View source for 05000219/LER-1994-001, :on 940113,discovered That Existing CS Systems 1 & 2 Minimum Recirculation Piping Configuration Did Not Meet Seismic & Thermal Expansion.Caused by Inadequacy Piping Design.Changed Configuration to Meet Criteria

{{Adams
| number = ML20072G723
| issue date = 08/16/1994
| title = :on 940113,discovered That Existing CS Systems 1 & 2 Minimum Recirculation Piping Configuration Did Not Meet Seismic & Thermal Expansion.Caused by Inadequacy Piping Design.Changed Configuration to Meet Criteria
| author name = Barton J, Schwartz S
| author affiliation = GENERAL PUBLIC UTILITIES CORP.
| addressee name = 
| addressee affiliation = NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
| docket = 05000219
| license number = 
| contact person = 
| document report number = C321-94-2129, LER-94-001, LER-94-1, NUDOCS 9408250076
| document type = LICENSEE EVENT REPORT (SEE ALSO AO RO), TEXT-SAFETY REPORT
| page count = 4
}}
{{LER
| Title = :on 940113,discovered That Existing CS Systems 1 & 2 Minimum Recirculation Piping Configuration Did Not Meet Seismic & Thermal Expansion.Caused by Inadequacy Piping Design.Changed Configuration to Meet Criteria
| Plant = 
| Reporting criterion = 10 CFR 50.73(a)(2)(iv), 10 CFR 50.73(a)(2)(v), 10 CFR 50.73(a)(2)(vii), 10 CFR 50.73(a)(2)(i), 10 CFR 50.73(a)(2), 10 CFR 50.73(a)(2)(x), 10 CFR 50.73(a)(2)(ii)
| Power level = 
| Mode = 
| Docket = 05000219
| LER year = 1994
| LER number = 1
| LER revision = 0
| Event date = 
| Report date = 
| ENS = 
| abstract = 
}}

=text=
{{#Wiki_filter:s GPU Nuclear Corporation
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:::e;t ; 88 Forked River, New Jersey 08731-0388 609 971-4000 Writer's Direct Dial Number:
August 16,1994 C321-94-2129 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

==Dear Sir:==
Subject: Oyster Creek Nuclear Generating Station Docket No. 50-219 Licensee Event Report 94-001 Enclosed is Licensee Event Report 90001, Revision 1. Changes have been indicated by a bar in the right hand margin.
If there are any questions please contact Terry Sensue at 609.971.4680.
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Jo in J. Barpn V'ce President and Director yster Creek JJB/JJR Enclosure cc:
Administrator, Region 1 Senior Resident Inspector 7%
Oyster Creek NRC Project Manager e
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'* S d C O f) npli Ni.cipat C.orporaf ton is a subsidiary of General Puunc Uti ibes Corporabon l
9408250076 940816 PDR ADDCK 05000219 S
PDR

NRC FORM 36Mt (5-92)
U.S. NUCLEAR REGULATORY CGellSSIGH APPR Y
LICENBEE EVENT REPORT (LER)
R 5 i/9 FACILI1Y NAME (1)
DOCKET NUMBER (2)
PAGE (3)
Oyster Creek, Unit 1 05000219 1 OF 3 11TLE (4)
Core Spray Piping Exceeding the Code Allowable Stresses Due to Original Design Deficiency EVENT DATE (5)
LFR NIDeBER (6)
REPORT DATE (7)
OTHER FACILITIES INVOLVED (8)
SEQUENTIAL REVISION MONTH DAY YEAR YEAR M3 NTH DAY YEAR NUMBER NUMBER 01 13 94 94 001 01 08 16 94 DOCKET NUMBER THIS RNT IS m!Um MNT TO M RMIRNNTS OF 10 CER D ( N ck e or more) ( M)
OPERATING y
C)DE (9) 20.402(b) 20.405(c) 50.73(a)(2)(iv) 73.71(b) 20.405(a)(1)(1) 50.36(c)(1) 50.73(a)(2)(v) 73.71(c)
PMR LEVEL (10) 20.405(a)(1)(ii) 50.36(c)(2) 50.73(a)(2)(vii)
OTHER 20.405(a)(1)(lit) 50.73(a)(2)(i) 50.73(a)(2)(viil}(A)
Specify in
#b a 20.405(a)(1)(iv)
X 50.73(a)(2)(ti) 50.73(a)(2)(vill)(B) n e 20.405(a)(1)(v) 50.73(a)(2)(lii) 50.73(a)(2)(x)LICENSEE CONTACT FOR THIS LER (12) hAME:
TELEPHONE NUMBER (include Area Code)
Sylvain L. Schwartz 609-971-4558 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)
T E

==CAUSE==
SYSTEM COMPONENT MANUFACTURER

==CAUSE==
SYSTEM COMPONENT MANUFACTURER 0
DS p
l t
l I
SUPPLEMENTAL REPORT EXPECTED (14)
MONTH DAV YEAR EXPECTED YES SUBMISSION (If yes, couplete EXPECTED SUBMISSION DATE).
X NO DATE (15)
ABSTRACT l
During the design process for a Core Spray piping modification, it was discovered that the existing Core Spray Systems 1 and 2 minimum recirculation piping configuration did not meet I the seismic and thermal expansion criteria allowables specified in the UFSAR. The root cause l
of this condition was inadequacy of the original design.
The safety significance is considered to be minimal, as the existing configuration meets the ASME Section III allowable seismic and thermal expansion criteria and satisfies the operability limits.
A modification presently scheduled for the upcoming 15R refueling outage will change the-subject piping configuration to meet the UFSAR criteria.
N3C FORM 366 (5 92)
N

d
!(5-92)
U.S. NUCLEAR RE(ULATORY (XBellSSION LICENSEE EVENT REPORT ( L"J Tt) y,7Q76 TEXT CONTINUATION FACILITY KAM (1)
DOCKET MLBIBFR (2)
LER IRDEBER (6?
PAGE (3)
SEQUENTIAL REVISION YEAR oyster creek, Unit 1 05000219 2 OF 3 94 1
-- 0 01 --
i 1
DATE OF DISCOVERY The condition described in this report was identified on January 13, 1994.
IDENTIFICATION OF DISCOVERY The Core Spray Systems 1 and 2 (E!IS-BM) minimum recirculation valves (CFI-FSV) V-20-92, V-20-93, V-20-94, and V-20-95 are not sufficiently supported for design seismic loading. The Core Spray System 1 minimum recirculation line is not flexible enough to accept design thermal
]
expansion, and two of the System 2 piping supports may exceed AISC requirements for supplemental structural steel. This condition is considered to be reportable in accordance with 10 CFR 50.73(a)(2)(ii).
CONDITIONS PRIOR TO DISCOVERY At the time of discovery, the plant was operating at approximately full power. The condition has been present in all plant. modes throughout the plant's operating history.
DESCRIPTION OF OCCURRENCE The existing Core Spray System 1 and 2 minimum recirculation piping configuration does not l
meet the seismic and thermal expansion criteria allowables specified in the Updated Final Safety Analysis Report (UFSAR) Table 3.9-1, which is based on ANSI B31.1 code. For system 1, the l seismic plus maximum operating loads (i.e., dead weight and pressure condition) exceed the code allowables by 2.33 times, using the design basis seismic response spectra. In addition, the thermal expansion load at the design temperature of 350"F, exceeds the code allowables by 4.87 times. For system 2, the seismic plus maximum operating load condition is 1.52 times the Code
~
c allowables. Also, the thermal loads at the design temperature of 350 F on two system 2 supports excced the supplemental structural steel AISC requirements; one by a factor of 1.2 (deadweight plus thermal expansion) and on the other by a factor of 1.32 deadweight plus thermal expansion plus seismic. This condition was discovered during the design process of a Core Spray piping modification.
NRCFORM36IA(S-92)
J

.(5-92)
U.S. NUCLEAR REQJLATORY CG#tlSSION
. LICENSEE EVENT REPORT (LER)
APPROVE 50-0104 TEXT CONTINUATION FACILITY KAME (1)
DOCKET NUMBER (2)
LER WUMBER (6?
PACE (3)
SEQUENTIAL REVISION YEAR oyster creek, Unit 1 05000219 3 OF 3 94 O
001 __
APPARENT CAUSE OF OCCURRENCE The cause of this condition was the inadequacy of the original piping design.
ANALYSIS OF OCCURRENCE AND SAFETY SIGNIFICANCE This condition is considered to have minimal safety significance for the following reasons:
: 1. An analysis using the most accurate seismic spectra currently available showed that the existing configuration is within the ASME Section III Level D allowable stress.
: 2. At the Core Spray System design temperature of 350 F, the System I con 6guration is acceptable for 30 cycles according to ASME Section III criteria. The two System 2 supports meet the ASME Section III, Appendix F, level D criteria. There are no records indicating that this pipe has seen the design temperature specified in the analysis. Any significant back leakage past the parallel injection valves (CIF-INV) would be detected by the high pressure alarm (CIF-PA) and relief valve (CIF-RV) lifting alerting plant operators. There is no history of leakage with either the check valves (CIF-ISV)inside containment or the parallel injection valves. Therefore, there is no indication that the piping or supports were ever overstressed.
: 3. A visual inspection of the most highly stressed weld performed on January 14, 1994, revealed no anomalies.
l 4
Based on the above safety significance discussion, the Core Spray Systems are operable by determining that they satisfy ASME Section III criteria. In the unlikely event of a seismic occurrence or thermal transient, these systems would perform their designed safety function.

==CORRECTIVE ACTIONS==
The Core Spray piping modification presently scheduled for the upcoming refueling outage, will change the existing configuration to meet the UFSAR criteria, ANSI B31.1.

==SIMILAR EVENTS==
LER 85-023, Emergency Service Water System Seismic Concerns LER 86-014, Containment Spray System Seismic Concerns LER 86-021, Plant Systems Did Not Meet Seismic Design Basis CRC FORM 366A (5-92)
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