ML20070H744
| ML20070H744 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 03/08/1991 |
| From: | Crimmins T Public Service Enterprise Group |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| GL-89-10, NLR-N91043, NUDOCS 9103150011 | |
| Download: ML20070H744 (5) | |
Text
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Pubbe Service nectne and cas
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Thomas M. Crimmins, Jr.
Petec bm ce * < %c ami Gas Commy P.O !W P36 Hancock Bnage NJ 08038 609 339 4700 l
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MAR 0 8 1991 NLR-N91043 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Gentlemen:
120-DAY REPORT GENERIC LETTER 89-10, SUPPLEMENT 3 HOPE CREEK GENERATING STATION DOCKET NO. 50-354 In accordance with the provisions of 10CFR50.54 (f), Public Service Electric and Gas Company, (PSE&G) hereby provides the following 120-day report in response to Supplement 3 of NRC Generic letter (GL) 89-10 which we received on November 9,
1990.
CONCLUSION We have completed our evaluation regarding the "as-is" capability of the Hope Creek HPCI, RCIC ana RWCU system motor-operated isolation valves (MOV).
We conclude that these MOVs, as designed, tested and maintained, are capable of performing their safety function.
This conclusion is based on a comprehensive evaluation of thrust requirementa in light of design base parameters, standard methodologies and the latest diagnostic test results (See Attachment 1).
INEL TEST RESULTS In general, the INEL test results demonstrate that repeated stroking of an MOV at, and above, maximum design differential pressure, causes damage to 7alve internals which increases the thrust requirements to close the valve on subsequent cycles.
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MAR 0 8 1991 Docur nt Control Desk 2
i NLR-N91043 Prior to the flow interruption tests, thE INEL test valves were stroked numerous times at differential pressures up to 1700 psid.
This raises two general applicability issues.
First, because the test valves were not inspected for damage between the " qualification" tests and the flow interruption tests, it is impossible to determine the poir4 in time when valve internal damage first occurred.
As a result, we have no basis for concluding that the Hope Creek valves, which have not been exposed to potentially damaging test conditions, would respond similarly in a flow interruption scenario.
Secondly, because our design basis postulates one occurrence of maximum differential pressure for a motor operated isolation valve in a high energy applfcation, these valves are expected to close once in the avent of a guillotine break.
Consequently, we do not believe that the gradual performance degradation of the test valves as documented is an adequate basis for concluding that the Hope Creek isolation valves would fail on first stroke.
In addition to the above general issues, we have several specific concerns regarding the applicability of the INEL test results.
1.
The Hope Creek HPCI, RCIC and RWCU system isolation NOVs are 10",
4" and 6" (respectively) Anchor Darling 900# ANSI Class flexible wedge gate valves.
Because none of the valves tested by INEL were 4" Anchor Darling gate valves, no basis exists for application of the INEL test results to the Hope Creek RCIC MOVs.
2.
Given the sample size and the utility population, the claim for representation is statistically unfounded.
3.
The lack of a QA-type assurance program for the INEL testing program detracts from its effectiveness and its credibility.
We have found such programs to be very useful over the years and strongly recommend their implementation.
4.
Supplement 3 states that elements that a licensee may consider in determining applicability of the INEL test results include internal dimensions.
However, detailed dimensional characteristics (i.e., mean seat diameter and seat face width) were'not available from published INEL reports.
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MAR 0 81991 Document Control Desk 3
NLR-N91043 5.
Based on discussions with Anchor Darling, we conclude that the 6" and 10" valves tested were comparable to the Hope Croek RWCU and HPCI isolation valves.
However, a review of the " qualification" test results for these valves indicates that both valves sustained damage-to their internals during this testing.
Also, Anchor Darling has advised that the-valve refurbishment performed on the 6" valve between the-Phase 1 and Phase 2 tests was not representative of an Anchor Darling valve.
Specifically, the stellite repair and disc machining were considered "Below Acceptable Standards" for an Anchor Darling valve.
-Consequently, we do not consider the Phase 2 test data to be applicable to the Hope Creek RWCU and HPCI isolation valves.
6.
Step 25 closely parallels the isolation function these valves are designed to perform.
We feel that this Step should have been performed first, immediately followed by valve inspection and refurbishment, as required.
Absent any intra-test inspection and the lack-of specificity for when maximum stem force was recorded, the data documented for this Step is inconclusive.
HOPE CREEK BYPASS CIRCUITRY The Hope Creek design incorporates a close torque switch bypass circuit for the six MOVs in question. 'This circuit-automatically cuts out the close torque switch until the valve reaches the 5% open position..The 5% open position assures isolation through disc to seat overlap.
.Between l
100% and 5% open, the valve will: continue to close-against resistive forces up to the MOV. stall thrust.
This circuit provides additional assurance of valve closure upon receipt of an isolation signal.
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4 Document Control Desk
'4-MAR 0 81991-NLR-N91043 COMMENT' l
We do consider the concern underlying Generic Letter 89-10 to be serious and worthy of careful engineering assessment, and we are aggressively pursuing scheduled resolution 4-foriour!
units.
However, we do not feel that this iccue.was properly handled by the Staff.
We1believe that additional and better test data should have been developed prior to issuancetof--
Supplement 3 to GL 89-10.
We also-believe that GL 89-10 improperly addresses.the topic of^ valve mispositioning-in
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light of single failure criteria and requires extensive discussion with the-industry, f
Sincerely,
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e Attachment 1
C Mr.
S.
Denbek Licensing Project-Manager Mr.-T. Johnson Senior Resident Inspector Mr. T. Martin, Administrator Region I Mr. Kent Tocch, Chief New Jersey Department of: Environmental Protection Division:of Environmental QualityJ Bureau of Nuclear Engineering-CN 415 Trenton, NJ 08625-
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THRUST EVALUATION
SUMMARY
l SlSIEH VALVE LATEST DIAG.
DESIGN' PERCENT ISOLATION IDENT.
TEST THRUST BASE SEATING THRUST POSITION DATA (LBf)
THRUST (LBf)
MARGIN RWCU INBOARD 1BGV-001 19,179 12,597
'52%
OUTBOlaD 1BGV-002 17,872 12,597 42%-
B.Cl.C INBOARD 1FCV-001 7,230
-6,567 10%=
OUTBOARD 1FCV-002 7;500 6,507 14%
HPCI INBOARD 1FDV-001 30,150 26,508 14%
OUTBOARD 1FDV-002 30,100 26,508 14%'
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Refi NLR-N91043 STATE OF NEW JERSEY
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COUNTY OF SALEM
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T. M. Crimmins, Jr., being duly sworn according to law deposes and says:
I am Vice President - Nuclear Engineering of Public Service Electric and Gas Company, and as such, I find the matters set l
forth in our letter dated March 8, 1991 concerning the Hope Creek Generating Station, are true to the best of my knowledge, information and belief.
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l Subscribed and Sworn to before me this [d/
day of '[4/A//
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1991
/ dana /?sf L Notary PU'.lic of New Jersey DF.Lonl8 0 HA000s Notary Punc of NewJersey My Commission expires on