ML20198L459

Notice of Violation from Insp on 970721-0827.Violation Noted:From Initial Licensing Until 970615,suction Piping for HPCS & RCIC from CST

ML20198L459
Person / Time
Site:	Perry
Issue date:	09/23/1997
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20198L449	List: ML20198L444 ML20198L459 ML20198L512 ML20198L524
References
50-440-97-08, 50-440-97-8, EA-97-430, NUDOCS 9710270046
Download: ML20198L459 (8)

Text

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i NOTICE OF VIOLATION Centerior Service Company Docket No. 50-440 Perry Nuclear Power Plant License No. NPF-58 EA 97-430 During an NRC inspection conducted on July 21, through August 27,1997, violations of NRC requirernents were identified, in accordance with the " General Statement of Policy and Procodure for NRC Enforcement Actions," NUREG-1600, the violations are listed below:

A. Design Control Violatl .ns:

1, 10 CFR Part 50, Appendix B, Criterion 111 states in part that measures shall be established to assure that the design basis for structures, systems and components are correctly translated into speedications, drawings, procedures, and instructions.

Section 3.5.1.4 ' Missiles Generated by Natural Phenomena,' of the Updated Safety Analysis Report (USAR) atates in part that: those systems or components listed in Table 3.2-1 that are required to ensure the integrity of the reactor coolant pressure houndary or maintain safe shutdown conditions are provided with tomado missile protection by location within seismic Category I structures, unique missile barriers or by the shielding of an adjacent seismic Category I structure.

Contrary to the above requirements, from initial licensing until June 15,1997, the suction piping for high pressure core spray (HPCS) and reactor core isolation cooling (RCIC) from the condensate storage tank (CST), and the emergency service water discharge piping were listed in Table 3.2-1 of the USAR; however, they were not protected from tornado generated missiles as described in the USAR. These components respectively, are required to ensure the integrity of the reactor coolant pressure boundary and to maintain safe shutdown conditions.

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

2. 10 CFR Part 50, Appendix B, Criterion Ill states in part that the design control measures shall provide for verifying or checking the adequacy of the design.

The Perry Operations Quality Assurance Program, USAR 17.2 commits to compliance with Regulatory Guides and Standards as listed in USAR Table 1.8-2.

USAR Table 1.8-2 commits to following ANSI N45.2.11 - 1974 for Quality Assurance Requirements For The Design of Nuclear Power Plants.

Nuclear Engineering Instruction NEl-0341 Revision 5 " Calculations

• applies to all calculations to establish design bases or to change design documents.

Paragraph 6.2, Calculation Revisions states " Design Engineers are to monitor calculations to 9710270046 PDR 970923 0 ADOCK 05000400 PDR

i Notice of Violation 2 <

4 determine if a revision is required e.g. receipt of new/ revised design input, f confirmation of assumption etc." Paragraph 6.3, Review and Approval, states,

" Verification / review and approval of calculation _should precede use of the results for design, but must be completed prior to the component, system, or structure being declared operable."

ANSI N45.2.11 states that design analyses shall be performed in a planned, controlled and correct manner, in addition, the design activities shall be prescribed and accomplished in accordance with procedures which provides adequate checking or verifying the results of the i ctivity.

(a) Contrary to the above requirements, as of March 27,1997, the. ,

licensee had modified various systems as reflected in design drawings and did not update / revise the calculations as described in the following examples:

(1) Electrical drawing D 206-029, " Electrical One- Line Diagram, 4 Class IE,480 V Bus EF10," Revision BB, identified the installation of a 10-hp electric motor for compressor 1E22-C004A, Calculation PRMV-0017 "EHF-1 E Transformer .

Breaker EH1305," Revision 0, did not list the compressor motor. USAR Table 8.31 also did not correctly identify the moior loads. Calculation PRMV-0017 was last updated on March 11,1985 (12 years ago), and did not reflect the current plant loads and settings.

(2) Calculation PSTG 0003 "480 V Safety-Related Motor Starting Voltage Drop," Revision 2, dated June 29,1995 (page 6),

contained an open assumption that required confirmation.

Calculation PSTG-0001, "PNPP Auxiliary System Voltage Study," Revision 2, approved on August 24,1995, provided the information to resolve the open assumption. As of March 27, 1997, calculation PSTG-0003 was not updated to close the -

open assumpiion, even though the information to close the assumption was available on August 24,1995.

(3) Calculation PRDC-0006, " Load Evaluation and Battery Sizing of Division 111 Class IE DC System," Revision 0, dated April 8, 1991, did not address Division ill high pressure core spray

-(HPCS) pump 1E22C001 breaker EH1304 spring charging motor load at t=0 second, the load profile for 0-1 min for continuous (L2) load, and the DC control circuit loads (L2 -

loads) of the breakers.

Notice of Violation 3 (4) Calculation PRDC-0004, " Class IE DC Control Circuit Coordination," Revision 2, dated May 30,1995, did not address switch #12 added to drawing D206-051, ' Electrical Main One-Line Diagram, Class IE DC System" Revision RR, dated May 15,1992, in accordance with DCP 90 0012.

(5) Calculation PRLV-0004, '480 V Breaker Coordination,"

Revision 2, dated April 30,1996, was reviewed against associated electrical D 206 series drawings for 480-V motor control centers (MCCs). Various discrepancies and typographical errors were found between the calculations and the drawings as noted below:

Dem# Calculation PRLV-0004 Drewino D-206 series 1B21-F065A 6.6 HP 6.4 hP P42-F551 MISSING 0.13 HP P45-D004A 7 HP 1 HP P42F550 MISSING 0.13 HP M25-C001B 100 HP 60 HP 1G33 F001 3.0 HP 3.9 HP This is a Severity Level IV violation (Supplement I).

(b) Contrary to the above, as of March 27,1997, the Perry Plant design control measures did not ensure that calculations or analyses were verified and controlled adequately for the following examples:

(1) Calculation P11-12, "P11 - Level Setpoints in CST for E22 and E51 Instruments' dated March 12,1985, determined the CST low-level swapover setpoint limit required to ensure that the HPCS system has adequate net positive suction head and that no vortex occurs before suction valve swapover to the suppression pool. The calculation was inadequate, in that, it used a non-conservative (improper) HPCS flow rate and did not consider the impact of valve timing when establishing the CST low-level swapover setpoint.

(2) Engineering input for safety evaluation 96-128 dated October 10,1996, associated with USAR change request (CR)96-150 for Section 9.2.2 revision, which changed the emergency closed cooling (ECC) system surge tank sizing basis from a 7-day supply without necessary makeup to a 30-minute supply, used a non-conservative flooding rate of 60 gpm for surge tank overflow on the basis of the minimum calculated emergency service water (ESW) makeup flow to one ECC surge tank.

Since the operating procedure directs the operator to initiate ESW makeup to both tanks, the flooding rate should have consider the flooding of both l

,=.

Notice of Violation 4 tarAs (i.e.,120 gpm). Additionally, the flooding potential should have been calculated using assumptions that ESW flows would be the maximum (117 gpm to each tank) rather than minimum (60 gpm to each tank).

(3) Calculation P42 31,'ECC A Heat Exchanger Test Results-1995," Revision 0, dated September 15,1995, contained an assumption . hat the test instrumentation was within calibration limits. Post-test calibration was specified to confirm this assumption. As of March 27,1997, the outstanding assumptions, open for 18 months, which could affect calculation acceptability, had not been confirmed or closed.

(4) Calculation E22 2, ' Overpressure Protection Analysis," Revision 0, dated February 23,1983, which performed an overpressure protection analysis on a portion of the HPCS system was inadequate. The analysis did not identify operating conditions under which pressure relief devices were required to function.

The maximum pressure considered for the suction piping was 31.25 psig whereas the suction side relief was set at 100 psig.

Maximum discharge pressure considered was 1130 psig whereas the discharge side thermal relief valve was set at 1560 psig. Further, the analysis for the suction piping did not evaluate other pressurization potentials such as post accident alignment from the suppression pool with consideration of containment overpressure, or conditions of back-leakage from the reactor pressure vessel (RPV).

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

B, Corrective Action Violation 10 CFR Part 50, Appendix B, Criterion XVI requires conditions adverse to quality such as malfunctions and deviations be promptly identified and corrected.

The Perry Operations Quality Assurance Program, USAR Chapter 17.2, requires in Section 17.2.16 that significant conditions adverse to quality be identified and have action taken to prevent recurrence.

Contrary to the above, as of March 27,1997, the following examples of untimely or ineffective conective action were identified:

1. The testable rupture discs for emergency diesel generators had failed to properly operate on more than 12 occasions over the past 12 years. The root cause evaluation for the most recent failure (February 19,1997) documented in PlF 97-0325 attributed one of the causes to untimely corrective action.

Notice of Violation 5

2. _ Test results for the HPCS keep full pump (from TXI-229, dated March 19,1996) showed that the pump was not capable of delivering the flows and pressure specified in USAR Section 6.3. This degraded condition had existed since July 24,- 1993, when the surveillance test was conducted. As of March 27,1997, this issue had not been resolved.

3. The licensee's safety system functional inspection of the HPCS system in 1992, recognized that the basis for not installing the HPCS pump motor overfrequency protection relay during construction, as documented in FDDR KLI 3890, dated May 28,1985, was not well founded. Consequently, the licensee performed Calculation E2219," Justification for Elimination of HPCS Overfrequency Relay,"

Revision 1, dated July 23,1992, to evaluate the effect of not installing the relay.

The licensee's corrective action decision in 1992 was improper for the following reasons:  !

• The calculation referenced Section lil, NB36541, of the ASME BSPV Code in order to justify exceeding the system design pressure in the event that the Division 111 emergency diesel generator frequency goes above 60 Hz.

The licensee did not identify the specific edition or addenda of the Code; however, Design Specification (DSP) E221-4549-00, Revision 3, dated April 18,1986, specifies the 1974 ASME B&PV Code with addenda up to and including the winter 1975 issue, Section Ill, Division 1. This code edition and addenda did not provide adequate basis to enable the licensee to justify the allowances for exceeding the design pressure in accordance with NB 3654-1.

• The licensee's calculation methodology provided a relief path to limit pressure using the minimum flow valve and its actuation circuitry as overpressure protection devices. However, the licensee was unable to demonstrate compliance with the requirements for this valve and its actuation circuitry, as specified in ASME Code Section 111, Article NC-7000,

' Protection Against Overpressure."

4. On October 8,1996, emergency closed cooling boundary valves were included into the licensee's ln service Test program as part o' corrective actions. This corrective action was ineffective in that test procedure PTI-P42-P0008, 'P42 (ECC) System Leak Rate Test Procedure " revision 1, was inadequate and did not demonstrate that equipment could perform satisfactorily under accident conditions. Specifically, as of March 27,1997, the differential test pressure specified for system boundary valve seat leak testing was only approximately %

of the pressure that the valves would be subjected to under accident conditions, No extrapolation of test data to compensate for this difference in test conditions was included in the procedure.

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

i

Notice of Violation 6 C. Changes to the Facility Violation 10 CFR 50.59, " Changes, tests and experiments," permits the licensee, in part, to make changes to the facility as described in the safety analysis report without prior Commission approval provided the change does not involve an unreviewed safety question (UEQ). A proposed change, test, or experiment shall be deemed to involve a USQ if, in part, a possibility for an iccident or malfunction of a different type than any evaluated previously in the safety analysis report may be created. The licensee shall maintain records of changes in the facility and these records must include a written safety evaluation which provides the bases for the determination that the change does not involve a USQ.

USAR Sections 6.3.2.2.1 and 6.2,4.2.2.2 describe the normal alignment of the HPCS system and the position of containment isolation valves for the suppression pool cleanup system (SPCU). Specifically, Section 6.3.2.2.1 Indicates that HPCS suction is normally aligned to the CST and that it will automatically switchover to the suppression pool. Section 6.2.4.2.2.2 for the SPCU system containment isolation valves states that the compliance for containment isolation requirements for the retum line is satisfied, in part, on the basis that the line is normally closed.

Contrary to the above requirements, from the Spring of 1993 until March 31,1997, the SPCU system was in operation almost continuously resulting in a HPCS alignment different than described in the USAR and resulting in the SPCU system con'.ainment isolation valves to be normally open versus closed, as specified in the llMR. This continuous operation of the suppression pool system was not supported by a written safety evaluation as required by 10 CFR 50.59.

This is a Severity Level IV violation (Supplement I). (50-440/97-201-09)

D. USAR Update Violat or.

10 CFR 50.71(e) requires the licensee to update the Final Safety Analysis Report (FSAR) originally submitted as part of the application for the operating license to assure that the information included in the FSAR contains the latest material developed. The updated FSAR shall be revised to include the effacts of, in part, all safety evaluations performed by the licensee in support of conclusions that changes did not involve a unreviewed safety question. Updates must be filed annually or 6 months after each refueling outage. The updates must reflect all changes up to a maximum of 6 months prior to the date of filing.

Contrary to the above, as of March 27,1997, the licensee failed to update the FSAR (currently referred to as the Updated Safety Analysis Report (USAR)) to reflect plant conditions, which existed more than 6 months prior to the previous USAR update, as evidenced by the following examples of inaccurate or non updated USAR information:

Notice of Violation 7

1. USAR Table 8.31 listed fuel oil transfer pumps 1R45C001C and 2C as 0-second loads.1R45C001C and 2C were 40-minute automatic cyclic loads for both loss of offsite power (LOOP) and loss of coolant accident (LOCA),

2. USAR Table 8.3-1 identifieed a 9-kW load for 1E22C004B, which did not agree with the 8 kW load in Calculation PSTG-0014.

3. USAR Table 8.3-1 listed the inrush currents for HPCS fuel oil transfer pumps 1R45-C001C and 2C as 109A, whereas Calculation PRMV-0017 listed the inrush current as 130A.

4. USAR Table 8.31 listed the inrush currents for HPCS diesel generator room fans OM43-C001C and 2C as 362A, whereas Calculation PRMV-0017 listed the inrush current as 376A.

5. USAR Tabie 8.31 listed the Full Load Amperage (FLA) of HPCS diesel generator starting air compressor 1E22 C004B as 13A, whereas Calculation PRMV-0017 listed the FLA as 11 A.

6. USAR Table 8.31 listed the HPCS ESW pump 1P45-C002 is as 75 hp,88.5 FLA, and 557A inrush, whereas Calculation PMRV-0017 listed the same load as 75 hp,85.4 FLA, and 543A inrush.

7. USAR Table 8.3-1 listed the rating of HPCS diesel generator space heater 1E22-D011 as 2 kW, with a load current of 3 amp. Calculation PRMV-0017 listed the same space heater as 1.6 kW, with the load current of 2.01 amp.

Drawing D-206-029/BB, ' Electrical One Line Diagram, Class IE,480-V Bus EF10," listed the same space heater as 2.4 kW.

8. USAR Table 3.9-30 listed active valves not associated with the nuclear steam supply system (NSSS). This table had not been updated to reflect several ECC system modifications. Valves P42 F315A,B.C should have been deleted from the table, since they were converted from automatic to manual valves by DCP 92-0060. Valves P42-F550 and P42 F551 should have been added to the table, since they were converted from manual to automatic valves by DCP 90-0012,

9. USAR Tables 9.2-18 (ECC Pumps) and 9.2-19 (ECC Heat Exchangers) listed two different values for ECC system operating flow rate (1860 versus 1920 gpm). Since all pump flow was delivered to the heat exchanger, the flow values should have been the same.

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

d w

Notice of Violation 8 Pursuant to the provisions of 10 CFR 2.201, Centerior Service Company 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, Region ill, and a copy to the NRC 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 violation: (1) the reason for the violation, or, if contested, the basis for disputing the violation, (2) the corrective steps that have been taken and the results achieved (3) the corrective steps that will be taken to avoid 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 time specified in this Notice, an order or a Demand for Informa' ion may be issued as to why the license should not be modified, suspended, or revoked, or why such other action as may be proper should not be taken. Where good cause is shown, consideration will be given to extending the response time, Under the authority of St.ction 182 of the Act,42 U.S.C. 2232, this response shall be submitted under oath or affirmation.

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. If personal privacy or proprietary information is necessary to provide an acceptable response, then please provide a bracketed copy of your response that identifies the information that should be protected and a .edacted copy of your response that deletes such information. If you request withholding of such material, you must specifically identify the portions of your response that you seek to have withheld and provide in Jetail the bases for your claim of withholding (e.g.,

explain why the disclosure of information will create an unwarranted invasion of personal privacy or provide the information required by 10 CFR 2.790(b) to support a request for withholding confidentia: commercial or financial information). If safeguards information is necessary to provide an acceptable response, please provide the level of protection described in 10 CFR 73.21.

Dated at Lisle, Illinois, this day of September 1997