ML20071N514

Interim Deficiency Repts 216 & 223 Re Gaseous Release from RHR Relief Valve Vents & Inadequate Standby Liquid Control Sys,Respectively.Final Deficiency Rept 232 Re Improper Installation of RHR Support Pin Receptacle Also Encl

ML20071N514
Person / Time
Site:	Columbia
Issue date:	05/17/1983
From:	Carlisce C, Carlisle C WASHINGTON PUBLIC POWER SUPPLY SYSTEM
To:	Martin J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V)
References
10CFR-050.55E, 10CFR-50.55E, 216, 223, 232, GO2-83-436, NUDOCS 8306070095
Download: ML20071N514 (6)

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P.O. Box 968 3000 GeorgeWashington Way Richland, Washington 99352 (509)3 -5 Docket No. 50-397 $ g May 17, 1983 h.,#

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/0 G02-83-436 Mr. J. B. Martin Regional Administrator U.S. Nuclear Regulatory Commission Region V 1450 Maria Lane, Suite 210 Walnut Creek, California 94596

Subject:

NUCLEAR PROJECT N0. 2 10CFR50.55(e) POTENTIALLY REPORTABLE CONDITIONS #216, RHR RELIEF VALVE VENTS; #223, STANDBY LIQUID CONTROL SYSTEM (SLCS); AND #232, LPCS AND RHR SUPPORT PIN RECEPTACLES

References:

a) Telecon dated October 22, 1982, R.T. Johnson to John Elin b) Telecon dated January 27, 1983, L.C. Floyd to R. Dodds (QA2-83-029) c) Telecon dated January 21, 1983, L.C. Floyd to John Elin (QA2-83-023)

In accordance with the provisions of 10CFR50.55(e), your office was informed by telephone, of the above subject conditions. Attachments A and B provide the Project's interim reports on Conditions #216, RHR Relief Valve Vents and #223, Standby Liquid Control System (SLCS).

Attachment C provides our final report for Condition #232, LPCS and RHR Support Pin Receptacles. Uc will continue to provide your office with quarterly updates on Conditions 216 and 223.

If you have any questions or desire further information regarding these subjects, please contact Roger Johnson, WNP-2 Project QA Manager, (509) 377-2501, extension 2712.

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. . Carlisle Program Director, WNP-2 Attachments: (3) As stated cc: W.S. Chin, BPA A. Forrest, Burns and Roe - HAPO N.D. Lewis, EFSEC WNP-2 Files /917B/917Y A. Toth, NRC Resident Inspector Document Control Desk, NRC B306070095 830517 gDRADOCK05000 pl0g7 l

ATTACHMENT A WASHINGTON PUBLIC POWER SUPPLY SYSTEM NUCLEAR PROJECT NO. 2 D0CKET NO. 50-397 LICENSE NO. CPPR-93 10CFR50.55(e) CONDITION #216 RHR RELIEF VALVE VENTS INTERIM REPORT Description of Deficiency There are 4 RHR relief valves which have a 2" vent nole on the valve body.

The valves are RHR-V-55A, RHR-V-95A, RHR-V-55B, and RHR-V-95B. These valves are situated such that the failure of a single motor operated valve, RCIC-V-ll3, would allow an open leakage path from the wetwell (primary containment) directly to the reactor building (secondary containment).

The attached sketch illustrates the situation. Containment penetration X-ll6 is an open path into the wetwell's gaseous volume. In the steam con-densing mode, RHR-V-55A (or B) and RHR-V-95A (or B) protect the RHR heat exchanger from over-pressurization. In order to accommodate condensation in the line between these valves and containment, a vacuum breaker has been installed that consists of penetration X-116, RCIC-V-ll3, RHR-V-102, RHR-V-101A (or B), RHR-V-103A (or B), and RHR-V-179A (or B).

All of these valves are normally open. Upon a containment isolation signal, the only valve to close would be RCIC-V-113. If it failed to close, and a LOCA had occurred, the wetwell would pressurize, and the wetwell atmos-phere would vent down this path. Details of the RHR relief valve show that a flow path exists which would allow the wetwell atmosphere to vent directly to secondary containment.

Safety Implication Burns an Roc has estiaated the gaseous release from these four paths at 5.2 x 10 scfm during the first 450 seconds after a LOCA and at a rate of 4.9 x 10 scfm thereafter. This can be compared to the allowable release rate for primary containment of approximately 1.7 scfm.

Corrective Action After a review of possible corrective actions, including incorporation of a bellows seal, the Project has decided to remove the relief valves, elimi-nate the containment leakage path, and deactivate the steam condensing mode of the RHR system. Supply System and Burns and Roe Engineering are prepar-ing the necessary Project Engineering Directives and FSAR changes to imple-ment the corrective action. We will continue to provide your office with quarterly updates on this subject. The next report will be submitted by August 17, 1983.

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ATTACHMENT B l

l UASHINGTON PUBLIC POWER SUPPLY SYSTEM NUCLEAR PROJECT NO. 2 DOCKET N0. 50-397 LICENSE NO. CPPR-93 10CFR50.55(e) CONDITION #223 STANDBY LIQUID CONTROL SYSTEM (SLCS)

INTERIM REPORT Description of Deficiency The Standby Liquid Control System (SLCS) is a backup system for safe shut-down which receives supply power from AC and DC safety-related buses.

General Electric electrical elementary diagrams have one loop of SLCS equip-ment powered and controlled from safety-related (Class lE) electrical buses, while the other loop of equipment is not. Power supplies which are not safety-related are connected to safety-related buses without isolators and control circuits within various panels are not separated: both of which are requirements of the WNP-2 Electrical Separation Criteria.

Safety Implication The system, as designed, does not meet single faiiura criteria per current FSAR statements, either mechanically or electrically. Electrical Separation Criteria has been violated concerning prime circuits which degrades the reliability of a safety-related electrical division. These conditions may preclude the SLCS from performing its intended function and, electrically, could cause failure of safety-related circuits through interfaces with non safety-related circuits.

Corrective Action General Electric has clarified the design requirements of the SLCS and sub-mitted recommended document chariges to preclude misinterpretation of the system's function and the licensing commitments. Corrective action is proceeding in accordance with this clarification as follows:

Completed:

o System electrical design has been issued to bring the system into com-pliance with electrical separation criteria.

In Progress; o Loop A electrical shall be fully qualified to meet safety-related Class lE requirements, including Quality Class I and Seismic Category I requirements.

o The existing design documentation shall be brought in compliance with the above.

Construction, equipment qualification and necessary document revisions are scheduled to be complete before fuel load.

We will continue to provide your office with quarterly updates. The next report will be submitted on or before August 17, 1983. I

ATTACHMENT C WASHINGTON PUBLIC POWER SUPPLY SYSTEM NUCLEAR PROJECT NO. 2 DOCKET NO. 50-397 LICENSE N0. CPPR-93 10CFR50.55(e) #232 LPCS-RHR SUPPORT PIN RECEPTACLE FINAL REPORT Description of Deficiency Bottom casing support pin receptacles were noc installed ia accordance with vendor's operation and maintenance manual requirement for safety-related pumps, RHR-P-2A, B and C and LPCS-P-1.

Safety Implication Safety-related pumps involved perform essential safe shutdown functions and must have high reliability and capability to function in conjunction with seismic event. The support pin receptacle was intended as a safeguard to assure these functions.

Engineering analysis has determined that due to the existing physical piping and pump configuration, the receptacle is not required to mitigate seismic reactions. Evaluations of operating vibration data by Engineering concluded that the equipment installation is satisfactory operationally without the pin receptacle.

The Engineering evaluation (identified in GEWP-2-83-83) has substantiated that the deficiency does not require extensive redesign or repair for the affected components to meet the critieria stated in the FSAR. The condition is therefore, considered not reportable.

Corrective Action No corrective action is required based on the Engineering evaluatica.