ML20199G788

Notice of Violation from Insp on 970707-970824.Violation Noted:In Oct 1993,vendor Specification for Emergency Svc Water Pump Bowl Assemblies Was Changed from Cast Iron to Stainless Steel

ML20199G788
Person / Time
Site:	Oyster Creek
Issue date:	11/17/1997
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20199G780	List: IR 05000219/1997006 ML20199G788
References
50-219-97-06-01, 50-219-97-6-1, EA-97-241, NUDOCS 9711250290
Download: ML20199G788 (5)

Text

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ENCLOSURE .,

i NOTICE OF VIOLATION

' GPU. Nuclear incorporated _

Docket No. 50-219

-Oyster Creek Nuclear Generating Station License No. DPR 16..

EA 97 421 During an NRC inspection conducted from July 7,1997 to August 24,1997, for which an exit -

meeting was held on September 4,1997, violations of NRC requirements were identified. In accordance with the " General Statement of Policy and Procedure for NRC Enforcement Actions," NUREG-1600, the violations are cet forth below:

l. VIOLATlONS RELATED TO DESIGN CONTROL T

The, Code of Federal Regulations,10 CFR Pan 50, Appendix B, Criterion Ill,' requires, in part, that measures shall be established to assure that applicable regulatory-

- requirements and the design basis, as defined in i 50.2 and as specified in the license application, for those structures, systems, and components to which this appendix applies, are correctly translated into specl%3ons, drawings, procedures, and ,

instructions. Design control ceasures shall b, u > lied to compatibility of materials.

A. Contrary to the above, in October 1993,-the vendor specification for the -

emergency service water (ESW) pump bowl assemblies was changed from cast iron to stainless steel without adequately applying design control measures for use in considering the compatibility of materials specified for other ESW pump components. As a result, in November 1993, the "C" and "D" ESW pumps were installed with cast iron top case flanges and stainless steel bowl assemblies, and the materials in these components were incompatible in that they created the opportunity for accelerated galvanic corrosion. Subsequently, the "C" ESW pump was found to be inoperable on July 31,1997, because of a broken coupling (top case flange) between the pump bowl and the discharge head. (01013)

This violation is classified at Severity Level 111 (Supplement 1).

B. Contrary to the above, in April 1989, the licensee failed to establish adequate measures to assure that the design basis for reactor water cleanup (RWCU)

- system inlet isolation valve V-16 2 was correctly translated into specifications, drawings, procedures, and instructions. Specifically, in April 1989, the licensee modified the RWCU system procedures such that operators periodically opened V-16 2 at full reactor; pressure (1020 psig) in order to fill and pressurize the RWCU system. However, the design input of 125 psid differential pressure used in the design basis calculations that determined the capabilities of V-16-2 failed to consider that V-16 2 might be required to close against full reactor pressure. Therefore, there was no assurance that V-16-2 would be able to Wose to perform its required isolation function against full reactor pressure.

(02013)

This violation is classified at Severity Level lli (Supplement 1).

l i

lI 9711250290 971117 o PDR ADOCK'05000219 L e PDR l

, i Enclosure 2 C. Contrary to the above, in 1994, the licensee f ailed to take adequate measures to assure that specifications for the 4160 volt vital bus degraded voltage relay (DVR) setpoints were adequate to ensure that the onsite distribution system, in conjunction with the offsite distribution system, was capable of providing .

acceptable voltage under worst case station electric load and grid voltages as specified in the Final Safety Analysis Report (FSAR), Sections 8.3.1.2.a and 8.2.2.g. Specifically, the degraded grid analysis, performed in 1994, failed to I consider that the startup transformer voltage regulators could lower bus )

voltage. The analysis did not consider the design of the voltage regulators, l specifically the regulator response time and setpoints, in selection of the DVR l setpoints. As a result, on August 1,1997, when the output of the Oyster -l Creek unit was lost due to a manual reactor scram, the startup transformer voltage regulators failed to supply sufficient voltage to the vital busses to preclude an undervoltage (UV) condition. This UV condition lasted for greater than 10 seconds which caused the DVRs to trip (and not reset) which de-energized the vital buses and started the emergency diesel generators contrary I to the design bases. (03013)

This violation is classified at Sever;ty Level 111 (Supplement 1).

l

11. VIOLATION RELATED TO INOPELABLE CRD PUMPS Oyster Creek Technical Specification 3.4.D.1, requires that the control rod drive (CRD) hydraulic system shall be operable when the reactor water temperature is above 212 l degrees F.

Contrary to the above, for a substantial period of time prior to August 1,1997, with reactor water temperature above 212 degrees F, both CRD pumps were inoperable in that both CRD pumps tripped and could not be restarted remotely when they attempted to load onto the vital bus following a manual reactor scram and loss of offsite power to the vital busses. The anti-pumping feature of the CRD pump breakers prevented the breakers from closing after a trip signal with a start signal continuously supolied. The

. CRD pumps are designed to automatically load onto the vital busses following a loss of offsite power. (04014)

This violetien is classified Severity Level IV (Supplement 1).

Ill. VIOLATIONS RELATED TO INADEQUATE CORRECTIVE ACTIONS The Code of Federal Regulations,10 CFR Part 50, Appendix B, Criterion XVI (Corrective Action), requires that measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected. in the case of significant conditions adverse to quality, the measures shall assure that the cause of the condition is determined and corrective action taken to preclude repetition.

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, Enclosure . ~3

. A. Contrary to the above, between January 22;1997, and July 9; 1997, the licensee- failed to take timely and' effective corrective action to procludel repetition of a significant condition adverse to quality that resulted in both trains of the Standby Gas Treatment (SBGT) system being inoperable.- Specifically, the licensee failed to ensure that technicians assigned to' calibrate the reactor

. _ building; (RB) ventilation exh1ust- rediation monitors _ on -July 9, _1997, had - .

. received remediallogarithmic seine training.- Training should have been provided

- as corrective action for a previous event on January 22,1997, which resulted in miscalibration of the RB ventilation exhaust radiation monitors. However, all of the technicians that calibrated the radiation monitors on July 9,1997, had =

not received the training and the technicians' supervisor failed to ensure that

the technicians were properly trained. As a result, the RB ventilation exhaust '

< radiation monitor setpolnts were set nonconservatively high which would have delayed initiation of the SBGT system. (05014) i-This violation is classified Severity Level IV (Supplement- 1).

B. Contrary to the above, between September 13,1996, and August 3,1997, the licensee failed to take effective corrective action to preclude repetition of a significant condition adverse to quality that removed the low suction pressure F -protection feature from the shutdown cooling (SDC) pumps, snd resulted in inadvertent SDC pump trips. The SDC pump trips could have resulted in a loss of SDC. Specifically, the licensee failed to assure that corrective actions to revise procedures and provide training were adequate to ensure that the suction pressure switches were not isolated during SDC pump operation following

- similar events on September 13,1996, and April 24,1997. On April 24,1997,

f. and August 3,1997, a shutdown cooling pump inadverte- Jy tripped because F the suction pressure switches were isolated. (06014)

This violation is classified Severity Level IV (Supplement 1).

[ . Pursuant to tne provisions of 10 CFR 2.201, GPU Nuclear incorporated (Licensee) is hereby required to submit a written statement or explanation to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, D.C. 20555, with a copy to the Regional Administrator, U.S. Nuclear Regulatory Commission, Region I, and a copy to the NRC

. Senior Resident inspector at the_ facility that is the subject of this Notice, within 30 days of the date of the letter transmitting this Notice of Violation (Notice). This reply should be clearly marked as a " Reply to a Notice of Violation" and should include for each alleged violation:

(1) admission or denial of the alleged violation, (2) the reasons for the violation if admitted, and if denied, the reasons why, (3) the corrective steps that have been taken and the results achieved,-(4) the corrective steps _that will be taken to avoid further violations, and (5) the

-date' when full compliance will be achieved. Your response- may reference or include e previously docketed correspondence, if the correspondence adequately addresses the required L response, if an adequate reply.is nor received within the time specified in this Notice, an Order or a Demand for Information may be issued as to why_ the Ibense should. not be modified, suspended, or revoked or why such other action as_may be proper should not be taken. - Consideration may be given to extending the response time for good cause shown.

Under;the authority of Section 182 of the Act,. 42 U.S.C. 2232, this response shall be i submitted under oath or affirmation.

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1 Enclosure 4- 1 l

Because your response will be placed in the NRC Public Document Room (PDR), to the extent possible, it should not include any personal privacy, proprietary, or safeguards information so that it can be placed in the PDR, if redactions are required, a proprietary version containing brackets placed around the proprietary, privacy, and/or safeguards information should be e submitted. In addition, a non-proprietary version with the information in the brackets redacted should be submitted to be placed in the PDR.

Dated at King of Prussia, Pennsylvania this17thfay of November 1997 4