ML20197B741

Notice of Violation from Insp on 970729-0922.Violations Noted:Roving Fire Watches Were Not Implemented to Monitor Areas W/Inoperable Suppression Sys Intended to Protect Safe Shutdown Equipment

ML20197B741
Person / Time
Site:	Quad Cities
Issue date:	12/16/1997
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20197B025	List: ML20197B020 ML20197B241 ML20197B264 ML20197B741
References
50-254-97-14, 50-265-97-14, NUDOCS 9712240032
Download: ML20197B741 (5)

Text

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NOTICE OF VIOLATION Commonwealth Edison Company Docket Nos, 50-254;50-285 Quad Cities Station, Units 1 and 2 License Nos. DPR 29; DPR 30 During an NRC inspection conducted from July 29 through September 22,1997, four 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 listed below:

1. TS 4.0.A states," Surveillance Requiroments shall be met during the reacter operational mode (s) or other conditions specirled for Individual Limiting Conditions for Operations unless otherwise stated in an individual Surveillance Requirement."

s. TS Table 3.2.H 1 requires for en inoperabie main condenser offgas treatment explosive gas monitor system, grab samples be collected at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and analyzed within the following 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. If the racombiner temperature remains constant and thermal power has not changed, the grab sample collection frequency may be changed to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

b. TS 4.5.D.1 requires maximum scram insertion time for control rods be demonstrated for all control rods prior to thermal power exceeding 40 percent of rated thermal power,

c. TS 4.0.E requires inservice inspection of American Society of Mechanical Engineers (ASME) Code Class 1,2, and 3 components and inservice testing of ASME Code Class 1,2, and 3 pumps and valves be performed in accordance with Section XI of the ASME Doller and Pressure Vessel Code and applicable Addenda as required by 10 CFR Part 50, Sections 50.55a(g) and 50.55a(f).

ASME/ ANSI (American National Standards Institute) OMa Part 10,1988, paragraph 1.3.3.1(c)(2) states, ' Owners that satisfy testing requirements by installing a full complement of pretested valves in place of valves that had been inservice shall set pressure test the valves which were removed within 12 months of removal from the system."

d. TS 4.0.E requires inservice inspection of ASME Code Class 1,2, and 3 componcnts and inservice testing of AGME Code Class 1,2, and 3 pumps and valves be performed in accordance with Section XI of thu ASME Boller and Pressure Vessel Code and applicable Addenda as required by 10 CFR Part 50, Sections 50.55a(g) and 50.55a(f).

ASME/ ANSI OMa Part 10,1988, Sections 4.2.1.1 end 4.3.2.1, require active category B valves and check valves be tested nominally every 3 months.

e. TS 4.0.E requires insrirvice inspection of ASME Code Class 1,2, and 3 components and inservice testing of ASME Code Class 1,2, and 3 pumps and valves be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR Part 50, Sections 50.55a(g) and 50.55a(f).

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9712240032 971216 PDR ADOCR 05000254 G PDR

Notios'of Violation- -  ?

Section XI of the A8ME Boiler and Pressure Vessel Code 1989 (no addenda),

Section IWA 5214, ? Repairs and Replacement," states, "A component replacement shall be pressure tested prior to resumption of service."

- Contrary to the above:

a. - On September 4,1997, at 10:15 p.m., and on September 5,1997, at 8:00 p.m.

and at 11:40 p.m., the licensee changed the recombiner temperatures without increasing the Unit 2 main condenser offgas treatment explosive gas monitoring l

system grab sample frequency to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

On August 19,1997, with the Unit 2 "B" main condenser offgas treatment ,

, explosive gas monitor system (offgas hydrogen analyzer) inoperable, chemistry technicians failed to take an 8-hour grab sample fmm the Unit 2 main condenser offgas treatment system with the system in operation.

b. On June 23,1997, operators raised Unit 2 thermal power above 40 percent of '

rated thermal power without completing scram insertion time testing for five control rods.

c. As of June 1,1996, Target Rock Safety Relief Valve 2 203-3A, which was replaced on June 1,1995, had not been set pressure tested one year after removal from the system. As of March 15,1997, Target Rock Safety Relief Valve

, 1203-3A, which was replaced on March 15,1990, had not been set pressure tested one year after removal from the system.

. d. On May 9,1997, the licensee discovered that Manual Valves W3999-89

_ (an active Category B valve) and S3999-88 (chtli valve) from the % diesel generator cooling water pump to the emergency core cooling system room coolers were not tested during the second quarter of 1997.

e. - On April 29,1997, the licansee discovered that a pressurs test of the replaced high pressure coolant injection (HPCI) Check Valve 1-2301-45 was not performed prior to placing the system in service.

This is a Severity Level IV violation (Supplement 1). (50-254/97014-01; 50-265/97014-01)

2. Title 10 CFR 50,' Appendix B, Criterion XI, " Test Control," requires, in part, that testing required to demonstrate that systems and components will perform satisfactorily in F service be performed in accordance with written test procedores which incorporate r ~

acceptance limits contained in applicable design documents. Test procedures shall

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include provisions for assuring that adequate test instrumentation is available and used.

, a.' TS 4.9.5 requires the capacity of the Unit 2 safety-related 250 Vdc battery be

- verified at least the greater of either 80 percent of the manufacturer's rating or the .

minimum acceptab;e battery capacity from the load profile when subject to a l modified performance test. ,

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-L Notice of Violation 3'

b. . TS 4.8.J.2 states, "The safe shutfown makeup pump (SSMP) system shall be -

demonstrated operable: At least Gnos per 92 days by verifying that the SSMP develops a flow of greater than or equal to 400 gpm against a system head corresponding to reactor vessel pressure of greater than 1120 psig."-

Contrary to the above:

si On April 7,1997, TS 4.9.5 surveillance requirements were not incorporated into -

Quad Cities Technical Staff Procedure (QCTS) 0240-06, Revision 2, " Unit One (Two) Modified Performance Test 250 Vdc Safety Re;ated Battery." On April 7,

- 1997, the modified performance test on the Unit 2 safety-related 250 Vdc battery -

was performed with a battery capacity acceptance criteria of 70 percent instead of 80 percent as required by TS 4.9.5.

b. From September 23,1996, until approximately July 29,1997, Quad Cities -

Operating Surveillance (QCOS) 2900-01, Revision 12, "Quarterty Safe Shutdown Makeup Pump Flow Rate Test," did not adequately incorporate design requirements and provide assurance that adequate instrumentation would be .'

used to demonstrate that the safe shutdown makeup pump system would -

perform catisfactority in service. Test acceptance criteria did not incorporate instrument tolerances to ensure that the flow and system head design requirements would be met.

This is a Severity Lcvel IV violation (Supplement I), (50-254/97014-04; 50-265/97014-04; 50 265/97014-07; 50-265/97014-07)

3. Title 10 CFR 50, Appendix B, Criterion lil, " Design Control," requires, in part, that measures shall be established to assure that the design basis, as defined in 10 CFR 50.2 and as specified in the license application, are correctly translated into specifications and procedures and instructions.

Contrary to the above:

, s. On February 13,1397, design basis information from the license application was not correctly translated into the battery sizing calculations, dated February 13, 1997, included in Nuclear Design Information Transmittal SO40-QH-0296. The --

calculation used a lowest expected electrolyte temperature of 65'F as an input for

- the sizing of the 250 volt safety-related battery. However, tha surveillance requirements of TS 4.9.C.2.c required that the average electrolyte temperature of all connected cells be above 60'F.

b. From 1993 to the present, the design basis of the 250 volt battery was not correctly translated into battery load profile calculation PMED 891377-01. The <

calculation assumed that failure of the unit ememency diesel generator (EDG) would produce the worst case load profile. However, a 1993 modification to the ,

plant, removing the direct current (dc) turbir e emergency oil pump as a load on

the 250 voit battery, resulted in the worst case load profile coming from failure of -

the M (swing) EDG.

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Notice of Viole' - )

1 This is a Severity Level IV violation (Supplement I). - (50 254/97014-05a; -

50 26&97014 05a; 50 254/97014-05b; 50-265/97014-05b).

4. Th 6.8.A requires applicaole procedures recommended in Appendix A of Regulatory -

Guide 1.33, Revision 2, February 1978, be implemented. Paragraph 9.a. of Appendix A l recommends procedures for performing maintenance that can affect the performance of-safety related equipment. Paragraph 1.1. of Appendix A recommends plant fire protection program administrative procedures.

- a. Quad Cities Mechanical Maintenance (QCMM) Procedure 1515-07.- Revision 7,

" General Valve Packing Procedure," Stop I.2, requires the removal of packing, ,

and Stop 1.5.d. requires use of QCMM 1515 07, Attachment D, for replacing packing on valves with an active leak-off line,

b. . Quad Cities Administrative Procede e 1500-01, " Administrative Requirements of Fire Protection," Revision 6, dated February 17,1997, Step D.2.c.2.(b), requires an hourty roving fire watch be established if a water suppression system which protects a safe shutdown system is inoperable nd the affected unit is not in a

- safe shutdown condition.

Contrary to the above;

a. On March 11,1997, Steps 1.2 and 1.5.d. were not performed during use of s QCMM 1515-07 for valve repair associated with the 2A low pressure coolant -

injection (LPCI) loop air operated Check Valve 2-1001-68A, a valve with an active leak-off line.

b. On September 10,1997, roving fire watches were not implemented to monitor areas with inoperable suppression systems intended to protect safe shutdown toulpment.

This is a Severity Level IV violation (Supplement I). (50-254/97014-02; 50-265/97014-02; 50-254/97014-09; 50-265/97014-09)

Pursuant to the provisions of 10 CFR 2.201, Comrnonwealth Edison .s hereby required to submit a written statement of explanation to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington D.C. 20555, with a copy to the Regional Administrator, Region 111, and a copy to the NRC Resident inspector at the facility that is the F subject of this Notice, within 30 days of the date of tbs letter transmitting this Notice. This reply should be clearty marked as a " Reply to a Notice of Violation" and should include for each -

violation

-(1) the reason for the violation or, if contested, the basis for disputing the violation, t _ (2) the corrective steps that have been taken and the results achieved, (3) the corrective steps

that will be taken to svold further violations, and (4) the date when full compliance will be achieved. Your response may reference or include previous docketed correspondence, if the correspondence adequately addresses the required response if an adequate reply is not received within the tirr 3 specified in this Notice, an order or a Demand for Information may be issued as to why th6 license should not be modified, suspended, or revoked, or why such other c

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Notice of Violation action as may be proper should not be taken. Where good cause is snown, cor. sideration will be given to extending the response time.

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 without redaction. However, if you find it necessary to include such information, you should clearly indicate the specific information that you desire not to be placed in the PDR, and provide the legal basis to support your request for withho; ding the information from the public. .

Dated at Lisle, Illinois, this 16th day of December 1997 2

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