCE ON THE CLINTON ALLEGATION CONCERNING CONTAINMENT INTEGRITY AND THE INSERVICE TESTING PROGRAM (TIA III-2-88/ TAC NO. 67447)

This memorandum responds to your request for technical assistance dated March 1, 1988 relating to allegations concerning deficiencies with the containment integrity at the Clinton Power Station and the Inservice Testing (IST) Program. A summary of the evaluaticn of these allegations is given below.

The alleger contends that the regulations (GDC-55 and GDC-56,10 CFR Part 50, Appendix A) require, as a minimum, two containnent isolation valves: one inside, containment, and the other outside containment for each containment penetration. The penetration for high-pressure core spra core spray (LPCS), and low-pressure core injection (LPCI)y "C"/RHR(HPCS),

"C" lines low-pressure each have one outboard motor-operated valve and one inboard air-testable check valve, yet only the outboard motor-operated gate valves are being considered as containmentisolationvalves(CIVs). The alleger further contends that the inboard testable check valves should be considered as CIVs and should be tested as required in Appendix J.10 CFR Part 50, and the penetration with the most leakage should be added to the running total for the containment building.

The Plant Systems Branch has reviewed the allegation concerning the primary containment integrity ano agrees with the alleger that the four inboard testable check valves, IE22F005, 1E51F006, 1E21F006, and 1E12F0410, should be considered as CIVs.

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. . I The HPCS line penetrates the drywell to inject water into the reactor pressure vessel. Isolation of the reactor coolant pressure boundary is provided by an air-t2 stable check valve (1E22F005) located inside the dr manually controlled, motor-operated gate valve (1E22F004)ywell and athe located outside remote, containment. The containment isolation is maintained by this outboard motor-operated gate valve. Similarly, for LPCS, reactor core isolation cooling i (RCIC) and LPCI "C"/RHR "C" lines, the isolation criteria for the reactor coolant pressure boundary are accomplished by the use of inboard air-testable check valves (1E21F006, 1E51F066 and IE12F0410) and outboard remote, manually controlled motor-operated gate valves (1E21F005,1E51F013, and 1E12F042C) with position indicators in the control room. Both of these types of valves are nonnally closed, with the motor-operated valves receiving an automatic signal to open in the event of a loss-of-coolant accident. The licensee of the Clinton Plant has considered the outboard motor-operated valves as CIVS which are being tested as per the requirements given in Appendix J of 10 CFR Part 50.

However, the licensee has not considered inboard testable check valves for containment isolation. The inboard testable check valves are considered

, pressure isolation valves (P!Vs). These P!Vs are hydraulically tested for a system differential pressure of 1000-psi once every 18 months. The leakage acceptance criterion for PIVs is 0.5 gpm per nominal inch of the valve diameter.

On the basis of its review of the FSAR and the plant Technical Specifications, the staff concludes that the four inboard air-testable check valves (1E21F005 for HPCS, 1E51F066 for RCIC, 1E21F006 for LPCS and 1E12F041C for LPCI "C"/RHR "C") should be considered as CIVs; therefore, these testable check valves should be included in Table 3.6.4-1 of the containment isolation valve in the plan', Technical Specification, and should be tested in accordance with the requirements of Appendix J of 10 CFR 50. It should be noted that other Mark I!! plants confonn to the arrangement that the inboard check valves are CIVs and are tested as required in Appendix J.10 CFR Part 50.

I It should also be noted that the testable check valves (1E12F041A and 1E12F0418) in LPCI "A" and LPCI "B" lines need not be considered CIVs. LPCI "A" and LPCI *B" lines that penetrate the containment have inboard remote-manually controlled, motor-operated, nonnally closed CIVs (1E12F042A and 1E12F0428), and outboard remote-manually controlled, motor-operated, nonnally I open CIVs (1E12F027A and 1E1200278). The outboard CIVs can be closed to

! provide containment isolation in the event of a high-energy line break inside containment. Thus, the design of containment isolation provisions satisfy the requirements of GDC 55, 56 and 57.

The decision to not allow the closed system as the second barrier for the i penetrations containing the check valves in question centers upon the staff's desire to have two independent barriers for each containment penetration. For the ECCS systems, the suction penetration has a remote manual valve and a closed system. Although the system does not fully meet all staff requirements for a closed syste:n, the staff has accepted it 46 the second barrier since the addition of a second valve would reduce the availability of the system. This is not true for the discharge line. For the discharge penetration, giving credit for the closed system outside containment would require using the same

barrier for two penetrations. Consideration of the inside check valve as a containment barrier also does not reduce system availability. It is for the above reasons that the staff has concluded that the containment isolation barriers should include the check valves.

With respect to the issue of leak testing, the staff is currently discussing with the licensee the specific testing procedures to satisfy the leak testing requireu nts. It is the contention of the licensee that leak testing at full system pressure (i.e.1000 psi) is equivalent to air testing at 15 psi. This full system pressure test is in a sense continuous since the check valve is always exposed to system pressure during operation. The staff is evaluating the merits of this approach. Until the staff completes the evaluation, adequate safety margins exist due to the testing that has been done as well as the fact that the systems are expected to function following a LOCA. Therefore isolation is not needed. It is only for the low probability event when the safety system needs to be isolated.

The above infonnation completes our response to the allegation concerning Clinton containment integrity issues. As a separate NRR review, the Plant Systems Branch is evaluating the unique correlation between a 1000-psi water test and an Appendix J. Type "C" air test for inboard testable check valves as discussed above. A copy of this evaluation will be provided to you upon distribution.

The alleger also contends that can in valves in the Clinton plant that should have been inservice tested were not because they were not included in the Clinton IST program. There are about 108 valves involved in the allegation.

The Mechanical Engineering Branch (EMEB) and the Idaho National Engineering Laboratory (INEL) have evaluated the safety-related function, if any, of all involved valves as well as 12 additional valves, i.e., IB21-F098A, B, C, 0; IE12-F051A, B; IE12-F065A, B; and ICC065, 067, 068, 070.

The Code of Federal Regulations, paragraph 10 CFR 50.55a(g) requires safety-related valves in water-cooled nuclear reactor facilities to meet IST require-a nts stated in the ASME Code,Section XI, Subsection IWV, "Inservice Testing of Valves in Nuclear Power Plants." Per code requirements, a valve must be inservice tested if it perfoms an active safety function in shutting down the reactor or mitigating an accident. However, a valve may be exempted from inservice testing if it is only used for operating convenience, system control, or maintenance.

The EMES anti INEL's review of all specified valves indicates that 23 of the 120 1 valves are required to perfom a safety-related function. Per Section XI requirements, these 23 valves should be included in the Clinton 15T Program and inservice tested. With regard to the m maining 97 valves, 18 of them are used only for operating convenience and maintenance and are not required to be tested. The remaining 79 valves perform certain system functions; however, none of these valves perfonn an active safety function. Therefore, they are not required to be tested in accordance with Section XI.

Among the 23 valves that are required to be tested,16 of them were added to the Clinton IST program in Revision 2, which became effective on June 30, 1987. The connercial date of the Clinton Power Station was April 24, 1987, a

The NRC regulations and ASME Code,Section XI require that the first inservice test for most valves be performed within three months after connercial

.i operation. For Clinton, the first IST occurred during July 1987. Ten (10)of J these 16 affected valves (Table I, Valve No.1 thru 16) were not tested during the first scheduled IST but were all tested during September / October 1987.

Since then, these valves have been tested in accordance with Section XI requirwnents and have been verified operable. This information is based on verbal input from Region !!!. Although these 16 valves might not have been included in the IST program at the time of the allegation, they were in-corporated shortly after the first scheduled inservice testing. Thus, although 4

the allegation had merit at the time it was made, the early omission of valves from the IST program has not resulted in any real ssfety impact to the Clinton Plent.

As a result of interactions with the licensee, seven additional valves (Table I, Valve No.17 thru 23) were added to the Clinton IST program, Revision 5, dated May 27, 1988. This was about one year after connercial operation. A safety evaluation of each valve, missing from the IST for slightly more than a year, is discussed below, j Four of these valves (IE12-F051A, B and F065A, B) were designed to be used i during the steam condensing mode of the Residual Heat Removal System. This I

mode is not intended to be used any longer at Clinton and steps were taken to delete this mode and also these valves from the IST program. Although the staff finds that these four valves should have been subjected to IST prior to their removal from the IST program, not testing them during the past year presented minimal impact to the safety of plant operations simply because this j mode of operation is not to be used at the Clinton Plant.

i Valve IFC091 is a relief valve.Section XI requires a relief valve to be

. tested approximately on a five-year cycle. Not testing this valve during its first year of the IST program does not violate Section XI requirements, and therefore presents no safety concerns.

i Valves IE12-F040 and -F049 are part of the Residual Heat Removal System which

provides shutdown cooling for the reactor. These two valves were added to the l IST program (Rev. 5) and were tested on July 27, 1988. They were both found operable. This information is based on verbal input from Region !!!. Since 1 the recent test has verified the operability of these valves, the safety l significance of not testing them more frequently as required by Section XI is minimal, i

, Based upon the discussion above, the staff finds that approximately 20 percent of the valves in the allegation are required to perfonn a safety-related function while the rest are not. Those valves that are required to be in-service 'ested are now included in the Clinton IST Program. Furthennore, the 7

staff fin.

that the safety significance of not performing inservice testing of i

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certain affected valves during the first year of plant operation is minimal.

As such, the staff concludes that no further action is required and that this allegation is resolved.

For further infonnation or clarification, please contact me at 492-1397.

Janice A. Stevens, Project Manager l Project Directorate !!!-2 1 Division of Reactor Projects !!!, l

!Y, V and Special Projects, NRR cc: R. Cooper, RI!!

D. Danielson, RI!!

J. Kudrick, PSB J. Huang MEB l

, TA8LE I VALVES THAT ARE REQUIRED TO BE INSERVICE TESTED Valve No. Valve Identification 1,2 OVC10A, 8 3,4 GVC17A, B 5,6 OVC20A, B P 7,8 OVC25A, B 9,10 1821-F001, F002 11,12 1E12-F037A, 8 13.14 1FC085A, B 15,16 1E51-F004, F005 l 17,18 1E12-F051A, B 19,20 1E12-F065A, B i 21.22 1E12-F040, F049 23 IFC091 l

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certain affected valves during the first year of plant operation is minimal.

As such, the staff concludes that no further action is required and that this allegation is resolved.

For further information or clarification, please contact me at 492-1397.  :

Janice A. Stevens, Project Manager Project Directorate !!!-2 Division of Reactor Projects !!!,

l IV, Y and Special Projects, NRR -

l cc: R. Cooper, Rill D. Danielson, RI!!

J. Kudrick, PSB J. Huang, MEB '

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