IR 05000369/1990012
| ML20044B142 | |
| Person / Time | |
|---|---|
| Site: | McGuire, Mcguire  |
| Issue date: | 07/03/1990 |
| From: | Belisle G, Lenahan J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20044B140 | List: |
| References | |
| 50-369-90-12, 50-370-90-12, NUDOCS 9007170425 | |
| Download: ML20044B142 (7) | |
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UNITED STATES
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NUCLEAR REGULATORY COMMISslON '
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101 MARIETT A STREET.N.W.
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ATLANTA. GEORGI A 30323
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' Report Nos.:: 50-369/90-12 and 50-370/90-12
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Licensee: Duke-Power Company-422. South Church Street Jo % s
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- Charlotte.-NC~ 28242'
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Docket Nos.:
50-369 and 50-370 License Nos.: NPF-9 and NPF-17
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Facility Name: ~McGuire l'and 2
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l Inspection Conducted:.-June 4-7 1990
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. In'spector: -
J.-J.Lenahang('
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" Approve by:.
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G. A._Belfsle, Chiefv Date Signed,
. Test Programs Section
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J Engineering Branch
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Division of Reactor Safety
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SUMMARY
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, Scope:.
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- This ' routine,--unannounced inspection was. conducted in the areas of licensee;
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ii: antified item'(LERs) and licensee action on' previous inspection findings.
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' :Results:
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In. the; areas inspected, violations or deviations were not identified.
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g The licensee's: approach to. resolution of technical issues was conservative and.
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timely.
Responsiveness to'NRC initiatives was generally sound, thorough, and
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REPORT DETAILS
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Persons Contacted.
-Licensee Employees-
- N Atherton, Compliance
- J. Boyle, Superintendent of Integrated Scheduling W.-Harris, Maintenance Engineer K. Kelly, Maintenance Engineer
- T.,Mathews, Manager, Design Engineering
- M. Nazar, Performance Engineering Supervisor J. Oswald, Nuclear Production Engineer Performance Other licensee employees contacted during this inspection included engineers and technicians q
NRC Resident Inspectors
- T. Cooper, Resident Inspector S. Ninh, Resident Inspector K. Van Doorn, Senior _ Resident Inspector
- Attended exit interview l2.
Onsite' Followup of Written Reports of Non Routine Events at Power Reactor'
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Facilities (92700)
(C1.osed) Licensee Event Report (369/85-29):- Analysis Error in Containment
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Pressure Non-Conservative On October 1,1985, Westinghouse notified the licensee that a reanalysis
of the peak containment pressure calculations for Final Safety Analysis
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-(FSAR) Report Section 6.2.1 indicated that the internal containment pressure in the event of a Loss of Coolant Accident (LOCA) would be 15.8-psig, versus the design valve of '15.0 psig. The analytical model used to predict the internal containment pressure considered the mass of ice in i
the ice condenser starting and stopping times for the various containment spray pumps, volume of. water in the refueling water storage tank, and time required for operators to perform various procedural actioni.
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. containment pressure analysis, which resulted in a 15.0 psig containment pressure, was performed using the following assumptions:
Westinghouse WCAP-8246 containment model
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RHR pumps, in containment spray mode (ND Spray), initiated at 3600
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seconds after event 2.22 E+6 pounds mass of ice
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Containment spray (NS Spray) pumps off at 2193 seconds and stopped
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for 374 seconds
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Refueling Water. Storage Tank (FWST) initial volume =.350,000_ gallons
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FWST Volume at' low alarm = 80,000 gallons
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- FWST Volume at 10-10 alarm = 10,000 gallons
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y However, the initial FWST volume is 372,100 callons (per Technical i
Specification 3.5.5), the volume at low alarm t. Aually 117,094 gallons,
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while the-volume! at lo-lo alarm is 13,121 gallons.
In August 1985,.
Westinghouse recalculated the containment pressure using the WCAP-10325
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containment model to justify a reduction in the ice mass.
This calcula-
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tion disclosed that the higher _FWST volumes would probably shift the NS j
pump-off time with respect to ice melt off which would result in a higher containment pressure of 15.8 psig.
The immediate licensee's corrective
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-. action to resolve this problem was to revise the times specified-in f
g" affected emergency procedures (EPs) for no spray initiation from 3600 to-3000-seconds.
The inspector reviewed procedure EP-1A 5000.15.1, Response j
to High Containment Pressure, step 8, and verified that the no containment J
spray initiation was specified as 3000 seconds. Other corrective actions-by the licensee-involved reevaluation of the EP for transfer to cold leg recirculation to determine if it would be permissible to allow the NS
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pumps to continue running during the transfer, and recalculation of the i
FWST level setpoint.
The licensee performed this reevaluation and
reported the results to the NRC in a letter dated September 3,1987, Subject:
Licensee Event Report 369/85-29.
The licensee stated that it
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was not necessary to adjust the FWST level setpoints based on their reevaluation.
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Within this area, no violations or deviations were identified.
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Action on Previous Inspection Findings (92701)
a.
(Closed) Unresolved -Item (URI) (370/85-41-01):
Decrease in Core
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-Delta Temp - Evaluate < Westinghouse Analysis
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Six weeks after power escalation commenced for Cycle 2 Unit 2 experienced a gradual decrease of approximately one-degree Fahrenheit
in the indicated Delta T,(temperature) across the core. The concern regarding this ind dent involved possible underpredicM on of reactor thermal power,- and a ootential impact on FSAR accidents which take
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credit for a reactor trip on over temperature delta T or over power delta T trip functions. The inspector reviewed a-Westinghouse report titled:
Analysis Of Corlant Delta T Reduction Observed at McGuire
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Unit 2, dated October 30, 1985.
The Westinghouse report concluded that the cause of the delta T reduction might have been caused by changes _in hot' leg temperature streaming patterns, possible slight
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_ reduction in thermal power, and the usual instrument drift.
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water venturi fouling may have also contributed to the delta T decrease.
Westinghouse concluded that the reduction in delta T was within the allowed margin, and there was no impact on the safety analysis.
Westinghouse recommended that the channel calibration to set the over temperature setpoint, over power setpoint and T average L
channels _be performed at the beginning of each cycle upon completion o
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m of'the precision heat' balance. The licensee submitted a request in a-letter < dated May~9, 1988, to amend the Technical Specifications (TS)
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- to clarify several areas in the TS as a result of-evaluating the
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delta T incident.
The licensee submitted additional information to i
the NRC-in letters dated, April 1,1988, and January 5,1989, in
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response to an. NRC request.
The licensee proposes to amend TS-Section 4.2.3.5 by specifying that the reactor coolant system total
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flow shall be determined by the precision heat balance measurerent at the -beginning of 'each cycle -in lieu of-once per 18 months as is
.-I currently required and amending TS Table 4.3-1 and referenced notes to require the over temperature delta T and over power delte T calibrations to be performed following the completion of the
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precision heat balance, in -lieu of one per 18 months.
These TS
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6mendment requests are currently being-reviewed by the NRC Office of
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- Nuclear Reactor: Regulation.
The licensee has administratively
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controlled-performing the precision heat balance measurements at the i
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beginning; of each cycle to conform with the recommendations of the
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Westinghouse report.
This also complies with the current TS requirement of. performing the precision heat balance once per 18 months.1The inspector examined performance test (PT) procedure i
PT/1/A/4150/13 and PT/0/A/4150/13, NC Flow Calculation, completed October '7,1986; December 4,1987; December 14,1988; and May 22, 1989, for Unit I and on August 8,1988 and October 17, 1989, for Unit 2.
These pts were performed at the beginning of the cycles per the Westinghouse recommendations, i
b.
(Closed) Inspector Followup Item (IFI) (369/87-40-02):
PORV Corrections q
Closure force for the pressurizer power operated relief valves (PORVs)
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is provided by spring force and air pressure.
Since the air pressure
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was provided by a non-qualified system, ~ the licensee implemented a j
design modification,' number NSM-MG-1-2102, to provide a seismically i
and thermally qualified air (nitrogen) supply to.these valves.
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inspector examined the quality records documenting implementation of this modification during an inspection-conducted in August 1989 which is documented in NRC Inspection Report Number 50-369/89-25.
The licensee has also completed a design study to detarmine how \\alve packing changes reduce stem friction, and if spring force, wiehout air assist, would be sufficient to obtain closure of the PROVs. The inspector examined the completed design study, number MGOS-0132/90,
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Pressurizer Power Operated Relief Valve Packing Study, dated September 1989.
This study reconnends a modification t3 the packiag whi;h the licensee has determined, through studies cor. ducted in a
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fossil plant, should reduce stem friction sufficiently to preclude
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the need for air assist to obtain PORV closure.
Ihe study recommended installing the modified packing on one P0P.V. testing its performance for approximately one year, and performing an additional evaluation at that time.
However, based on completion of the
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qualified air supply to the valves, the licensee's corrective actions
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are completed to resolve any safety problems.
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5(0 pen):IFI (370/87-40-02): PORY Corrections The inspector examined the safety evaluation for Nuclear Station Modification NSM 2-2102 which," when implemented, will provide a-qualified air supply to the Unit 2 oressurizer PROVs.
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plans to implement this modification in the upcoming Unit 2 outage.
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IFI' 370/87-40-02 will-remain open pending review by NRC of - the t
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completed modification.
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(Closed) URI=(369,370/88-31-18):- Include Certain Check Valves _in the-l
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IW Program This item was previously -examined during an inspection and _is documented in NRC Inspection Report Numbers 50-369, 370/89-25.
The item ' concerned not testing check valves 1-SA-5 and 1 SA-6 in the Inservice Testing -(IST) surveillance program.
Subsequent to. the x
inspection -the licensee provided an alternative program for this testing which is documented in Relief Request RR-SA-1.-
The licensee has conducted a detailed design study to determine if any edditional check. valves should be included in the surveillance testing program.
The licensee currently tests 1100 valves under '.he IST program. The inspector examined the design study, titled MGDS-0106 IWV Program Evaluation, dated October 31, 1989.
The study is comprehensive and included all types of. valves. The design study recommended adding-to the existing inservice inspection program 171 check valves, 95 relief valves, and 19; air operator valves per unit plus 10 check valves shared between. the units, and performing local leak rate tests on 8 '
additional valves.
The design study considered the requirements of
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the proposed ASME standard, OM 10, Inservice Testing of Valves in Light-Water Reactor Power Plants, and also included valves in' the safe shutdown system (SSS).
The licensee's IST and onsite performance group are currently reviewing the completed design study and. preparing procedures to ' incorporate the additional testing requirements.
However, the need for including the SSS valves in the -
i these valves perform Appendix R (g reexamined since the majority offire pro inservice testing program _ is bein
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not need to be included in the inservice testing program.
The unresolved item was originally identified as a result of not including check valves 1 SA-5 and 1 SA-6 in the IST program, however, NRC.will followup on the resolution of the Design Study 106 recommendations during future inspectiont.
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(Closed) IFI (369,370/88-31-19):
Followup of Check Valve Design Study-
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Design Study 106, discussed in paragraph 3.d above, included a review of check valve design and acceptable testing methods followup on.any
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revisio'ns to check valve testing will be part of IFI 369,370/90-12-01
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-(Closed)IFI(369,370/88-31-20): Vehify Testing of Relief Valves This. item'was erroneously indicated to be closed in NRC Inspection -
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Report _ Numbers 50-369,370/89-25.
The licensee has developed a'listL
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of.'all safety-related relief valves under design study 106.
The
.-inspector reviewed procedure MP/0/A/7650/37 Relief Valve Set
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s Pressure' Testing land Adjustment.
This procedure contains the test procedure, prerequisites, and acceptance criteria for" testing of.
fsafety relief. valves.
The licensee has also developed a list of
nonsafety-related : relief valves which they. also plan to include in-the relief valve testing program.
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(Closed) IFI (369,370/88-31-21):
Verify Improvements in Testing of JAir Operated-' Valves.
Review of test data for air-operated valves showed that valve stroke'
times (VSTs) were erratic.
The licensee's test procedures were inconsistent regarding the requirements for recording VSTs.
Review of Unit:1 and 2 auxiliary feedwater air-operated valves CA-20AB,
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CA-27A,.'CA-328, CA-36AB, CA-408, CA-44B, CA-4BAB,- CA-52AB, CA-56A, CA-60A, and CA-64AB disclosed cases where the VSTs varied by more
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than 50 percent or were not recorded.
However, the licensee had been
granted an exemption from ASME Section Xi requimments for some of these cases.' Subsequent to the inspection,-the licensee reviewed these test procedures to clarify requirements for recording stroke times, and calculating the variation in strcke time.from test to test.
The inspector examined the procedures listed below which specify:the requirements for stroke timing of the auxiliary feedwater
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system air operated valves.
- The inspector verified.that the procedures contained acceptance criteria for valve stroke times, the requirements for comparing-the current stoke times with previous results to determine the percent change in valve stroke times, and requirements for increasing the test frequency when the percent change in the VST exceeds the acceptance criteria.; Procedures examined were as follows:
PT 1/A/4252/02A CA Valve Stroke Timing - Quarterly
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1A Motor Driven Pump Flowpath
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PT 1/A/4252/02B CA Valve Stroke Timing - Quarterly
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1B Motor Driven Pump Flowpath PT 2/A/4252/02A CA Valve Stroke Timing - Quarterly i
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2A Motor Driven Pump Flowpath PT 2/A/4252/02B CA Valves Stroke Timing - Quarterly
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28 Motor Driven Pump Flowpath
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PT 1/A/4252/034 CA Train A Valve Stroke Timing
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and PT 2/A/4252/03A-Quarterly Turbine Driven Pump Flowpath
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' The' inspector reviewed the completed procedures listed below to
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. verify - the revised program for air-operated valves was being
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m implemented. Completed procedures examined were as.follows:
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PT 1/A/4252/2A, - completed January 1989, March 1989,~ May 1989,
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June 1989 September 1989 December 1989, and May 1990 i~"
PT 2/A/4252/2A, completed September ;189, October 1989,
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November 1989, January 1990, and April LO PT 2/A/4252/3A, completed October 1989, November 1989,
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c bruary 1990, and April 1990 e
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Based on review of the above completed procedure, the inspector verified that the licensee has implemented ASME Section XI require-ments for these valves.
The-licensee has also identified additional s
air operated valves in Design Study 106 which may need to beLincluded in this-program.
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Exit Interview The _ inspection scope and results were summarized on. June 7,1990, with
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those' persons indicated in paragraph 1.
The inspectors described the
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- areas. inspected and discussed in. ' detail-the -inspection results listed below.
. Proprietary information is not contained in this. report.
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Dissenting comments were not received from the licensee.
' Licensee management was informed that the following items were closed:
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LER 369/85-29, paragraph 2
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- URI: 370/85-41-01, paragraph 3.a
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IFI 369/87-40-02,. paragraph 3.b
URI-369,370/88-31-18.- paragraph 3.d IFI 369, 370/88-31-19 - paragraph 3.e IFI 369, 370/88-31-21, paragraph 3.f
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IFI 369, 370/88-31-27, paragraph e.g
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