IR 05000369/1981030
| ML20039A065 | |
| Person / Time | |
|---|---|
| Site: | McGuire, Mcguire  |
| Issue date: | 11/20/1981 |
| From: | Bemis P, Bryant J, Falconer D, Graham M, Jape F NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20039A059 | List: |
| References | |
| 50-369-81-30, 50-370-81-18, NUDOCS 8112160189 | |
| Download: ML20039A065 (11) | |
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UNITED STATES
'8 NUCLEAR REGULATORY COMMISSION-o
e REGION 11
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101 MARIETTA ST, N.W., SUIT E 3100 o
ATLANTA, GEORGIA 30303
Report Nos. 50-369/81-30 50-370/81-18-Licensee: Duke Power Company 422 South Church Street Charlotte, NC 28242 Facility Name: ficGuire Docket Nos. 50-369 and 50-370
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License No. NPF-9 and CPPR-84 Inspection at Lake Norman, NC
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Inspector:
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Datd Sigfied
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Inspector:
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Dat6 Sighed Inspector:
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///2c/d7 Approved by:
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'.A. 'Bryant, Seqf>6n Chief, Division of Date~ Signed Resident and Reactor Project Inspection i
SUtil1ARY Inspection on September 15, 1981 - October 15, 1981 Areas Inspected.
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This routine, announced inspection involved 300 resident inspector-hours on site in the areas of operational safety, maintenance, surveillance, power ascension, security, LER review, and followup on inspector identified items.
Results Of the seven areas inspected. no violations or deviations were identified.
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8112160189 811123 PDR ADOCK 05000369'
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DETAILS 1.
Persons Contacted-Licensee Employees
- H. D. McIntosh, Station Manager R. Wilkinson, Superintendent of Administration G. Cage, Superintendent of Operations N. McGraw, Operations M. Sample, Projects and Licensing
- D. B. Lampke, Projects and licensing
- D. M. Franks, Quality Assurance
- C. M. Fish, Administration
- W. A. Rondlett, Security
- R. E. Jones, Security-I
'Other licensee employees contacted included technicians, operators, mechanics, security force members, and office personnel.
- Attended exit interview 2.
Exit Interview The inspection scope and findings were summarized on October 16,1981 with those persons indicated in paragraph 1 above.
3.
Licensee Action on Previous Inspection Findings The inspector reviewed the licensee's response to the following violations and noted that the corrective actions had been taken as described and appeared to be adequate to prevent recurrence.
Based on the inspectors review, these items are closed, a.
(Closed) Violation (81-13-03): VUCDT pumped to environment.
b.
(Closed) Violation (81-16-02):
Failure to follow tagging procedures resulting in removal of both trains of control room air from service.
4.
Unresolved Items Unresolved itenis were not identified during this inspection.
5.
Operational Safety The inspector observed operational activities in the plant and the control room throughout the inspection interval.
Shift operation and turnovers were
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observed, and control room, tagging, and work request logs were reviewed.
The status of instrument calibration, equipment tags, and radiation work permits was verified. Selected technical specification operating conditions were independently verified through confirmation of component position or operability or through observation of indicated parameters. Operations, maintenance perfonnance, health physics, chemistry and security personnel were observed at work. Operational areas reviewed in depth are discussed in detail in the following paragraphs.
6.
flaintenance liaintenance activities were observed in progress throughout the inspection interval.
The inspector verified that these activities were accortplished by qualified personnel using approved procedures.
Radiation controls, fire
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prevention measures, and QA/QC hold points were observed as appropriate.
Test equipment used was verified to be calibrated, and data recorded were compared to those observed. The following maintenance activities were observed and reviewed in depth.
a.
NV-219 replacement. NV-219 is a three-inch, manually operated valve in the charging system. Af ter unsuccessful attempts to repair a recurrent leakage problem, a modification was authorized to replace the valve with a different brand.
The inspector reviewed the modification package, witnessed the isolation and tag-out of the valve, and observed the establishment of radiation controls and fire prevention measures while work was in progress. Hydrostatic testing of completed work was witnessed, and discussions of the work and worker qualification were held with craftsmen and foreman.
The following procedures in use were reviewed:
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IIP /0/A/7600/28 Borg Warner Gate Valves MP/0/B/7650/09 Cutting, Welding, Open Flame Safety MP/0/A/7650/52 Controlling procedure for Piping flodifications f1P/0/A/7650/55 Hydrostatic Testing of Class B and C sy >tems SRUP 81-31 b.
ND-1 leakage. ND-1 is the first isolation valve in the Residual Heat Removal System suction line off the Reactor Coolant System. While the unit was shutdown, ND-1 was found to be leaking with a body to bonnett leak.
The licensee's design group performed an analysis which would allow the drilling of up to 10 holes into the seal ring area of the valve (above the pressure boundary).
The purpose of these holes was to allow the introduction of Furmanite in an attempt to seal the leak:
Furmanite is the trade name of a fiberous sealant which uses a grease vehicle for introduction into the araa to be sealed.
The licensee produ::ed the recov~ry analysis and paperwork for use in a safety class system. Af ter tN sealant was injected into the ninth hole the leak was stoppe.
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NY 242 leakage. NV242 is a manual valve located in the charging system which is used as isolation for the reactor coolant pump seal backpres-sure supply valve and return to service testing.
The valve was re-placed and the leakage stopped.
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7.
Surveillance Surveillance activities were-observed in progress throughout the inspection, interval. The inspector verified that these activities were accomplished by qualified personnel using approved procedares.
Test equipment used was verified to be calibrated and data recorded were. compared to those observed where appropriate radiation controls were observed in effect. The following surveillance activities were observed in greater depth.
a.
Ice Basket Weighing.
On September 30 - October 2,1981, PT/0/A/42200/
18 Ice Bed Analysis and HP/0/A/7150/07 Ice Condenser Intermediate Deck Doors, Corrective Maintenance, were performed to accomplish the sur-veillance requirements for ice condenser operability outlined in Technical Specification 4.6.5.1.
The inspector observed portions of both procedures in use and verified that basket selection confomed to the distribution required by Technical Specifications.
The inspector noted that additional baskets were weighted as required when light baskets were found, and that the final calculated weight met the acceptance criteria.
b.
The following surveillances were observed and the -data was reviewed for adequacy:
1.
PT 0/A/4600/14C Source Range Neutron Flux Functional 2.
PT 1/A/4350/03A Electrical Power Systems Alignment Verification 3.
PT 0/A/4601/02 Protection System Channel II Functional 4.
PT 1/A/4252/04 CA Valve Movement Test
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5.
PT 0/A/4600/14A Power Range NIS Functional 6.
PT 1/A/4252/01 Turbine Driven CA Pump Performance Test 7.
PT 1/A/4600/08 Precritical Surveillance Items for Unit S/U 8.
PT 1/A/4150/14 PORV Channel Functional Test 9.
PT 1/A/4202/01A Kf Pump 1A Performance Test 10.
PT 1/A/4150/17 Pressurizer Heaters Capacity Tests Based on this observation and review, no violations or deviations were identified.
8.
Power Ascension The inspector witnessed selected portions of power ascension testing at the 20-30% plateaus.
The inspector verified that tests were performed as required by the FSAR, that approved procedures were in use, and that per-sonnel performing the procedures had been properly briefed and were familiar with the test.
Further, the inspector independently verified selected
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4; limiting conditions required for performance of the test and verified that acceptance criteria were met.
The following tests were witnessed and reviewed in depth, a.
TP/1/A/2650/04 Unit Load Steady State b.
TP/1/A/2150/14 Steam Generator Water Hammer Test c.
TP/1/A/2600/11 Pressurizer Pressure and Level d.
TP/1/A/2650/03 Loss of Control Room The loss of control room test was performed on September 14 and September 17,-1981. The unsuccessful attempt to complete the test on September 14 resulted in a primary system cooldown of 117 F in one hour. This event is discussed in paragraph 9.
The second performance of the test' resulted in successful shutdown and cooldown. The inspector observed the following procedures in use:
TP/1/A/2650/03 Loss of Control Room Test AP/1/A/5500/17 Loss of Control Room 0P/1/A/6100/04 Shutdown Outside Control Room.
Pending resolution of the items discussed in paragraph 9, the inspector has no further questions in this area.
No violations were identified in the power ascertion program review.
9.
Cooldown during loss of Control Room test On September 14, 1981, during performance of the loss of control room test, operators had difficulty controlling the auxiliary feed system. An ex-cessive amount of feedwater was pumped to the steam generator, resulting in
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an abnormally fast cooldown rate and a safety injection.
As part of the test, the reactor was tripped from outside the control room.
Almost immediately after the reactor was tripped, the main feedwater (CF)
isolation valves closed on coincident reactor trip and Tave less than 564 F.
The S/G levels dropped to the Lo-Lo level setpoint which resulted in auto-matic start signals of the three CA pumps. All three CA pumps started and pumped water'into the S/G's at maximum flow-rate since the CA throttle valves had been fully opened by the automatic start signal. Steam generator levels began increasing.
During the next several minutes operators -tried unsuccessfully to gain control of the CA throttle valves. As a result, the S/G's filled rapidly with the excessive CA flow.
Pressurizer level and NC primary (NC) pressure fell at a corresponding rate.
Operators started the second centrifugal charging pump to recover pres-surizer level and NC pressure.
As NC pressure dropped below the P-11 setpoint (1955 psig), operators blocked SI to prevent further complication of the transient.
When pressurizer level indication dropped below zero, all four reactor coolant pumps were tripped to avoid damage.
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As the two charging pumps refilled the pressurizer, NC pressure was also
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restored.
the minimum NC. pressure reached was approximately 1900 psig.
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When the NC pressure exceeded the P-11 setpoint, the SI block was cleared and SI was initiated by S/G low pressure. The main steam isolation valves (
closed due to the SI which helped stem the heat loss from the NC system.
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llecovery from the cooldown was interrupted by a second safety. injection
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initiation on Train-A when the reactor trip circuit breakers were reset
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without sufficient steam generator pressure.
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During the resulting transient, NC temperatures exceeded the 100*F/ hour
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' cooldown rate addressed in. Technical Specification 3.4.10.1.
Maximum W
temperature change during the hour following the reactor trip was 116.8 F.
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ExcessivA cooling by the S/G's, the large amount of water injected into the
- NC system during the initial level recovery and the tripping of the NC pumps-
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The inspector noted that the licensee complied with the Action Statement of
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, Technical Specification 3.4.10.1 in shutting down and performing an s
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.sngineeringlevaluation of the effects of the transient on the RCS structural
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integrity: The licensee and vendor concluded that continued use of the j- -
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system was-acceptable as the absolute temperatures remained high and the ASME Code tection III Appendix G stress units were not exceeded. The fatigue usage factor contribution for the transient was found to be negli-gible.
The griatest contributing factor to the magnitude of the transient was y~
inability of the' operators to control auxiliary feed addition.
Analysis of
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J the,e, vent resulted in identification of the following areas of concern:
ss ar ' Procedural deficiencies:
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' N1) Tripping the Reactor. CF isolation on low Tave' coincident with
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reactor trip will occur on virtually every reactor trip.
This
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response and the rapid S/G level drop which resulted in the
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~Vs automatic start of all three CA pumps are characteristic of
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Westinghouse units similar to McGuire Unit #1.
S/G levels dropped
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from the nonnal operating values to the Lo-Lo setpoint in less s
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than five seconds.
Since the reactor was tripped as soon as the first group of operators reached the reactor trip breakers, the transient was i
well underway when the second group of operators reached the r'
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. auxiliary shutdown panels (the auxiliary shutdown panels are much
further from the control room than the reactor trip breakers).
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The procedures in use, AP/1/A/5500/17, Loss of Control Room, and OP/1/A/6100/04, Shutdown Outside' Control Room, did not anticipate the transient in the steam generator or specify when the reactor
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(2) Auxiliary feed control valve resets
~0n an automatic start signal, the auxiliary feed control valves go fully open and remain in that position until reset.
The pro-cedures in use did not address reset of the auxiliary feed.
As a result of identification of this area of concern, both procedures were revised to require depressing the CA flow control reset switches before leaving the control room, and the establishment of communi-cations between the reactor trip breaker panels and auxiliary control panels prior to tripping the reactor. The order of the steps perfonned was changed to prevent initiation of the transient before the operators were at all panels.
b.
Design deficiency Resetting of the CA control valves can only be performed in the control room.
The reset may be locked in prior to tripping the reactor.
Therefore the transient on the plant need not be initiated before the operators are in place at the panels.
The need to perfonn any action in the control room represents a design deficiency.
The licensee has committed to install duplicate reset switches on the auxiliary shutdown panels. The modification will be completed by December 31, 1981.
In the interim, the revised copy of AP/1/A/5500/17 is in effect. This commitment is Inspector Follow-up Item 50-369/81-30-01. The same design change will aslo be in effect on Unit 2; Inspection Followup Item 50-370/81-18-01.
c.
Personnel error Some confusion was present during the test resulting from the rapidity and severity of the transient and the lack of experience in controlling the plant from the auxiliary shutdown panels. The operators were unaware of the rapid level drop characteristics of the S/G's following reactor trip and CF isolation.
Upon reaching the auxiliary shutdown panels, the operators tried unsuccessfully to reset the CA flow controls. Operators who were stationed in the control room dep,'essed the reset pushbuttons located there with no apparent results.
The CA flow logic has since been tested completely including the equipment, wiring and functional tests of the circuitry.
Logic involved with the two CA flow control reset switch modules is completely separate and the possibility of a double failure is judged to be remote.
It is possible that the reset worked after being depressed in the control room but the operators did not recognize it due to inadequate training or experience in controlling CA flow from the auxiliary panels.
The additional clarity provided by procedural changes appears to address this problem in the interim. The second loss of control room test was performed without incident.
In addition, Duke has planned to modify the auxiliary shutdown panels to clarify operation of equipment
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and monitoring of operational parameters. This modification will be reviewed at a later date (Inspector Follow-up Item 50-369/91-30-02).
Pending resolution of the above identified items, the inspector had no further questions in this area. This closes LER 50-369/81-151.
10. Zero Power Physics Test Review The inspector reviewed the following r.ompleted zero power physics test procedures and data packages:
a.
TP/1/A/2100/22, Zero Power Physics Controlling Procedure b.
TP/1/A/2150/03 A-H, Boron Endpoint Measurement Test c.
TP/1/A/2150/0/6A, RCCA Psuedo Ejected Rod Test d.
TP/1/A/2150/10, Stuck Rod Worth Measurement Test e.
TP/1/A/2150/11 A-E, Isothermal Temperature Coefficient of Reactivity Test f.
TP/1/A/2150/12A-D, Zero Power Flux !!ap The above test procedures and data packages were reviewed to verify the following:
a.
The procedure was annutated to identify all test changes.
b.
No test change altered the basic objectives of the test.
c.
Each test deficiency was resolved and the isolation was accepted by appropriate management.
d.
Any system or process changes necessitated by a test deficiency were properly documented and reviewed.
e.
Deficiencies that constituted a reportable occurrence as defined by Technical Specifications were properly reported.
f.
Data sheets were completed and individual test steps and data sheets were properly initialed and dated.
g.
All data were recorded as required and within acceptance toler-ances.
h.
The cognizant engineering function evaluated the test results and signified that the test demonstrated system met design require-ments and the results met established acceptance criteria.
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Personnel charged with the responsibility for review and acceptance of test results documented their review and acceptance.
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Within -the areas' inspected there were no violations or deviations identi-fled.
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The all rods withdrawn isothermal temperature coefficient of reactivity was
. measured to be less negative than.that required by the FSAR. The measured isothermal -temperature coefficient corresponded to a positive moderator-coefficient and pursuant to Technical Specification 3.1.1.3, the licensee
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established interim control rod withdrawal limits to maintain a negative
' moderator temperature coefficient during all mode 1 and mode 2 operations
' not covered by special test exception. Technical Specification 3.10.3.
The interim limits were placed in the appropriate operating procedures and shall remain in effect until sufficient fission product poisons are produced to preclude a positive moderator temperature coefficient.
This closes inspector follow-up item IFI-50-369/81-24-02.
11. LER Review The inspector reviewed the LERs listed below to verify that the report
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details met license requirements, identified the cause of the event,
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described corrective actions appropriate for the identified cause, and'
adequately addressed the event and any generic implications.
In addition, the inspector examined selected logs and records and interviewed licensee personnel to verify that the report was accurate, described corrective action were taken, and that the event was reviewed as required.
The following event reports were reviewed:
LER Title 81-12 NV-250 Surveillance 81-13 EMFs inoperable 81-14 High ground water level 81-15 Fire protection 81-16 Fire protection i
81-17 Fire protection -
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81-19 Station security system 81-20 EMF-49 y
81-21 EMF-33
81-27 Diesel generator inoperable 81-29 Inograble radwaste flow monitor 81-31--
EMF-33 81-32 Fire protection 81-33 Containment purge system ducting -
81-34 Diesel generator exhaust in TSC 81-38 Both NV pumps inoperable 81-39 PORV setpoint low 81-40 SSPS instrumentation inoperable:
81-41 Fire Protection 81-42 Seismic instrumentation inoperable
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81-43 SSPS instrumentation inoperable
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' 81-44 Control room ventilation 81-45 Fire protection 81-47 EMI-38 81-48 Steam generator shutdown instrumentation inoperable 81-50 Containment P > 0.3 psi 81-51 SSPS out-of service 81-53 Lower containment T high 81-59 Inoperable radwaste flow monitor 81-60 Cold leg accumulator level 81-62 Steam generator level instrumen-tation 81-63 Personnel air lock seals 81-64 Failure to follow surveillance
. procedure 81-65 Sump level 81-67 UHI boron 81-68 EMF - 35, 36, 37 81-69 Cold leg accumulator level-81-70 Cold leg accumulator level 81-72 ND isolated 81-73 ND-1 leakage 81-75 Control Room ventilation 81-76 EMF-33 81-77 Personnel air lock welds 81-79 EMF-33 81-81 High groundwater level 81-82 Fire protection 81-84 UHI check valves damaged 81-89 EMF-33 81-90 Remote shutdown instrumentation 81-92 Accident monitoring 81-94 Liquid radwaste 81-95 Liquid radwaste 81-97 Fire protection 81-98 CCW isolation 81-100 Meteorological instrumentation 81-102 Seismic instrumentation 81-103 Meteorological instrumentation 81-106 EMF-51B 81-107 Steam generator level instrumen-tation 81-108 EMF-33 81-109 Meteorological instrumentation 81-114 Aux feed pump valves81-115 Personnel air lock open
- 81-116 Centrifugal charging pump valve 81-119 Diesel generator failure 81-120 Containment isolation valve limit-
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switches81-121 CCW 81-122 BIT conceatration 81-123 BIT concentration 81-125 SSPS instrumentation 81-130
. Fire protection 81-132 RCS leakage 81-133 Remote shutdown monitoring instrumentation 81-134 EtlF-434 81-137 Control room ventilation 81-141 Meteorological instrumentation 81-142 Sump level 81-143
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EMF-43A 81-150 Steam generator sample line 12.
Review of Inspector Followup Items The following inspector followup items were reviewed a.
(Closed) IFI (50-369/81-13-02):
T.S. change for breaker testing. The June 1981 technical specification change included the test method issued by the licensee to perform required surveillances.
This item is closed.
b.
(Closed) IFI (50-369/81-13-04):
Radwaste procedure review. The licensee reviewed all radwaste procedures as committed, and made changes where appropriate. This item is closed.
c.
(Closed) IFI (50-369/81-16-06):
Radwaste procedure peview.
The licensee reviewed all radwaste procedures as cannitted, and made changes where appropriate. This item is closed.
d.
(Closed) IFI (50-369/81-21-01):
Power ascension.
The licensee made the committed changes to the power ascension controlling procedure.
This item is closed.
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