ML20116M056
| ML20116M056 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 08/09/1996 |
| From: | Dawn Powell Public Service Enterprise Group |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| LR-N96211, NUDOCS 9608190281 | |
| Download: ML20116M056 (8) | |
Text
"OPSEG Public Service Electric and Gas Cornpany P.O. Box 236 Hancocks Bridge, New Jersey 08038-0236 Nuclear Business Unit AUG O 91996 LR-N96211 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Gentlemen:
RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION DRYWELL TO SUPPRESSION CHAMBER VACUUM BREAKER TEST SCHEDULE HOPE CREEK GENERATING STATION FACILITY OPERATING LICENSE NPF-57 DOCKET NO. 50-354 The purpose of this letter is to provide additional information relative to the Drywell to Suppression Chamber Vacuum Breaker Test Schedule.
This letter provides the responses to questions that were discussed during a teleconference between PSE&G and the NRR Hope Creek Project Manager on June 20, 1996. of this letter contains the questions asked by the NRC, along with PSE&G's responses.
This information supports the previously submitted test schedule.
The trending program that has been put into place will enable future failures to be predicted.
PSE&G believes that had this trending program been
-put into place previously, the prior failures could have been predicted and prevented.
Should you have any questions or comments on this transmittal, do not hesitate to contact us.
- ncerely,
/
l D vi R.
Powell Manager -
Licensing and Regulation Ok Attachment f
9608190281 960809 i
PDR ADOCK 05000354 p
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' Document Control Desk AUG 0 91996 LR-N96211 C
Mr. H. J. Miller, Administrator - Region I U.
S.
Nuclear Regulatory Commission 475 Allendale Road
. King of Prussia, PA 19406 Mr. D. Jaffe, Licensing Project Manager - Hope Creek U. S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Mail Stop 14E21 Rockville, MD 20852 Mr. R.
Summers USNRC Senior Resident Inspector (X24)
Mr. K. Tosch, Manager IV Bureau of Nuclear Engineering 33 Arctic Parkway CN 415 Trenton, NJ 08625 i
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' Document Control Desk AUG O 91996 LR-N96211 RAR/tcp BC Senior Vice President - Nuclear Engineering (N19)
General Manager - Hope Creek Operations (H07)
Director - QA/NSR (XO1)
Operations Manager - Hope Creek (H01)
Manager
. Nuclear Safety Review (N38) l Manager - Nuclear Business Relations (N28)
Onsite Safety Review Engineer - Hope Creek (Hil)
Station Licensing Engineer - Salem (XO9)
Station Licensing Engineer - Hope Creek (XO9)
General Solicitor, R.
Fryling, Jr. (Newark, SG)
Mark J. Wetterhahn, Esq.
Records Management (N21)
Microfilm Copy File Nos.
1.2.1, 3.7 (HC LER 354/95-031) l l
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l ATTACHMENT 1 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION LR-N96211 HOPE CREEK GENERATING STATION DOCKET NO. 50-354 NRC OUESTION I.A:
Where is the testing described in the UFSAR and how did previous I
testing differ?
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PSEEG Response:
The Drywell to Suppression Chamber Vacuum Breaker testing requirements are described in UFSAR Section 6.2.6.5, "Special Testing Requirements."
This section specified the following:
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The valve and system orientations are the same as those for the Type A test, except that any open paths l
equalizing drywell and suppression chamber pressure j
during the Type A test are closed.
The drywell atmosphere is allowed to stabilize for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after attaining test pressure.
The leakage rate is calculated from pressure and l
temperature data, taking elapsed time into account.
During the surveillance tests that were conducted prior to and including the November 10, 1995 surveillance, a pressure decay test methodology was utilized.
However, the above three UFSAR requirements were not met.
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An UFSAR Change Notice has been generated in accordance with l
10CFR50.59 and has been approved by Hope Creek.
This Change l
Notice removed the requirement for the Type A test valve and l
system orientations and removed the type of test method.
However, it retained the stabilization period and the temperature i
calculation requirements.
l The basis for this change was that the change reduces the extent of the containment isolation that is required to perform an accurate leak rate measurement of the leakage through the vacuum relief valves between the Drywell and the suppression Chamber.
Any system that penetrates the Drywell or Suppression Chamber, and has the capability to produce volumetric or temperature changes must be controlled to obtain valid test results.
Systems i
l such as Residual Heat Removal and Drywell Chilled Water would be required to be in service to provide temperature stability over the test duration.
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i ATTACKMENT 1 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION LR-N96211 HOPE CREEK GENERATING STATION DOCKET NO. 50-354 In addition, the surveillance procedure has been revised to l
incorporate a bypass leakage methodology and to perform the surveillance test in accordance with the revised UFSAR.
NRC OUESTION I.B:
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What did the review of previous test data reveal?
l PSEEG Response:
l During the past surveillance tests, the three UFSAR requirements discussed above were not met.
The System Manager reviewed the previous tests and determined that the Drywell ventilation system i
provides adequate air volume mixing through fixed fan and coil l
system configuration during the performance of the test; no transients that could have non-conservatively affected the results occurred during any of the tests; and there were no changes in the system configuration during any of the tests.
Therefore, no significant temperature changes would have occurred over the 10 minute test period.
l NRC OUESTION I.C:
Confirm that LER 92-006 is the only previous report of failures.
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PSEEG Response:
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The previous Suppression Chamber to Drywell Vacuum Breaker surveillances were reviewed.
The failure reported in LER 92-006 l
is the only failure prior to LER 95-031.
Following the failure reported in LER 92-006, PSE&G submitted letter NLR-N92076, which established a test schedule.
The NRC provided its concurrence with this schedule by letter dated July 2, 1992.
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ATTACKMENT 1 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION LR-N96211 HOPE CREEK GENERATING STATION DOCKET NO. 50-354 NRC OUESTION II.A:
What was the cause of the failures reported in LER 95-031?
PSEEG Response:
The cause of the equipment failures was misalignment of the pallet to its valve body.
Evidence of this was shown in the
'C',
'F',
and
'H' valves where the seal was not centered on its body.
In the case of the
'C' and
'H' valves, test pressure (0.8 psid) i could be achieved, but not maintained, in the test stand.
These valves were repaired by replacing the seals, making valve pallet to pivot block adjustments to re-center the pallet, and drilling new pivot block holes to maintain the pallet centered in a stress free condition.
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'F' valve could not achieve test pressure in the test stand.
Upon disassembly of the pallet, one of the hinge bolts broke, providing indication that the valve was misaligned and in a stressed condition.
The
'F' valve pallet was re-centered on its valve body, re-assembled with new pivot block holes drilled, and successfully retested prior to being placed back onto the ring header.
l When LER 95-031 was submitted on December 11, 1995, PSE&G l
suspected that the
'G' valve was a major contributor to the surveillance test failure.
This was due to the results of a qualitative " paper test."
However, the surveillance test that was conducted on the
'G' valve while it was in the test stand proved the valve to be satisfactory.
NRC OUESTION II.B:
Is there any commonality with the LER 92-006 failure,s?
PSEEG ResDonse:
There is commonality between the 1992 and 1995 failures with regard to the
'F' and
'H' valves.
As described in LER 92-006, these valve pallets were found misaligned, and the alignment pins for the hinge arm were found sheared upon disassembly.
Corrective actions in 1992 included replacement of the hinge alignment pins, and adjustment of the pallet to achieve proper sealing.
PSE&G believes that the new pivot block holes will remove the stress on the pallet and resolve the issue.
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ATTACHMENT 1 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION LR-N96211 HOPE CREEK GENERATING STATION DOCKET NO. 50-354 NRC OUESTION III.A:
When was the final testing conducted?
PSEEG Response:
The testing was performed satisfactorily on March 13, 1996.
NRC OUESTION III.B:
What were the test results?
PSEEG Response:
0.04 inches of water per minute for a 10 minute period.
NRC OUESTION III.C:
What were the acceptance criteria?
PSE&G Response:
0.24 inches of water per minute for a 10 minute period.
NRC OUESTION IV:
Does the current surveillance test require a shutdown?
PSEEG Response:
The current surveillance test procedure is required to be performed in OPERATIONAL CONDITION 4, (COLD SHUTDOWN).
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ATTACHMENT 1 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION LR-N96211 HOPE CREEK GENERATING STATION DOCKET NO. 50-354 l
MFC OUESTION V:
i Describe the trending program for future testing.
l PSE&G Response:
A procedural step was added to forward the completed procedure to the System Manager for review and trending of test results.
The system Manager is expected to evaluate the data within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of test result receipt.
This will allow the System Manager to examine the leak rate versus test data, extrapolate the information for the next 18 months, and recommend valve repairs or inspections, if necessary, at the beginning of a refuel outage.
Corrective maintenance will be performed on individual valves as required.
The purpose of the trending program is to be able to predict if l
the vacuum breakers will successfully perform their safety function over the next operating cycle.
A review of the results of the surveillances prior to the last failure indicates that a i
step change in each consecutive set of test results existed (see chart below).
If the trending program had been implemented, the failure could have been predicted and corrective actions implemented to preclude the failure.
DATE RESULTS 5/30/92 0.03 11/3/92 0.08 3/4/94 0.15 i
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