IR 05000461/1988020
| ML20195C273 | |
| Person / Time | |
|---|---|
| Site: | Clinton  |
| Issue date: | 10/18/1988 |
| From: | Danielson D, Huber M, James Smith NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20195C262 | List: |
| References | |
| 50-461-88-20, NUDOCS 8811020426 | |
| Download: ML20195C273 (21) | |
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-U.S. NUCLEAR REGULATORY COMMISSION ,
REGION III
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Report No. 50-461/88020(DRS)
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Docket No. 50-461 License No..NPF-62 Licensee: Illinois Power Company 500 South 27th Street Decatur, IL 62525
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Facility Name: Clinton Power Station, Unit 1 ;
Inspection At: Clinton Site, Clinton, Illinois
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Inspection Conducted: . July 27-28 and October 18, 1988
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Inspectors- . SmitA a - , -,,
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t1 H. P. Huber l
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Approved By:
ob79dk&W D. H. Danielson, Chief / /8 tr ;
Materia: and Processes Section Date
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Inspection Sumary i
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Inspection on July 27-28 and October 18, 1988 (Report No. 50-461/88020(DR52 Areas Inspected: Special safety inspection to follow-up on Allegation .
RIII-87-A-0027 (99014). :
Results: Review of this allegation identified one violation for failure to
, meet IST requirements as stated in the ASME Code,Section XI, Subsection IWV.
i No reply to the violation is require *
The valves required to be inservice tested are now included in the Clinton IST Progra *
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The safety significance of not performing inservice testing of certain
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affected valves during the first year of plant operation is minima I.
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DETAILS
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, ,P_ersons Contacted -
Lillinois' Power Company (IP) ,
T*J.' Weaver.. Director, Licensing ;
J. Brownell,l.icensing Specialist ;
- S. R. Bell, Supervisor, Inservice Inspection
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-*T. Stevenson, Staff Engineer
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- Denotes those who participated in the telephone exit interview on s, October 18, 198 !
" Followup on Allegations -
e (Closed)AllegationRIII-87-A-0027 >
s Background ,
In March 1987, an individual identified potential deficiencies with the inservice testing (IST) of safety-related valves and the lack of containment integrit To respond completely to the alleger, the NRC's Headquarters staf provided assistance in addressing two of the concerns. Additional inspectior.s were conducted to obtain all the necessary information ,
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neeoed to thoroughly evaluate and address the alleger's concerns. These issues are identified in NRC Inspection Report No. 50-461/07027(DRS). :
t Specifically, these allegations were:
(1) Pressure isolation check valves inside the drywell were not being considered containment isolation valves (CIV) and appropriately tested as required by 10 CFR Part 50, Appendix J; and (2) Certain valves were not included in the IST program although l they should have bee (
The chronology of major activities associated with this allegation are provided in detail as Attachment 1; t!e NRC Headquarters evaluation of the two concerns is provided in Attachment 2; and a summary of the NRC's response to the concerns provided by the alleger in his June 16, !
1988, letter is provided in Attachment j NRC Review (1) With respect to item (1) above, the alleger contended that ,
certain testable check valves should be considered as CIV's, i
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and tested as such. The specific valves in question were:
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p Valve Number Valve 1E22F005 HPCS Testable Check IE21F006 LPCS Testable Check 1E12F041A LPCI from RHR A Testable Check 1E12F0418 LPCI from RHR B Testable Check 1E12F041C LPCI from RHR C Testable Check 1E51F066 RCIC Testable Check-Regulations (GDC-55 and GDC-56,10 CFR 50, Appendix A) require, as a minimum, two containment isolation valves: one inside containment, and the other outside containment for each containment penetration. The penetration for High-Pressure Core Spray (HPCS), Low-Pressure Core Spray (LPCS), and Low-Pressure Core Injection (LPCI) "C"/RHR "C" lines, each have one outboard motor-operated valve and one inboard air-testable check valve; yet only the outboard motor-operated gate valves were being considered as containment isolation valves (CIVs).
NRC Region III forwarded this information to NRC's office of NRR for their evaluation. By memo dated September 16, 1988, from D. R. Muller to H. J. Miller the staff discussed their resolution of the two allegations, Atachment 2. On the basis of its review of the FSAR and the plant Technical Specifications, the staff concluded that the NRC's previous assessment of this issue was in error and that the four inboard air-testable check valves (1E21F005 for HPCS, IE51F066 for RCIC, 1E21F006 for LPCS and-IE12F041C 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 valves in the plant Technical Specifications, and should oe tested in accordance with the requirements of Appendix J of 10 CFR 5 It should also be noted that the testable check valves (1E12F041A and 1E12F041B) in LPCI "A" and LPCI "B" lines need not be considered CIVs. The LPCI "A" and LPCI "B" lines that penetrate the containment, have inboard remote-manually controlled, motor-operated, normally closed CIVs (1E21F042A and 1E12F0428), and outboard remote-manually controlled, motor-operated, nonnally open CIVs (IE12F027A and 1E12F0278).
The outbcard CIVs can be closed to provide containment isolation in the event of a high-energy line break inside containmen Thus, the design of containment isolation provisions satisfies the requirements of GDC 55, 56 and 5 Portions of this allegation are substantiated. The licensee was formally notified by NRR on August 26, 1988, that the four inboard air-testable check valves identified above are CIVs and that appropriate corrective action was necessary. This includes adding these valves to the FSAR and Technical Specifications as CIVs and testing them accordingly. This portion of the allegation is close * '
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(2) Item _(2) above concerns several valves identified by the alleger as valves that were not being tested as they should have been because they were not included (n the Clinton IST program. At the time of the allegation, sorr.e of the safety-related valves identified by the alleger were omitted from the program and the Clinton IST program was not complete. This failure to meet IST requirements as stated in the ASME Code,Section XI, Subsection IWY, "Inservice Testing of Valves in Nuclear Power Plants" is a violation of paragraph 10 CFR 50.55a(g) (461/88020-01). There were about 108 valves involved in the allegation as well as 12 additional valves which were evaluated as part of the NRC revie The NRC and its contractor's review (See Attachment 2) of all involved valves indicates that twenty three (23) of the 120 valves are required to perform a safety-related function. Per ASME Section XI requirements, these 23 valves should be included in the plant IST program and be inservice tested. With regard to the remaining 97 valves,18 of them are used only for operating convenience and maintenance and are not required to be teste The remaining 79 valves perform certain system functions; however, none of these valves perform an active safety function. Therefore, they are not required to be tested in accordance with ASME Section X The 23 valves that were required to be included in the IST program are listed below:
Valve N Valve Identification 1, 2 OVC10A, B 3, 4 OVC17A, B 5, 6 OVC20A, B 7, 8 OVC25A, B 9, 10 IB21-F001, F002 11, 12 1E12-F037A, B 13, 14 1FC085A, B 15, 16 1E51-F004, F005 17, 18 1E12-F051A, B
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19, 20 1E12-F065A, B
. 21, 22 1E12-F040, F049 23 1FC091 The NRC inspectors reviewed the licensee's test records to
, determine to what extent testing had been performed on these l valves. For Clinton, the first IST occurred during July 1987.
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Ten of these 16 affected valves (Valve No. 1 thru 16) were i
not tested during the first scheduled IST but were all tested during September / October 198 Since then, these valves have been tested in accordance with ASME Section XI requirements and
, have been verified operab4.
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Valves 17 thru 20 (IE12-F051A, B and F065A, B) were designed to be used during the steam condensing mode of the Residual Heat Removal System. This 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 found that these four. valves should have been subjected to IST prior to their removal from IST program, not testing them during the past year presented minimal impact to the safety of the plant operation simply because this mode of operation is not to be used at the Clinton plan Valve IFC091 is a relief valve. ASME 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 ASME Section XI requirements, and therefore presents no safety concern 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 IST program (Rev. 5) and were tested on July 27, 1988. They were both found operable. Since the recent test has verified the operability of these valves, the safety significance of not testing them more freqaently as required by ASME Section XI is minima Although the allegation had merit at the time it was made, all appropriate valves have .ww been incorporated into the Clinton IST
' program. Furthermore, the staff found that the safety significance of not performing IST of certain valves during the first year of plant operation is minimal. Therefore, the staff concluded that no further action is required and that the allegation is resolve Examination of S&L Plant Configuration Review A point of concern identified by the alleger during the Knoxville, Tennessee, April 13, 1988, meeting was the statement by General Electric, "However, the overall impact due to stroke time changes ...
should be evaluated by others, since this is dependent on plant configuration details." This observation was made when approving increased stroke time The plant configuration review was performed by S&L before licensin Specifically, this review addressed areas such as, "What type environment will the valve be in and how will the change affect the corresponding area?"
The Engineering Change Notices were prepared, reviewed and approved by the same level of engineers and executives as would have been used to produce an original design document. The scope of the work, identification of the valves and the results of the review were clear and concise. The S&L review Concluded that "the increased stroking tinie will not have an impact on the environmental ... and radiological concerns."
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The stroke time of these valves is considered a design characteristi Design baseline reviews are required to determine if a design change affects the design baseline. If the design affects the baseline, changes are required to restore the baseline. Scme of the specific areas that are included in the design baseline reviews are:
ALARA (radiation consideration)
Environment
Flood Protection Once an ECN is completely reviewed, the approvei, who is a Registered Professional Engineer, is required to sign the document. The NRC <
inspectors did not feel that it was necessary to further confirm the competence of the S&L approve Furthermore, the NRC inspectors examined documents involved in S&L's review of the valves in question, including the following:
Action /Inforestion Requests (A/IR) 646 and 1022 t
Engineering Change Notice (ECN) 8693 Field Engineering Change Notice (FECN) 15012
Document Transmittal Form (Identifying affected valves)
JX-1183 Based on this review, the inspectors concluded that the valves with the modified stroke timos are acceptabl Inasmuch as the overall impact due to stroke time changes was evaluated and accepted by S&L, the process was properly documented and approved, and the NRC independently sampled the S&L reviews, this area of concern is considered close ,
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3. Exit Interview ,
The Region !!! lead inspector contacted the licensee representatives ,
(denoted in Paragraph 1) by telephone on October 18, 1988, and sunnarized i the results of the inspection and of the NRC Headquarters review of parts of the allegation. The inspector discussed the likely informational content of the inspection report. The licensee acknowledged this ,
information and indicated that none of it was proprietar ;
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ATTACHMENT 1 CHRONOLOGY OF MAJOR ACTIVITIES IN THE INSERVICE TESTING (IST)
AND CONTAINMENT INTEGRITY ALLEGATION AT CLINTON POWEk STATION (RllI-67-A-0027)
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March 11, 1987 - Telephone conversation with allege Concerns identified by alleger:
(1) Valve stroke times for solenoid operated containment isolation valves did not match FSAR, preoperational testing, or Technical Specification value (2) Positions for valves were not properly verifie '
(3) Docunentation did not exist for many test (4) Many valves were not tested at al (5) Illinois Power was notified of these problems but did nothin March 12, 1987_- Letter to NRC from allege Concerns identified by alleger:
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(1) ... no containment integrity ..."
- TS 3/4.6.4 could not be me No testing of ECCS injection line check valves to Appendix J criteri (2) "... failure to meet pre-engineered licensing commitments ..."
- Changing stroke times by 150% and subsequently changing design time Differences exist between FSAR, Tech Spec, design, and IST stroke time March 25, 1987, through June 1, 1987 - Inspections conducted in response to conversation and letter (Reference NRC Inspection Report No. 50-461/87014(DRS)).
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Inspections consisted of:
(1) Review of the following valves against concerns identified in phone conversation and March 12, 1987, letter:
Process Sampling System (PS)
1PS004 IPS005 1PS009 1PS010 1PS016 1P5017 1P5022 1PS031 1PS032 1PS034 1P5035 1P5037 1PS038 IP5043A 1P5043B 1PSO44A 1P5044B 1P5047 1PSO48 IP5056 1PS069 1P5070
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Containment Monitoring System (CM)
ICM011 1CM012 1CM022 1CM023 1CM025- ICM026 1CM047 1CM048 (2) Review of concerns against progra More specifically, a review of the IST program in the areas identified during the inspection with possible weaknesse Only this sample was reviewed due to the apparent similarity between concerns raised in phone conversation and lette Inspection Results 2 violations identified:
j - Failure to meet TS 4.0.4, 4.0.5 and 3.6.4.
- - Inappropriate procedure, violation of Criterion V, 10 CFR 50, Appendix Unresolved item (Unrelated).
Conclusions
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Technical deficiencies existed in IST program. However, resolution of i the deficiencies was in progress during NRC review of the program.
! Allegation was substantiated to the extent that the program was
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deficient, not that all requirements were not me June 20, 1987. - Alleger response to NRC Inspection Report No. 50-461/87014(DRS).
Concerns identified were same as the questions raised in the alleger's March 12, 1987, letter, with addition of request for personnel approving license August 12, 1987 - Response to alleger from H. J. Mille Detailed explanation provided to alleger. Information included additional efforts undertaken to improve valve reliability, explanation of licensing basis for non-testing of ECCS check valves per Appendix J, and basis for stroking requirements for valves in the IST progra November 12, 1987 - Letter from alleger to Chainnan lec December 6, 1987 - Alleger response to A. B. Davis letter dated December 2, 198 No new concerns identified, however, the alleger requested a meetin The decision w6s made to meet with alleger to discuss all concerns and the resolution of his concern ._ _ _ _ _ - _ _ _ _
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January 23, 1988 - Memo to NRR detailing allegatio March 1, 1988 - Memo to NRR requesting assistanc March 8-9, 1988 - Inspection conducted to perform compre'iensive review of allegers concerns and licensee corrective action The following were reviewed:
(1) Current and past status of IST progra (2) Discrepancies between stroke times in various documents for valves provided by allege (3) Status of valves, prior to fuel load up to present, to ensure operabilit (4) Corrective actions taken to address violations identified in NP,C Inspection Report No. 50-461/87014(DRS) and actions taken by the licensee in response to alleger memos (provided to licensee while still employed).
April 13, 1988 - Meeting with the alleger at Knoxville, Tennessee to discuss concerns in order for the NRC to obtain a better understanding of the allegatio May 27, 1988 - NRC Inspection Report 50-46/87027(DRS) issue Inspections conducted to address allegations and concerns raised by alleger during Knoxville meeting. No problems note June 16, 1988 - Letter to NRC from alleger concerning NRC Inspection Report No. 50-4LT/B7027(DRS). .
i (1) Questioned details of A/IR No. 646, conducted by S& (
(2) Questioned details of position indication tests and retesting after maintenanc ,
(3) Questioned QA program and administrative test control September 16, 1988 - NRR response to Region I!! request for assistance l regarding containment integrity and the inservice test program. Sumarizes !
alleger's concerns in these areas and provides similar sumary of the staff's '
evaluations of these concerns. (Included in this report as Attachment 2.)
October 24, 1988 - Letter issued to alleger addressing questions forwarded to the NRC by letter from alleger dated June 16, 198 j
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October 1988 - This report issued detailing results of NRR review and closing allegation l
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TABLE I VALVES THAT ARE REQUIRED TO BE INSERVICE TESTED
Valve.N ,
Valve Identification
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'1,2 OVC10A, B 3,4 OVC17A, B !
5,6 OVC20A, B l
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7,8 OVC25A, B
9,10 1821-F001. F002 -
11,12 1E12-F037A, B !
13,14 1FC085A, B 15,16 1E51-F004, F005 17,18 1E12-F051A, B 19,20 1E12-F065A, B 21,22 1E12-F040, F049 23 1FC091
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UNITED STATES NUCLEAR REGULATORY COMMISSION
{i j WASHING TON. D. C. 20555 September 16, 1988 Docket No. 50-461 MEMORANDUM FOR: Hubert J. Miller, Director Division of Reactor Safety, RIII l
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THRU: Daniel R. Muller, Director Prcject Directorate III-2 k lh DivisionofReactorProjectIII,$
IV, Y and Special Projects, NRR FROM: Janice A. Stevens, Project Manacer 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 !!!-2-88/ TAC NO. 67447)
This memorandum responds to your request for technical assistance dated e 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 evaluation of these allegations is given belo The alleger contends that the regulations (GDC-55 and GDC-56,10 CFR Part 50, Appendix A) require, as a minimum, two containment 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 (HPCS),
"C"/RHR low-pressure
"C" lines 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 A.opendix J 10 CFR Part 50, and the penetration with the most leakage should be added to the running total for the containment buildin The Plant Systems Branch hrs reviewed the allegation concerning the primary containment integrity and agrees with the alleger that the four inboard testable check valves, 1E22F005, IE51F066, 1E21F006, and 1E12F041C, should be considered as CIV .;;-
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-2-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-testable check valve (1E22F005) located inside the dr manually controlled, motor-operated gate valve (1E22F004)ywell and a remote, located outside the containment. The containment isolation is maintained by this outboard motor-operated gate valve. Similarly, for LPCS, reactor core isolation cooling (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 1E12F0410) and outboard remote, manually controlled motor-operated gate valves (1E21F005, IE51F013 and IE12F042C) with position indicators in the control roo 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 5 However, the licensee has not considered inboard testable check valves for containment isolation. The inboard testable check valves are considered pressure isolation valves (PIVs). These PIVs are hydraulically tested for a system differential pressure of 1000-psi once every 18 mor.tns. The leakage acceptance criterion for P!Vs is 0.5 gpm per nominal inch of the valve diamete 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 plant Technical Specification, and should be tested ',n accordance with the requirements of Appendix J of 10 CFR 50. It should be noted that other Mark !!! plants conform to the arrangement that the inboard check valves are CIVs and are tested as required in Appendix J.10 CFR Part 5 It should also be noted that the testable check valves (1E12F041A and
. 1E12F041B) 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 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 5 The decision to not allow the closed system as the second barrier for the 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 system, the staff has accepted it as the second barrier since the addition of a second valve would reduce the availability of the system. This is not true for the discharge lin For the discharge penetration, giving credit for the closed system outside containment would require using the sane
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-3-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 valve 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 requirements. 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 isolate 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 distributio The allager also contends that certain valves in the Clinton plant that should have been inservice tested were not beccuse they were not included in the Clinton IST program. There are about 108 valves involved in the allegatio The Mechanical Engineering Branch (EMEB) and the Idaho National Engineering Laboratory (INEL) have evaluated the safety-relateu function, if any, of all involved valves as well as 12 additional valves, IB21-F098A, B C, D; IE12-F051A, B; IE12-F065A, B; and ICC065, 067, 068, 07 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-u 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 performs 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 maintenanc The EMEB and :NEL's review of all specified valves indicates that 23 of the 120 valves are required to perfonn a safety-related function. Per Section XI requirements, these 23 valves should be included in the Clinton IST Program and inservice tested. With regard to the remaining 97 valves,18 of them are used only for operating convenience and maintenance and are not required to be tested. The remaining 79 valves perfono certain system functions; however, none of these valves perform an active safety function. Therefore, they are not required to be tested in accordance with Section X . .
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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 commercial date of the Clinton Power Station was April 24, 198 The NRC regulations and ASME Code,Section XI require that the first inservice test for most valves be performed within three months after comercial operation. For Clinton, the first IST occurrvd during July 1987 Ten (10) of these 16 affected valves (Table I, Valve No. I thru 16) were not tested during the first scheduled IST but were all tested during September / October 198 Since then, these valves have been tested in accordance with Section XI requirements and have been verified operable. This infomation is based on verbal input from Region !!I. 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 the allegation had merit at the ti.ne it was made, the early omission of valves
- - from the IST program has not resulted in any real safety impact to the Clinton Plan !
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 commercial operation. A safety evaluation of each valve, missing from the IST for slightly more than a year, is discussed belo Four of these valves (IE12-F051A, B and F065A, B) were designed to be used during the steam condensing mode of the Residual Heat Pemoval System. This 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 mode of operation is not to be used at the Clinton Plan Valve IFC091 is a relief valve. Sectfon 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 concern 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 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 III. Since the recent test has verified the operability of these valves, the safety significance of not testing them more frequently as required by Section XI is minima Based upon the discussion above, the staff finds that approximately 20 percent of the valves in the allegation are required to perform a safete related function while the rest are not. Those valves that are required to be in-service tested are now included in the Clinton IST Program. Furthennore, the staff finds that the safety significance of not performing inservice testing of
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certain affected valves during the first year of plant operation is minima As such, the staff concludes tnat no further action is required and that this allegation is resolve For further information or clarification, please contact me at 492-139 kM Janice A. Stevens, Project Manager Project Directorate III-2 Division of Reactor Projects III, IV, Y and Special Projects, NRR
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cc: R. Cooper, RIII D. Danielson, RI!!
J. Xudrick, PSB J. Huang, MEB
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ATTACHMENT 3 U.S. NUCLEAR REGULATORY C(- "ION
REGION III
With respect to the specific concerns with NRC Inspection Report N /87027(DRS) noted in your June 16, 1988, letter to the NRC, we have reviewed each concern and have the following responses. (The items refer to allegations addressed fa the above inspection report.) Concern
"In respect to item (1) who specifically perfonned the S&L review (A/IR No. 646). Their names, qualifications, previous experience specifically with these same radiological concerns, exactly what documents were reviewed, and specifically what were the documents reviewed for."
Response
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During the course of our inspection, we did not keep a record of the Sargent and Lundy (S&L) personnel who had performed the plant configuration reviews. Also, we did not review personnel qualification The following will provide the extent of the S&L review but we cannot provide you with information pertaining to all the specific documents that were reviewed by S& In order to address this issue thoroughly, we must clarify several points that might be the cause for some confusio Valves identified in Clinton FSAR Table 6.2-47, "Isolation Valve Summary for Linus Penetrating Containment", consist of several types. For this discussion, we will separate the valves into three categories: (1) valves with standard closure times; (2) valves required to close in a specific time because they are included in the off-site dose analysis; and (3)
valves that do not have automatic isolation signals and whose closure times are governed by the system requirement We believe your concern arose from the discussion contained in Document Transmittal Form (DTF) No. JK-1183, which was initiated for the review and approval of v,1ve stroke time changes. Specifically you identified the fo! lowing note on the DTF as a concern:
" Acceptable strnke time reflects system operability considerations. A valve by valve (See DTFs JK-980, JK-1013 and JK-1054 for applicable valves) analysis for radiological considerations was performed by G.E. per IPC request. G.E. evaluation was provided in G.E. responses to DTF's No. JK-980, No. JK-1013 and No. JK-1054. However, the overall impact due to the valve stroke time cha'iges (assuming the valves are closer tn the upper stroke acceptable time) regarding
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consideratior.s for radiological, environmental (e.g., flooding),
etc. should be evaluated by others since this is dependent on plant configuration detail."
The GE review of valves was to determine the effect of the changes on (1) Jf-site dose and (2) system requirements. GE reviewed the valves '
of concern and found them to be acceptable. NRC Inspection Report i No. 50-461/87027(DRS) provides the results of our review of this l matter and also concludes that the valves with the modified stroke '
times are acceptabl The following was identified as a point of concern by you during our i meeting in Knoxville, TN: "However, the overall impact due to ;
stroke time changes. . . should be evaluated by others since this is dependent on plant configuration detail."
The plant configuration review was perfomed by S&L befo.e licensing.
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This review addressed such areas as, 'what type of environment will the valve be in and how will the change affect the corresponding area?'
The Engineering Change Notices (ECN) which were used to effect the ,
modified *troke times were prepared, reviewed and approved by the same level of engineers and executives as would have been used to produce an original design document. The scope of the work, i identification of the valves and the results of the review are clear i and ccncise. The S&L review concluded that "the increased stroking times will not have an impact on the environmental . . . and l radiological concerns".
The stroke time of these valves is considered a design characteristic.
] Design baseline reviews are required to determine if a design change
, affects the design baseline. If the design affects the baseline, changes are required to restore the baseline. Scme of the specific areas that are included in the design baseline reviews are:
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ALARA (radiation considerations) !
Environment !
Flood Protection !
1 Once an ECN is completely reviewed, the approver, who is a Registered j Professional Engineer, is required to sign the documen ;
i Furthennure, the NRC inspectors examined a sample of documents involved :
in S&L's review of the valves in question, including the following:
Action /Information Request (A/IR) 646 and 1022
Engineering Change Notice (ECN) 8693 ,
Field Engineering Change Notice (FECN) 15012 '
Decument Transnittal Form (identifying affected valves) JK-1183
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Our review of these efforts is documented in NRC Inspection Reports No. 50-461/87027(DRS) and No. 50-461/88020(DRS) and they conclude that the valves with the modified stroke times are acceptable. Based on these efforts, no further action is necessar . Concern
"Also, for the record, on a change to a facility affecting almost every safety system in the plant and containment boundary, what is ,
the minimum review required by 10 CFR on such a proposed change to the Technical Specifications and (FINAL)SAR. This was not a change to the preliminary FSAR. Secondly, with all these changes affecting almost all T.S. 4.0.5 surveillances, when did the NRC conduct their review to these proposed changes to the licensing documents and procedures." .
Response The NRC reviews the FSAR and Technical Specifications and issues -
Safety Evaluation reports for all proposed changes to these document The NRC reviewed the Technical Specifications and FSAR prior to licensing to ensure that the containment integrity aspects of the valve stroke times were considered. This review included the new stroke times that Illinois Power Company had determined to 'oe accept-able operating characteristics for the plant. Any changes made to .
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the FSAR or Technical Specifications prior to the commercial licensing of Clinton could be made at the discretion of Illinois Power Company ;
and their Nuclear Steam System Supplier, General Electric, since these changes would be submitted to an' reviewed by the NRC orior to licensin The NRC reviewed and approved the FSAR prior to the commercial licensing of Clinton. The Technical Specifications review began in ,
1984, and they were approved and issued with the low power license in September 198 !
You have previously provided us with information concerning the !
differences between the FSAR and the Technical Specification *
Although there are differer.ces between the documents in the area of Inservice Testir.g. the FSAR will be updated by Illinois Power Company to correspond with the Technical Specifications during their i regularly scheduled submittal which is planned for September 198 lu the meantime, Illinois Power Company has added 23 valves in the Clinton IST program. These valves were identified as erroneously being excluded from the Clinton IST Program during the NRC review of your concern ;
When jou refer to ". . . all these changes affecting almost all T.S. 4.0.5 surveillances", you must note that Technical Specification !
4.0.5, which requires that inservice Testing (IST) be performed, was ;
not applicable until April 198 CFR 50.55a(g)(5)(iv) refers to l l
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the 120-month IST cycle commencing at the time of plant comercial operation. For Clinton, comercial operation was considered to be at the time of issuance of the full power license, which was April 17, 1987. Therefore, the Inservice Testing (IST) program requirements were not required to be met until that tim Our initial review of this concern is documented in NRC Inspection Report Nu. 50-461/87014(DRS). This effort identified a violation where Illinois Power Company failed to verify the remote position indication for certain valves in the Containment Monitoring System and a violation where certain procedures did not clearly delineate quarterly valve stroking requiremen . Additional reviews of this concern are documented in NRC Inspection Report No. 50-461/87027(DRS).
A sample of surveillance and test documentation and procedures were reviewed during this inspection as well as the Illinois Power Company ,
corrective actions to resolve the above violation In addition, in NRC Inspection Report No. 50-461/88020(DRS), Illinois Power Company was issued a violation for failing to include 23 valves in the Clinton IST Program. Many of these valves were included in the information you provided to Region !!! as part of your concern !
The above violations have been adequately addressed by Illinois Power Company end no other problems were identified as a result of our review of your concer . Concern
"In respect to Item (2) what was is the date of the preop test (note-test requirement frequency is once every two years), during the preop was the position indication light observed "open" while at rated flow injecting into the reactor vessel, is it the position of the NRC that ,
in order to meet the full exercise requirements of IWV-3420 the owner must only verify thet the disk will move promptly off its seat or does its ability to meet its safety function need to be verified."
Respense In response to your first question, the dates for the tests that were performed to: (1) verify valve position indication; and (2) fulfill exercise requirements are listed in paragraphs a and o below respec-tively. Also included in these paragraphs is the NRC position on acceptable test methods to fulfiH the requirement Your first issue concerns Valve Position Indicator Verification i defined in Subarticle IWY-3300 of the ASME Code Section XI. The '
second issue is the Valve Exercising Test defined in Paragraph IWV-3522 of the ASME Code Section XI. The two issues are discussed ,
separately below. The valves in question are as follows: '
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Valve Number Valve IE22F005 HPCS Testable Check 1E21F006 LPCS Testable Check 1E12F041A LPCI from RHR A Testable Check 1E12F041B LPCI from RHR B Testable Check 1E12F041C LPCI from RHR C Testable Check 1E51F066 RCIC Testable Che a. Valve Position Indicator Verification The ASME Code requires that valves with remote position indicators be observed at least once every two years to verify that valve operation is accurately indicated. This can be accomplished by simply observing that the disk is open when the contrc1 room light also indicates ope At Clinton, this was accomplished for the valves in a combination of ways: (1) using flow through the valvel (2) direct observation; and (3) as is the case with valve 1E51F006 no testing was required because the valve did not have remote position indicatio In addition to the information stated in NRC Inspection Report No. 50-461/87027(DRS), Paragraph 5.b., other test data was gathered to support the previous conclusion that adequate testing was done to meet the ASME Code Section XI requiret.snts for Position Indication Testing (PIT).
Valve 1E22F005: During the performance of testing defined in Procedure No. PTP-HP-01, "High Pressure Core Spray Pattern Test" on September 10, 1985, the check valve position was verified to correspond with the control room indication by using full system flow through the valv Valve 1E21F006: Check valve position and control room indication were verified to correspond by local observation of disk movement during perfomance of PTP-HP-01. Section 7.2.5 on July 15, 198 Valve IE!2F041A: Check valve position and control room indication ,
were verified to correspond by local observation of disk position !
during maintenance work initiated in September 198 '
Valve 1E12F0418 and C: Check valve position and control room indication were serified to correspond by local observation of :
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disk position by) performance
"(Non-Modulation Air / Hydraulic of Test Procedure Operated No. GTP-11 Valve Generic Test Procedure" on July 16, 198 Valve 1E51F066; As stated in the previously mentioned hRC Inspection Report No. 50-461/87027(DRS), the light indication for the valve does not reflect the valve position and no PIT is required for this valv i
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.' Valve Exercising Test The requirements for testing normally closed check valves are stated in Paragraph IWV-3522 of the ASME Code Section XI. These requirements are stated, in part, below:
"Check valves shall be exercised to the position required to fulfill their function . . . . Valves nonnally closed during plant operation and whose function is to open on reversal of pressure differential shall be tested by proving that the disk moves promptly away from the seat when the closing pressure differential is removed and flow through the valve is initiated, or a mechanical opening force is applied to the dis Confirmation that the disk moves away from the seat shall be by visual observation, by electrical signal initiated by a position indicating device, by observation of substantially free flow through the valve as indicated by appropriate pressure indications in the system, or by other positive means. This test may be made with or without flow through the valv The staff considers that these requirements may be met if any of the following four methods are used as confirmation:
(1) By demonstrating that the valve can pass the maximum, accident-design flow for which credit has been taken in FSAR analyses; (2) By showing that, for the measured flow, the pressure loss through the valve is such that the valve could only be fully opent (3) By using a mechanical exer:iser that can be observed to move through a full stroke; and (4) By partial disassembly of the valve and manually moving the disk through a full strok As stated in NRC Inspection Report No. 50-461/87027(DRS), ,
Paragraph 3.b., preoperational test results were reviewed and it was determined that the requirements of the ASME Code Section XI were met. The testing demonstrated full system design flow through the check valve and not just movement off the sea . Concern
"!P went to run the surveillance as written and innediately a MWR was required to be written in order to obtain an "Open" indication on the ECCS Panel in the control room. The Surveillance Test as written could not verify the re.quirements which responds to my original allegation. Is this your definition of a successful surveillance, attempt to run test, perform maintenance as required until desired
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results can be documented, then take credit for surveillanc After maintenance on the actuator was the surveillance as written ran successfully as required by the current ISI Test Program at Clinto Is the Clinton Plant plant currently taking credit for successful completion of PIT's on these testable check valves after maintenance when no actual surveillance was actually ran after maintenance."
Response i
As indicated in NRC Inspection Report No. 50-461/87027(DRS),we verified that these PIT tests were successfully completed after maintenanc The intent of the testing is to confirm that valves are either operating correctly or to determine the source of the problem, correct it through maintenance and then prove that the corrective :
action was effectiv ,
1 Concern
"Also, for the information of this agency, the reason that these particular PITS are important is not only derived from the ASME
- Testing Requirements but there exists a daily Tech Spec requirement
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to check the ECCS Status Board lights. Credit for this surveillance caa only be taken by observation of the lights if the lights themselves are calibrated. Your report states that 1E51F066 status light will never be required to be tested. Basically, which requirement do you ,
wish to be failing to met, ASME or Tech Specs. Its obvious this
- remote indication has not been tested in the la3t two years and on your advice will never be tested. I would suggest removing the light l if its not required, secondly, is that Clinton Plant currently in violation of their ECCS Status Board verification Tech Spec in lieu of tSe fact its taking credit for surveillances on non calibrated equipment."
Response Control room indication serves several necessary functions, including verification of correct valve lineups as required by Technical Specification The ASME Code Section XI requirements need to be separated from Technical Specification requirements. They might coincide, but conversely, one might have requirements that the other does no The issues will be discussed using the 1E51F066 valve as an example, ,
For PIT testing, the lights in the control room are verified to accurately reflect the actual valve position every two years. Once this has been accomplished, valve lineups can be perfonned from the control room to verify that the applicable valves are in their correct position, as required by the Technical Specificdtion In the case of the 1E51F066 valve, as was previously stated in Concern 3.a. there is no requirement to perform a PIT. The valve
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disk has no position indicating device, and therefore the valve does not have remote position indication. The ASME Code Section XI requires that testing be performed to verify that the position indication reflects the valve position for valves with remote position indicatio However, by design and system requirements, it is not necessary that all valves have remote position indication. It has been established and verified by the NRC that this valve does not require remote position indication. Consequently, there is no requirement for ASME Code Section XI testin The other concera you expressed is that the correct position of valve 1E51F066 (and the other valves) needs to be verified so that the system operability surveillance requirements specified in the Technical Specifications are met. The 1E51F066 valve is in the Reactor Core Isolation Cooling (RCIC) system and the operability requirements for the system are defined in Technical Specification 3/4.7.3 Reactor Core Isolation Cooling system. The surveillance requirements to demonstrate that RCIC is operable are defined in 4.7.3, including the performance of the valve lineup to 4.7.3a.2. which states "Verifying that each valve (manual, power operated or automatic) in the flow path . . . is in its correct position." (This is performed every 31 days, not daily). It is not appropriate to subject valve 1E51F066 to this surveillance requirement because it does not fall into the category of manual, power operated or automatic. Therefore since this require-ment does not apply to this valve there is no violation of the ECCS operability verificatio For the other check valves that were mentioned in Concern 3, no problems exist because the valve lineup can be verified since their remote positions were verifie . Concern
"The issue, did IP Management at the Clinton site have knowledge that certain minimum Technical Specification requirements had not been met prior to accepting their license; did IP Management have knowledge they could not meet the minimum T.S. requirements after accepting their fuel load license; what actions were taken on the part of IP management when said knowledge was known for certain; did IP manage-ment file any reports with this agency when it was known to be in violation of their Containment Integrity. I expect to reflect my allegations to the fuel load date rather than your current polic Looking at your report it might mislead a reviewer to think that the Containment Integrity was met for the PS and CM systems at the time of fuel load (i.e. 2.a. . . almost a year after my allegations, the CM valves are finally being test June 16,1987). What happened from September 26, 1986, was the plant in violation of containment integrity and did operations management initiate an LCO and track according to approved plant procedure If not, this again is another violation of Technical Specification and Test Control. For your office to make a statement "No violation of NRC requirements were identified during the course of this inspection" makes me feel that your office is
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looking at the current condition of the plant (2 years lateri and attempting to justify the actions on the part of the utility today made to allegations two years old. I feel its interesting, two years have passed, the plant operating at 100% power and *his office cannot answer safety concerns affecting the public that have been directed to the commissioner himself. Finally, I honestly can't understand why I am having to suggest these obvious questions that the inspector and consultant either overlooked or avoided."
Response The concerns identified above were difficult to address due to the
- many assumptions that were made by you and included in your questions and statements. First of all, it must be stressed that the NRC
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i understands the time period of concern. The reviews conducted by tha NRC focused on the time period beginning prior to the issuance of the l low power license in September 1986 up to the present.
i With regards to the ECCS check va?ves listed in Ccncern 3, the NRC informed you during our meeting in Knoxville, TN that these valves are containment isolation valves. However, this did not mean that the licensee violated Technical Specification 3/4.6.4 "Containment Isolation Valves" because these valve: are not currently listed as
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containment isolation valves in the Technical Specifications. Below 4 is an explanation of circumstances that allowed the plant to operate
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as they did and the actions that are being taken to resolve the matte Section 6.2.4 of the Clinton FSAR defines the containment isolation
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system. The containment isolation system has the function to i'olate
the containment in the event of accidents or other conditions which can lead to excessive releases of radioactivity to the environnen This is accomplished by complete isolation of system lines per.etrating the primary containment.
- Valves 1E12F041A and B located in the LPCI "A" and "B" lines respectively, need not be considered Containment Isolation Valves (CIV's). These lines have inboard remote-manually controlled, motor-operated, normally closed 1 CIV's and thus satisfy the requirements of General Design Criteria 55,
56, and 57. These testable check valves are therefore only considered J
Pressure Isolation Valves (P!V) and have been identified and tested as
suc However, a different situation exists for valves 1E22F005, IE51F066, 1E21F006, and IE12F041C. At the time the FSAR was initially approved by the NRC, these valves were excluded on the basis that isolation is
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provided by one motor-operated valve and the system is a closed system outside of containment, which provides redundant isolation. This design
was determined to be in accordance with the NRC's Standard Review Plan (NUREG-0800).
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Illinois Power Company did not include the four testable check valves in the FSAR that was submitted to the NRC prior to issuance of the low power license. At that time, the NRC accepted Illinois Power ,
Company's tasis for the exclusion of these valves from the FSAR and l approved the plant Technical Specifications, which for these valves, I agreed with the FSAR. Therefore, Illinois Power Company was in com-pliance with NRC regulations and there was no violation of containment j integrity requirements. On this basis there were no LC0 violations i t or reporting requirements as you sugges dubseqt.ently, the Technical 3pecifications were written to comply with the FSAR, excluding the forementioned testable check valves as CIV' However, af ter further review conducted in response to your allegations, the NRC has concluded that the four testable check valves, 1E22F005, IE51F066, 1E21F006, and 1E12F041C should be considered as CIV' These check valves should be included in Table 3.6.4-1 of the Containment Isolation Valve section in the plant Technical Specifications and correctly addressed in the FSA In a letter dated April 18, 1988, Illinois Power Company provided a calculation which attemptea to correlate leakage rates measured with water at 1000 psig with leakage rates measured with air at 9.0 psi The NRC staff's review of the Illinois Power Company proposal to use this water-to-air corrillation and thereby permit continued testing of the subject valves with water, concluded that the check valves should be Type C tested with air or nitrogen in accorda'ce with Appendix J to 10 CFR Part 5 Based on our recent review, the NRC considers this issue reso ed when the Technical Specifications and FSAR are revised to reflect the changes mentioned abov With respect to the 30 Process Sampling (PS) and Containment Monitoring (CM) valves, the letter written to you by Mr. H. J. Hiller of NRC Region 111 un August 12, 1987, addressed this issue in our response to Concern 4. Paraphrasing this letter, it was found that violations in fact did occur. At the time of the NRC inspection in March 1987 the test deficiencies were identified to Illinois Power Company. Tests performed subsequently showed no equipment deficiencies, and therefore there is no reason to believe that containment integrity at Clinton wds ever in jeopard With respect to the PS, CM, and testable check valve containment integrity requirements, to the best of our knowledge Illinois Power Company management did not have knowledge that Technical Specification requirements had not been met prior to accepting the low power licens In the case of the CM and PS valves, violations were issued for missed testing requirements and subsequent testing was conducted when this matter was brought to their attention. For the testable check valves, no requirements were violated and therefore, there was no reporting requiremen _ _ _ _ _ _ - _ _ _ _ _ _ .
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u Your allegations were substantiated with regard to these issues and generic corrective actions have been contpleted by Illinois Pcwer Company, with additional corrective actions under way. On this ba:,is, your allegations are considered resolve .