IR 05000324/1987013

Safety Insp Repts 50-325/87-13 & 50-324/87-13 on 870505-31. Violations Noted:Failure to Follow Integrated Leak Rate Procedure.Fuses for High Drywell Pressure Instrument Not Removed

ML20216C743
Person / Time
Site:	Brunswick
Issue date:	06/23/1987
From:	Fredrickson P, Garner L, Ruland W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20216C675	List: ML20216C671 ML20216C699 ML20216C743
References
50-324-87-13, 50-325-87-13, NUDOCS 8706300350
Download: ML20216C743 (11)

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' UNITED STATES -

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/ NUCLEAR REGULATORY COMMISSION.-

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101 MARIETTA STREET.N.W.

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Report'Nos. 50-325/13 and'50-324/87-13 n-

? Licensee:

Carolina Power and: light Company--

P. O. Box 1551-4

.Raleigh,-NC.27602 l

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IDocket'Nos. 50-325 and:50-324-License Nos. DPR-71 and.0PR-62

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' Facility Name: ' Brunswick 1 and 2-

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Inspection Conducted: 'May.5.- 31,.1987-b-NN '

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W. H. RVland Date Signed

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L.: W. Gh+ er Date Signed..

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- P. I., Fredridkson,' Secti.on. Chief.

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~' Division of-Reactor Projects

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~ SUMMARY

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. This routine safety inspection involved the ~ areas - of maintenance Scope:

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. observation, surveillance ' observation, operational safety;-l verification, ESF'

System walkdown,.and. Unit'2 forced outages.

3, Results: One. violation was-identified: Failure to follow'i.ntegrated leak rate

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procedure -' the. fus'es for a high.drywell pressure. instrument 'were not removed.

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B706300350 870'624

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Persons Contacted

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Licensee Employees

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P. Howe, Vice Precident - Brupswick Nuclear Project C. Dietz, General Manager - Brunswick Nuclear Project i

T. Wyllie, Manager'- Engineering and Construction" (

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G. Olivar, Manager - Site Planning and Control J. Holder, Manager - Outages R. Eckstein, Mana r% Operatior.sr - Technical Support, t:

E. Bishop, Manager, L.-Jones, Director's, Quality Assuchnce (QA)/ Quality Control (QC)

R. Helme, Director - Onsite Nuclear Safety - BSEP J. O'Sullivan, Manager - Maintenance G. Cheatham, Manager - Env Wonmental & Radiation Control

J. Smith, Manager m Admir,istrative Support

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K. Enzor, D1 rector -iRdgufatory Compliance

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R. Groover, Manager 2 Project Construction

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s V. Wagoner, Director xIPP,S/Long-Rangq Planning A. Hegler, Superintendent - Operations 1

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W. Hogle, Engineering Supervisor

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B. Wilson, Engineerinn Supervisor

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B. Parks, Engineeringi$upervisor s-x

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W. Biggs, Principal Engineer 7-R. Creech, I&C/ Electrical Maintenance Supervisor (Unit 2)

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d R. Warden, I&C/ Electrical Maintenance Supervisor (Unit 1)

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t t W. Dorman, Supervisor QA,

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W. Hatcher, Supervisor Security

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L R. Kitchen, MechanicalJ aintenance Supervisor (Unit 2)

C. Treubel, Mechanical faintenance Supervisor (Unit 1)

EJ R. Poulk, Senior NRC Regulatory Specialist w

W. Murray, Senior Engineer %y Specialist

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D. Novotny, Senior Regulator

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Nuclear Licensing Unit

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Other licensee employees contacted included construction craftsmen, engineers, technicians, operators, office personnel, and security force members.

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ExitInterview(30703)

The inspection scope and findings were summarized on May 29, 1987, with the

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general manager.

One Violation, failure to meet initial conditions

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required by the Unit 1 integrated leak rate test (paragraph 6), was discussed in' detail.

An Unresolved Item (paragraph 8.b)), concerning an inadequate procedure, was also discussed.

The licensee acknowledged the findings without exception. The licensee did not identify as proprietcry anyt, of the materials provided to or reviewed by the inspectors during the inspection.

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3.

Followup on Previous Enforcement Matters (92702)

Not inspected.

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Ma) baanceiChservation (62703)

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F The inspectod observed maintenande1 activities and ' reviewed records to

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verifhthat twor.K ' 's conducted ih accordat:ce withiapproved procedures,

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Techn11al Specifi$ 'tions, andr' applicable industry. codes and standards. The

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"inspecWrs. also verifjedithat:. dedundanticomponents were operable;

dministrit}ye; c6htlols' were followedptagouts were adequate; personnel

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y6 pre qualifded;' correct; replacement.partFw%re.lised;- radiological co'ntrols-J

.were proper; fire. protection was ade'qQte;- qdality control hold ; points

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.were.:Ladequate,and. observed; adequat9 post-maintenance' testin'g; was

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N performed; and' independent verification requiremsnts were L implemented.

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The. inspectors independently verified'that selected equipment.was properly.

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1 returned toLservice.

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Dutstandingiwork requests: wyre re'vieQd to knsureL that' the' licensee gave c-priority' to safe'ty-related maintenanc i,

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'The Linspectors observed / reviewed portions. of thifoHowingL maintenance-

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activities:

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,Repla ement of High Press ~ure Coolant Injection (HPCI)

a : tiydraulic Actuator Gear Drive Gears,.

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87-ANDX1-Disassemble 1-E41-F006 to Determine Cduse of Low' Actuator-g

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' Repair ofz24821-TR-R614, Safety Relief Valve (SRV) Taiipipe.

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Temperature Chart Recorder.

-MI-10-517C~

HPCI Hy'draulic Actuator Gear Drive and Speed Pickup Gear Assembly Inspection.

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r The' licensee contracted w'ith Babcock and Wilcox '(B&W): to take valve

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signatures and perform 1 diagnostics on 30 valves.

Twenty.seven of these are on the. HPCI and Reactor.. Core : Isolation-Cooling '(RCIC)-

l i3~.f systems. "The ' device used tis designated as a Motor Actuator Characte -

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. rized ?(MAC).

The MAQ machine geasures and/or. calculates spring.

f(Qand'orqueswitchactuations>

orce,. stem thrust, worm shaft displacement, motor current, and limit

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The raicro processor data is capable of.

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R being' transferred into; graphic-form and/or hard copy.

The project

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' repaired those valveshand/or actuators -which were - identified as

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having ' deficiencies. The licensay has'. supplied to the.' i n spector,

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copies of the MAC data 'for these' valves. The ~ inspector plans to i

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review this data and document <that review as well as a summary of the

. problems' which required: correhtion in a future report.

This.is an

f Inspector Followup Item:

Revi s Valve. Diagnostic Test Results 325/87-13-01).

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The inspector witnessed connection of sensors and/or testing of the HPCI suppression pool inboard suction isolation. valve, E41-F042,. the HPCI steam admission valve, E41-F001 and the HPCI inboard injection valve ' E41-F006.

During performance of the F006 valve test,.the measured stem thrust indicated. about one half of that specified by-the manufacturer (Limitorque).

Subsequent testing on other valves showed Lsimilar results.

The licensee, B&W and Limitorque have concluded that on large and/or fast actuating actuators, the correla-tion between the measured parameter, strain gauge output, and stem thrust is not correct. B&W, in cooperation with Limitorque, -.will-develop a new ccrrelation.

The licensee plans to have the stem thrust information on the affected actuators re-evaluated when the new correlation is available. The inspector will review the data as part of.the above IFI.

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Incorrect Spring Pack in Residual Heat' Removal (RHR) Valves ihe licensee declared both divisions of Suppression Pool Cooling.

inoperable for Unit 2 based on problems found in Unit'l Limitorque motor operators.

Units 1 and 2 were in operational conditions. 5 (refueling) and 1 (power operation), respectively.

The licensee had replaced the torque switch for valve 1-E11-F024A,

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Unit 1, Division I, Suppression Pool Cooling Isolation Valve, on l

April 5, '1987. The licensee had found the problem with the torque switch while trying to cycle the valve after routine preventative maintenance on the motor control center.

The valve was stopping '

.immediately.after dual indication was seen on the. main control board

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while shutting the valve. Main tenance, under work request 87-AKUP1, found the. torque switch contacts pitted and with high resistance.

-The valve was cycled and the motor-operator inspected. af ter the torque switch was replaced;:the results were satisfactory.

The 1-E11-F^24A valve then failed.to completely shut during perfor-mance of the quarterly RHR system operability test, PT-8.'2. 2. c, on

'May 2, 1987.' The licensee initiallyLfound the torque out current too

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low' with the required torque switch setting of 2.5.

The licensee

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continued tn troubleshoot the problem the next 20 days' in between l

vesse t ~ hydrostatic test. Integrated Leak Rate Test (ILRT), replace-

ment of au/iliary contractor work, and work on the Unit 1 HPCI F006 i

valve. 'On May 15. 1987, maintenance found that the~ handwheel torque'

required to open1the torque' switch in the' shut direction was too low and they suspected a bad spring pack. After the ILRT, on May 22, the licensee found the spring pack to be incorrect: it had 9 believille washers instead of 11.

Since all ' four F024 valves '(' both units) were originally purchased

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under 'the same order, the licensee inspected the Unit 1 F024B valve and found a liaht. spring pack also installed.

The licensee then i

i declared the vn.. 2 valves iroperable at 2:30 p.m. on May 22, 1987.

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The licensee concluded that a lighter than'specified spring pack was-installed in the Unit 2 F024A valve and adjusted the torque switch from 2-1/2 to 3-3/4 to compensate for it. The licensee has had the concurrence of Limitorque for the adjustment. The licensee concluded

.that ' the F0248 valve in Unit 2 has' a heavier spring pack installed

with characteristics similar to the specified' spring pack.

The licensee determined the spring pack characteristics of the Unit 2 valves by-observing valve handwheel torque when the torque switch operated while closing the valve.

The licensee had returned the Unit 2 F024A to operable status'at 9:00 p.m. on May 22 and the F024B at 12 midnight on May 22, complying with

.the applicable sections of technical specifications.

New spring packs will be installed when available.

The licensee has reported that they have no readily accessible record l

of. ever replacing the spring packs, indicating that the valves'may have been delivered with spring packs not matching Limitorque's own documentation.

The vendor program branch will followup on the generic issues rair,ed by the event. The inspectors will continue to j

follow the licensee's actions in the Motor Operated Valve'(MOV) area-as described in paragraph 4.a.

No violations or deviations were identified.

5.

Surveillance Observation (61726)

The inspectors observed surveillance testing required by Technical'

Specifications.

Through observation and record review, the inspectors

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tests conformed to Technical Specification ' requirements; administrative controls were followed; personne1' were qualified; instru-mentation was calibrated; and data. was accurate and complete.

The inspectors independently verified selected test results and proper return to' service of equipment.

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The inspectors witnessed / reviewed portions of the following test activi-I ties:

IMST-ADS 23R Automatic Depressurization System (ADS) Safety Relief Valve Primary Position Channel Calibration.

IMST-APRM29Q - Average Power Range Honitor (APRM) Flow Bias Flow Units C&D l

Channel Calibration.

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l 1MST-RPS34R Reactor Protection System (RPS) Main Steam Line Isolation Valve Closure Circuit Response Time.

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2MST-HPCI27M HPCI and RCIC Condensate Storage Tank Low Water Level

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Instrument Channel Calibration.

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PT-20.6 Drywell to Torus Leak Rate Test.

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PT-78.2 Liquid Radwaste Radioactivity Effluent Monitor Channel Calibration (012-RM-K604).

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During performance of IMST-RPS34R on May 14,.1987, the inspector observed

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that two procedural steps were incorrectly performed. Step 7.4.7 requires l

test leads to be connected to test panel points TPB-31 and 33; instead, I

they were attached to.the adjacent column of test points.

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in a failure, on the first attempt, to obtain the necessary data to -

determine the response time. During troubleshooting of the failure to-obtain data,' the system.was partially restored to-normal by reinstalling I

fuse C71A-F3F. Upon detection of the improperly connected test leads, a

technician resumed the procedure at the beginning of section 7.4.

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However, step '7.4.10,. remove fuse C71-F3F, was overlooked.

Thus, 'a subsequent run of the test again failed to measure the response time.

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This deficiency was corrected.

In addition, the inspector also observed a technician correct a step he i

had just indicated as'not applicable.

Step 7.4.35 is not applicable if j

the reactor mode switch is in RUN. - The technician initially indicated the j

step as not applicable, and then realized his mistake and performed the

step (the mode switch was in REFUEL at the time).

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l None of.the above items had a potential for resulting in an incorrect L

recnonse time test.

Each one of the above items'was associated with a

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different technician.

These items were discussed with maintenance l

l supervision.

Perhaps, because the unit was shutdown, the technicians may i

l-have been less attentive.

This concern was expressed to management.

L Maintenance management stated 'that the items would be discussed with

.j members of the surveillance staff to ensure that a " shutdown mentality"

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was not being developed.

No violations or deviations were identified.

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Operational Safety Verification (71707)

j The inspectors verified conformance with regulatory requirements by direct observations of activities, facility tours, discussions with personnel, reviewing of records and independent verification of safety system status.

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l The inspectors verified that control room manning requirements of 10 CFR l

50.54 and the technical specifications were met.

Control room, shift-

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supervisor and clearance logs were reviewed to obtain information

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concerning operating trends and out of service safety systems to ensure that there were no conflicts with Technical Specifications Limiting-l Conditions for Operations. Direct observations were conducted of control

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room panels, ' instrumentation and recorder traces important to safety to l'

verify operability and that parameters were within Technical Specification limits. The inspectors observed shift turnovers to verify that cnntinuity of system status was maintained.

The inspectors verified the status of selected control room annunciators.

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0perability' of. selected Engineered Safety Feature (ESF) trains were

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verified by. insuring that: each accessible valve in the flow path was in

its: correct: position; 'each power supply and breaker, including control-

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l room, fuses, were. aligned;for components that must activate upon initiation signal;.removallof power from those ESF motor-operated valves, so identi-fied by Technical Spec.ifications, was completed; there was no leakage of

major components; there =was proper lubrication and ' cooling water available; and a condition did not exist which might prevent fulfillment-

.of? the' system's L functional requirements.

Instrumentation essential.to

~ system actuati.on or performance was verified operable: by observing l

on-scale : indication and proper instrument' valve lineup, if accessible.

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Thel inspectors verified that the licensee's health physics policies /

procedures were followed. This included.a review of area surveys, radia--

tion work permits, posting, and instrument calibration.

LThe inspectors verified' that:

the security' organization was properly

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manned and security personnel were capable of performing their assigned functions;. persons :and packages were checked prior to entry into the-

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protected area. (PA); vehicles were properly authorized, searched and-

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escorted.within the PA; persons within the PA displayed photo identifica-

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tion badges;; personnel in vital areas were authorized; and effective q

compensatory ' measures were employed when required, n

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The ' inspectors also observed plant housekeeping control s,. verified.

position of certain: containment isolation valves, checked a clearance, and verified the operability ~of onsite and offsite' emergency: power sources, a..

'High Drywell Pressure Instrument Left Energized During ILRT

'The inspector found that Unit ' 1 Drywell Pressure'.High Instrument:

1-E11-N011A was reading greater than 5 psig while N011B, C & D -were less than 0 psig during an ILRT with containment pressure at about 48 psig at 3:00 p.m. on May 18, 1987.

The inspector questioned the.

d licensee concerning the discrepancy. The licensee reported that..the

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fuses should have been pulled for -all those instruments during the

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ILRT lineup. On May 19, 1987, the licensee informed the inspector that due to a labeling problem, the wrong fuses had been removed.

The licensee had removed the fuses for the Core Spray and Residual i

Heat Removal Injection Permissive Reactor Steam Dome Pressure Instru-i ment 1-B21-N021A instead of the E11-PT-N011A fuses. :The fuse labels were reversed in the back of the Emergency Core Cooling System (ECCS)

cabinet 1-XU-63 that contained both instruments.

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The ILRT was not affected by the error. No valves changed positions r

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as a result of the error. The N011A instrument could not by itself

cause an inadvertent actuation and was only required to be operable j

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in conditions 1, 2 and 3.

The unit was in operational condition 5

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(refueling).during the ILRT.

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-i The steam dome pressure instrument, B21-PT-N021A, had been inadver-tently de-energized, was required to. be operable, per technical specification-(TS)'3.3.3, in operational condition 5.

The applicable i

action statement requires the licensee to declare the associated ECCS systems inoperable. Both Core Spray (CS) and RHR Division I should have been declared inoperable. The CS ACTION statement in TS 3.5.3.1 for operational condition 5 requires that at least one Low Pressure Coolant Injection (LPCI) system is operable within four hours.

The licensee had complied with the TS based on the inspector's review of the' operator's log and the ILRT procedure.

The fuses were not labeled in accordance with the applicable drawing-for the XU-63 cabinet, F-39031, sheet 6, Revision 8.

Fuses in block RF, which were labeled B21-F1-A and B21-F2-A on the drawing, were labeled E11-F1-A and E11-F2-A in the. cabinet.

Fuses in block RT, which were' labeled E11-F1-A and E11-F2-A on the drawing, were labeled 821-F1-A and B21-F2-A in the cabinet.

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The Integrated Primary Containment Leak Rate Test (IPCLRT) procedure, PT-20.5, Rev.13, section VI, Initial Conditions, step DD., requires

"high drywell pressure instruments defeated by completing Section ' A of Appendix I.

Appendix I, step A.1.e, states, " Pull the fuses in -

i Figure.1-1 and place under clearance."

Figure I-1 requires that fuses E11-F1-A and E11-F2-A be pulled for instrument E11-PT-N011A.

i Contrary to. the. above, on May 18, 1987, fuses E11-F1-A and E11-F2-A

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for N011A were not pulled prior to the start of the -IPCLRT.

This failure to. follow procedure is a violation of TS 6.8.1.c',

which requires that written procedures be established and implemented covering. surveillance and test activities of safety-related equipment, i.e.,

the IPCLRT.

This is a Violation:

Failure to Properly: Implement Surveillance Procedure (325/87-13-02).

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On May 8, the inspector observed that the anchor plate associated with an adjustable rigid strut (support No. E41-3PG62), on the Unit 1 HPCI injection line had one side of the anchor plate not in contact l

with the concrete ceiling.

It appeared that both anchor bolts had been partially pulled out of.the concrete. Work request 87-AQBL1 has been issued to repair the condition prior to the unit refueling startup. The support was installed in 1985 as part of Plant Modifi-cation PM-84-381. This modification moved the inboard HPCI injection valve'into the main steam line valve pit. The licensee has concluded i

i that the as found condition did not render the support inoperable.

J An ' evaluation is being conducted to determine the probable cause of i

the as found condition.

The inspector will review this evaluation I

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On May 21, the inspector observed the following items on Unit 1.

The drywell-suppression chamber vacuum breaker X18F had its backup close j

position indication switch, LS-4, inoperable. The suppression spray

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header support brackets had E five nuts with less than' full thread l

engagement and eight additional nuts were not in contact with the bracket top. ' These items were corrected under work requests 87-ARDF1 and 87-ARDC1, respectively, bne violation and no deviations were identified.

7.

ESF System Walkdown (71710)

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f During the report period, the inspector performed an inspection of the i

accessible components of the Unit 2 Automatic Depressurization System (ADS).

This verification included the following items:

Initiation and permissive instrumentation are valved into service.

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Backup reactor building air compressors (air supply to ADS valves)

are energized.and operable.

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Backup nitrogen supply system is pressurized and operable.

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Both temperature and acoustical monitors of the ADS valve discharge

. piping are operable.

.The ADS actuation logic.3 energized.

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ADS override switches are in the NORM position.

j No. problem was 'found which would render the system inoperable. Two items were.noted.

Sporadically during the month, point 10, associated with Safety Relief I

Valve (SRV) B21-F013K, of the tail pipe temperature chart recorder,

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2-B21-TR-R614, has been fluctuating around the alarm setpoint. The normal value for this point during this cycle 'has been right below-the alarm point.

However, with the onset of warmer weather, the corresponding increase in the average drywell temperature has resulted in this point's

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normal operating band to sometimes overlap with the alarm setpoint. This

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results' in an erroneous annunciation on the main control panel of a leaking SRV. Sometimes this condition exists for several continuous hours before it clears. Because this could potentially mask a real problem with another SRV, the licensee is in the process of evaluating the feasibility of defeating this input when it is in the alarm state.

i Also, on May 15, the inspector observed that the 2-B21-TR-R614 chart recorder was printing scattered points versus its usual slightly wavy lines; however, the wavy lines were still discernible.

The licensee declared the recorder inoperable and repaired it under work request 87-AQIF1. The chart had been initialed by two different operators while it was printing in'this erratic pattern before the inspector observed the j

condition.

One of these signatures was right after the problem had

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sta rted.'

Because there was an abrupt change in the recorder's perfor-mance, which was clearly visible on the chart, the inspector believes that this condition should have been detected by this. operator.

This concern was discussed with operations management.

At the time, the primary o

indication of SRV positions, the acoustic monitoring system, 'was: fully j

operational and TS required minimum number of channels was met during the l

time the recorder was malfunctioning.

No violations or deviations we e identified.

8.

Unit 2 Forced Outages'(93702)

6.

Vacuum Leak in Main Condenser y

The ' licensee rapidly removed the generator off the grid due to condenser ' vacuum problems.

On May 22, 1987, at 4:40 p.m.,

vacuum started decreasing and off gas flow increased.

Power reduction was started at,4:45 p.m. and the main turbine was manually tripped at 5:16 p.m.

The licensee found the vacuum. leak:

the low pressure turbine bearing dirty oil drain header that passes through the

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condenser was' cracked.

The pipe was plugged and capped.

The generator _ was synchronized to the grid on May 23,.1987 at 4:16 p.m.

Main generator reverse power.resulted in an' anticipatory start of all four emergency Diesel Generators (DG). The DGs were not required to tie onto'the' emergency buses because offsite power was available.

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Water in Off gas Line On May.27, 1987, Unit 2 experienced a forced outage due to decreasing vacuum. caused' by water in. the off gas header.

At 3:55 a.m.,

condenser vacuum took a 2 to'3 inches sudden decrease. The operators reduced power by running back the recirculation pumps to minimum

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speed and inserting control rods'. ' An auxiliary ' operator found the.

.l off gas loop ' seal reservoir fil1 valve, 2-0G-SV-4906, wide open,

a thereby allowing demineralized water to flow into the off gas header.

He manually. isolated the fill line. At 4:20 a.m., the turbine was manually tripped. Reactor power was maintained at approximately 20%

until the off gas header was drained.

The turbine generator was i

synchronized back to the grid at 12:02 p.m.

The event was attributed to an inadequate surveillance procedure and equipment failure.

The problem occurred during performance of PT-4.1.8, Off gas System

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Automatic Isolation Operability Check,.on Unit 1, as required by TS , 4.3.5.9.

The procedure steps caused the Unit 2 off gas loop seal reservoir tank drain valve, 2-0G-SV-4907, to open.

The procedure

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should have resulted in the Unit i valve, 1-0G-SV-4907, to open. The opening of the Unit 2 drain and resulting level loss, caused the fill

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valve, 2-0G-SV-4906, to open.

This valve then stuck in the open j

position allowing demineralized water to back up into the 30 minute j

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off gas holdup _line. This partially restricted off gas flow, thereby causing the decreasing vacuum. On May 28, 1987, a temporary revision was Jissued to 'PT-4.l.8 to correct its errors and provide additional

clarification The temporary revision added steps to open the Unit 1 l

valve.

Specifically, the. steps require lifting of wire LY6 from terminal 58 and jumping terminal'56 to 57 in panel XQ9. Apparently, when the Unit 1 and 2 procedures, PT-4.1.8-1 and-2, respectively, were combined into one PT-4.18 (revision 24 dated February 2,1987),

these steps were deleted.

The inspectors need to review the licensee's procedure development i

process in; general and how the PT-4.1.8 revision was done in parti-cular.

Pending the inspectors'- ' review, this will remain an Unresolved Item:

PT-4.1.8, Off gas Automatic Isolation Operability j

Check Procedure Inadequate (325/87-13-03).

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i No violations or deviations were identified.

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