05000440/LER-1995-001, :on 950307,identified Condition Which Could Allow Primary Containment Leakage Rates to Exceed TS Limit. Caused by Inadequate Design of Airlock Lcs.Initiation of Permanent Design Change
| ML20082E673 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 04/06/1995 |
| From: | Conran D, Shelton D CENTERIOR ENERGY |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| LER-95-001, LER-95-1, PY-CEI-NRR-1938, NUDOCS 9504110300 | |
| Download: ML20082E673 (6) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2) 10 CFR 50.73(a)(2)(viii)(8) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(x) |
| 4401995001R00 - NRC Website | |
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a ENE.GY PERRY NUCLEAR POWER PLANT Mail Address:
DOnSid C. SheltOn SENIOR VICE PRESIDENT G CENTER ROAD RR, O 10 44081 PERRY, OHIO 44081 NUCLEAR (216) 259-3737 April 6, 1995 PY-CEI/NRR-1938L United States Nuclear Regulatory Commission Document Control Desk Washington, D.C.
20555 Perry Nuclear Power Plant Docket No. 50-440 LER 95-001 Gentlemen:
Enclosed is Licensee Event Report 95-001 concerning Loss of PNPP's Airlock Leakage Control System Resulted in the Potential to Exceed TS Containment Leakage Rates.
If you have questions or require additional information, please contact Mr. James D. Kloosterman, Manager - Regulatory Affairs at (216) 280-5833.
Very truly yours,
'l CRE sc
Enclosure:
LER 95-001 cc NRC Project Manager NRC Resident Inspector Office NRC Region III 1
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Perry Nuclear Power Plant. Unit 1 05000 440 10F 5 N I*3 Loss of PNPP's Airlock Leakage Control System Resulted in the Potential to t,xceed TS Containment Leakage Rates EVENT DATE (5)
LER NUMDER (6' REPORT NUMBER (7)
OTHER FACILITIES INVOLVED (8)
MONTH DAY YEAR YLAR MONTH DAY YEAR NUMBER NUMBER 05000 F ACIU1Y NAME DOCKET NUMBER
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03 QJ 95 95 001 00 03 07 95 05000 OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR O (Check one or more)(11)
MODE (9)
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20.402(b) 20.405(c) 50.73(a)(2)(iv) 73.71(b)
POWER 20.405(a)(1)(i) 50.36(c)(1)
X 50.73(a)(2)(v) 73.71(c)
LEVEL (10) 100 20 405(a)(1)(ti) 50.36(c)(2) 50.73(a)(2)(vii)
OTHER 20.405(a)(1)(iii) 50.73(a)(2)(i) 50.73(a)(2)(viii)(A)
IhP'cdv 'n Akimi NAi 20.405(a)(1)(iv) 50.73(a)(2)(ti) 50.73(a)(2)(viii)(8) r 20.405(a)(1)(v) 50.73(a)(2)(iii) 50.73(a)(2)(x)
LICENSEE CONTACT FOR THIS LER (12)
NAuc TELEPHONE NUMBER lIncluce Area Cocie)
Derek W. Conran. Compliance Encineer (216) 280-5274 COMPLETE ONE LINE FOP EACH COMPONENT FAILURF. DESCRIBED IN THIS REPORT (13)
CAUSE
SYSTE M COMPONENT MANUFACTURER
CAUSE
SYSTEM COMPONENT MANUFACTURER SUPPLEMENTAL REPORT EXPECTED (14)
EXPECTED MONTH DAY
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,o lW yes. compiete EWECTED SUBMISSON DATO y
DATE (15)
ABSTRACT (Umit to 1400 spaces, i e, approximately 15 single-spaced typewntten lines) (16)
On March 7, 1995, engineering personnel identified a condition in which an open or leaking Containment Airlock inner door concurrent with a LOCA could alone allow Primary Containment Leakage rates to exceed the Technical Specification 3.6.1.2.b limit of 0.60 L.
In addition on March 9, 1995, engineering personnel identified a separate condition in which the loss of Division I power concurrent with a LOCA could in the past have allowed Secondary Containment Bypass Leakage rates to exceed the Technical Specification 3.6.1.2.d limit of 0.0504 L,.
The root cause of the condition which could have alone allowed Primary Containment Leakage rates to exceed the Technical Specification limit of 0.60 L, is attributed to inadequate design. The root cause of the condition which could have allowed Secondary Containment Bypass Leakage rates to exceed the Technical Specification 3.6.1.2.d limit of 0.0504 L, is attributed to inadequate configuration control.
Corrective actions taken to maintain leakage rates within Technical Specification requirements included isolation of the Containment Airlock Leakage Control System (LCS) and initiation of a permanent design change to PNPP's LCS which vill ensure leakage rates are maintained within acceptable values with respect to Technical Specification limits.
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I.
Introduction
On March 7, 1995, engineering personnel identified a condition in which an open or leaking Containment [NH] Airlock [AL] inner door [DR] concurrent with a LOCA could alone allow Primary Containment Leakage rates to exceed the Technical Specification 3.6.1.2.b limit for measured combined penetration leakage of 0.60 L,.
Thin potential leakage path from primary containment to secondary containment was not adequately accounted for with respect to Primary Containment Leakage totals. Therefore, a four hour notification was made to the NRC at 1854 hours0.0215 days <br />0.515 hours <br />0.00307 weeks <br />7.05447e-4 months <br /> as required by 10CFR50.72(b)(2)(iii)(C). This requirement describes a loss of safety function needed to control the release of radioactive material.
As part of the ongoing investigation of a correlated concern identified in December 1994 and the above concern, engineering personnel also confirmed on March 9,1995, a condition in which the loss of Division I power concurrent with a LOCA could, in the past, have allowed Secondary Containment Bypass Leakage rates to exceed the Technical Specification 3.6.1.2.d limit of 0.0504 L,.
A four-hour notification was made to the NEC at 1821 hours0.0211 days <br />0.506 hours <br />0.00301 weeks <br />6.928905e-4 months <br /> as required by 10CFR.72(b)(2)(iii)(C). This requirement describes a loss of safety function needed to control the release of radioactive material.
As both of these conditions relate to the design of the Containment Airlock Leakage Control System, they are being reported together in accordance with the requirements of 10CFR50.73(a)(2)(v)(C).
At the time of discovery of these events, the reactor was in Operational i
condition 1, Operations, at 100% power with reactor pressure approximately 1025 psig.
II.
Description of Event
On December 7, 1994, Perry Nuclear Power Plant (PNPP) identified and began investigating a potential concern regarding loss of the Containment Personnel Airlock Leakage Control System (LCS) due to a single failure of its associated
)
Division I power. PNPP's Airlock LCS directs leakage past the inner seals of the i
four airlock doors (two airlocks each with an inner and outer door) through associated solenoid valves to the containment annulus (Secondary Containment).
The four solenoid valves are normally closed, but receive signals to open on a LOCA. However, these four valves all receive power from Division I.
Because these valves need power to reposition, a single failure of the Division I power supply following a LOCA could render PNPP's Airlock LCS inoperable and airlock leakage could become Secondary Containment bypass leakage.
During this initial evaluation, it was determined that loss of Division I power would not immediately result in exceeding Technical Specification leakage rates due to current known plant bypass leakage rates.
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., ARC w, => c n on March 7, '1995, < luring investigation of the concern addressed above, i
engineering personnel identified a separate condition in which an.open or leaking Containment Airlock inner door concurrent with a LOCA could alone allow Primary Containment Leakage rates-to exceed the Technical Specification 3.6.1.2.b limit for measured combined penetration leakage rate of 0.60 L. It was discovered that when an inner airlock door is opened or an inner aillock inner seal fails, J
the LCS piping limited by a one eighth inch orifice is exposed to containment i
atmosphere. Also following a LOCA, the associated downstream solenoid valve would receive a signal to open. An engineering calculation verified'that this leakage path alone could contribute greater than 0.60 L, to PNPP's treated leakage totals.
On March 9, 1995, engineering personnel confirmed that during at least one previous occasion prior to Refueling Outage 4, a condition existed in which the
- - loss of Division I power concurrent with a LOCA could have alloved Secondary Containment Bypass Leakage rates to exceed the Technical Specification 3.6.1.2.d limit of 0.0504 L. This was confirmed by adding known (measured) secondary
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containment bypasI leakage rates from past tests to the known air lock leakage that could become bypass leakage if the LCS was inoperable.
3 Interim corrective actions taken with respect to both of the conditions addressed l
above included issuance of a standing instruction requiring station personnel to isolate the Containment Airlock Leakage Control System (LCS) if an inner airlock door became inoperable.
Subsequently, the normally open manual valves upstream of the inner doors' solenoid valves were closed to maintain continuous isolation of the penetration. Previous test data provided assurance that these manual valves provided negligible leakage.
III. Cause of Event
The root cause of the condition in which an open Containment Airlock inner door concurrent with a LOCA could alone allow Primary Containment Leakage rates to-i exceed 0.60 L, is attributed to inadequate design.
PNPP's Airlock LCS was installed as designed and described in the USAR. However, this design allows a single leakage path that alone could have exceeded the measured combined penetration leakage rate of 0.60 L value during a design basis accident, if the door happened to be open concurrently.
Because this has existed since initial licensing, it is difficult to determine why this issue was not previously identified and incorporated into the original design. A review of design documentation could not identify why this leakage path was not designed to limit leakage significantly belov 0.60 L, during a design basis accident.
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5 uxT tu m sw. a wea. we aamw ua, orwc mm on The root cause of the condition in which the loss of Division I power concurrent with a LOCA vould have allowed Secondary Containment Bypass Leakage rates to exceed 0.0504 L, is attributed to inadequate configuration control.
PNPP's Airlock LCS was originally designed such that the solenoid valves were normally open. In this configuration, loss of the Division I power supply would not have affected PNPP's Airlock LCS from performing its intended function.
However, on October 12, 1985, plant drawings were revised to reflect the solenoid valves as normally closed instead of normally open. No engineering analysis or authorization document could be found to establish why the valve positions were changed.
In addition, it could not be determined why the solenoid valves associated with the Inner / Outer Door Airlock LCS vere not powered from separate divisions when this change was performed. Analysis prior to changing the configuration of PNPP's Airlock LCS should have identified the need for integration of divisional power concepts into the design change if the valves were to be maintained normally closed.
IV.
Safety Analysis
The personnel access airlocks are velded steel assemblies with double doors, each equipped with double inflatable seals and designed to provide the capability of leak rate testing the airlock between doors and the cavity between door seals at a pressure of P, (7.80 psig). The airlock doors are designed ac pressure seating doors.
Containment atmosphere could potentially bypass Secondary containment and the Annulus Exhaust Gas Treatment System (AEGTS) by leaking past the double seals on each door of the personnel airlocks into the Intermediate Building. The personnel air lock leakage control system was designed to eliminate this potential bypass leakage path past the double seals on each door.
Piping was routed from between the airlock door seals to the annulus (Secondary Containment). The AEGTS is designed to maintain the annulus at a vacuum. Any post-LOCA containment atmosphere in the airlock leakage control lines was designed to flow into the annulus.
The performance objective of the shield building is to collect and retain any fission product leakage from the containment vessel into the annulus during and following a design basis accident and, in conjunction with the AEGTS, process and release the fission products to the environs in a controlled manner. This release is accomplished such that the resultant offsite doses to the general public are within the values given in 10 CFR 100 and the doses to the control room operators are within the values given in 10 CFR 50, General Design Criterion 19.
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i Tsui or mm som a reemo, a.. coma,n eco.n or unC ram, aesy on Technical Specification limitations on containment leakage rates ensure that the total containment leakage volume vill'not exceed the value assumed in the-accident analyses at the peak accident pressure of 7.80 psig, P. As an added conservatism, the measured overall integrated leakage rate is f0rther limited to less than or equal to 0.75 L (including the 0.60 L measured combined penetration. leakage rate) duIing performance of the, periodic tests to account for possible degradation cf the containment leakage barriers between leakage tests.
Because the potential leakage rates identiff.ed during this investigation could have exceeded the Technical Specification Containment Leakage Rates, these events are considered to be safety significant.
V.
Similar Events
Previous events concerning problems with containment and dryvell air locks have been documented by LERs87-061 (problem with locking pins),88-032 (problems with 3-vay ball valves and a blistered seal),88-035 (problem with a mechanical interlock),90-007 (problems with a ruptured seal combined with personnel errors), and 92-020 (problems with improper labeling of plant equipment and personnel error). These previously reported concerns and their associated corrective actions could not have been reasonably expected to prevent the issues discovered in March, 1995.
VI.
Corrective Actions
Interim corrective actions taken with respect to both of the conditions addressed-above included issuance of a standing instruction requiring station personnel to-isolate the Containment Airlock Leakage Control System (LCS) if an inner airlock door became inoperable. Subsequently, the normally open manual valves upstream of the inner doors' solenoid valves were closed to maintain continuous isolation of the penetration.
Previous test data was found which provided assurance'that these manual valves provided negligible leakage. Completion of this interim corrective action assures compliance with Technical Specifications; however in i
addition, engineering personnel are currently planning a permanent design change to PNPP's Airlock LCS.
Energy Industry Identification System (EIIS) codes are identified in the text as [XX].
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