IR 05000259/1981009
| ML18025B518 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 04/08/1981 |
| From: | Cantrell F, Chase J, Paulk G, Sullivan R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML18025B512 | List: |
| References | |
| 50-259-81-09, 50-259-81-9, 50-260-81-09, 50-260-81-9, 50-296-81-09, 50-296-81-9, NUDOCS 8105210209 | |
| Download: ML18025B518 (38) | |
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I UNITED STATES NUCLEAR REGULATORY COMMISSION
REGION II
101 MARIETTAST., N.W SUITE 3100 ATLANTA,GEORGIA 30303 Report Nos.
50-259/81-09, 50-Z60/81-09 and 50-296/81-09 Licensee:
.Valley Authority 500A Chestnut Street Tower II Chattanooga, TN 37401 Facility:
Browns Ferry Nuclear Plant.
Oocket Nos.
50-259, 50-260 and 50-296
License Nos.
OPR-33, OPR-52, and OPR-68 Inspection at Browns Ferry site near Athens, Alabama.
Inspectors:
R. F. Sul ivan Senior Resident Inspector Oate Signed
. W. Cha e, esident Inspector Oate Signed G. L. Pau k, R sident Inspector Approved, by:
F.
S.. Cantrell, Secti p', Oivision of Resident and React Project Inspection SUMMARY Inspection on February 26 - March 25, 1981 Areas Inspected Oate Signed te. Si ned This routine inspection involved 140 resident inspector-hours on site in the areas of operational safety, reportable occurrences, plant physical protection, radiation protection, reactor trips, surveillance testing, maintenance, startup report review and. reactor water level instrumentation.
Results Of the 9 areas. inspected, no violations or deviations were found in 7 areas,
violations were found in 2 areas; (Maintenance performed on safety-rel'ated equip-ment without proper documentation, paragraph 6;
Inadequate administrative t
controls on keys. to high radiation doors, paragraph 12).
SX 052Xop 09
DETAILS 1.
Persons Contacted Licensee Employees 2'.
H. L. Abercrombie, Power Plant, Superintendent J.
R. Bynum, Assistant Power Plant Superintendent J.
L. Harness, Assistant Power Plant Superintendent R. T. Smith, guality Assurance. Supervisor R. G. Metke, Engineering Section Supervisor A. L. Clement, Chemical Unit Superviso~
D. C. Mims, Engineering and Test Unit Supervisor R. G. Cockrell, Reactor Engineering Unit Supervisor J. B. Studdard, Operations Section" Supervisor
'A. L. Burnette, Assistant Operations Supervisor.
Ray Hunkapillar, Assistant Operations Supervisor T.L. Chinn, Plant Compliance Supervisor.
M. W. Haney, Mechanical Maintenance Section Supervisor J. A. Teague, Electrical Maintenance Section Supervisor J.
R.. Pittman, Instrument Maintenance Section Supervisor J.
E.. Swindell, Outage Director.
8. Howard,, Plant Health Physics R. E. Jackson, Chief, Public. Safety R. Co'le, gA,Site Representative Officer of Power Other licensee employees, contacted included licensed senior reactor operators and, reactor operators, auxil'iary operators; craftsmen, techni-cians', public safety officers, gA personnel and engineering personnel.
I Management Interviews Site. management interviews were conducted on February 27, March 6, 13, and 20, 1981, with the Power Plant Superintendent and/or his. Assistant Superintendents and other selected members. of his staff.
The inspectors summarized the scope and findings of their inspection activities.
The licensee.
was informed of two violations identified during this report period..
3..
Licensee Action on Previous Inspection Findings Not,inspected.
Unresol'ved Items
Unresolved items:
are matters about which information is required to determine whether they are acceptable or may involve violations or deviations.
An unresolved item identified by this inspection is discussed
. in paragraph ' Operational Safety'he inspectors kept=informed on a daily basis of the overall plant status, and any significant safety matters related to plant operations.
Oaily discussions were held each morning with plant management and various members of the plant. operating staff.
C The inspectors made frequent visits to the control room such that each was visited at. least daily when. an inspecto~
was on, site.
Observations included instrument readings, setpoints and recordings; status of operating systems; status and. alignments of emergency standby systems; purpose of temporary tags, on equipment. controls and switches; annunciator alarms; adherence to procedures; adherence to limiting conditions for operations; temporary alterations in effect; daily journals and data sheet entries; and control room manning=.
This. inspection activity also included numerous informal discussions with operators and theii supervisors, General plant tours-were, conducted on at least.
a weekly basis.
Portions of'he turbine building, each reactor building and outside areas were visited.
Observations included valve positions and system alignment; snubber and hanger conditions; instrument.
readings; housekeeping; radiation area controls;, tag controls.on equipment; work activities in progress", vital area t
controIs; personnel badging, personnel search and escort; and vehicle search and.escort.
Informal discussions were held with selected plant personnel in their functional areas during these tours.
On February 21, 1981, after shutting down Unit 1 for routine maintenance:,
the.,control room-operator reported erratic, operation of the scram discharge'olume (SDV) continuous. water monitoring system (CMS)
west-header.
The.
operator reported that the CMS alarmed, but cleared almost immediately.
The alarm* should have stayed in-for five to ten minutes..
Investigation by the licensee concluded that the possible problem with the CMS was.
a broken center conductor in the cable from. the transducer to the ultrasonic instru-ment.
The cable was replaced; and tested.
No further problems have been identified..
The inspector determined. through interviews with quality assurance (gA)
personnel, that the defective cable, was replaced on February 20, 1981, to reduce; intermittent. al'arms caused by welding in the area.
gA personnel stated that no work authorization (trouble report (TR)) was issued for the replacement of the cable, nor was there a post-maintenance test document.
filled out, although the test was reported to have been performed.
In addition, na. TR was fi.ljed out for the replacement of a defective cable on February.
21, 1981, but a, post-maintenance test was performed and properly documented.
The. failure to document work performed on safety-related equipment was identified to the Power Plant. Superintendent on March 20, 1981, as an
.apparent violation of Technical Specification 6.3.A which requires written procedures be approved and adhered to for maintenance operations which coul have an affect on the safety of the reactor.
The Power Plant Superintendent accepted the apparent violation with no comment. (259/81-09-01)
An inspector conducted a plant tour on March 18, 1981, and noticed that the pressure suppression chamber head tank (PSC), was isolated for all three units.
The PSC system fs used to maintain the applicable emergency core
.cooling system (ECCS) filled with water from the pump discharge check valves to the last normally closed valve fn the ihjection path.
Ma/ntaining the ECCS discharge lines full of water is necessary to prevent water hammer upon system initiation and reduces the injection time.
The condensate system is used to backup the function of the PSC, system.
Technical Specification 3.5.H states.
that the PSC head tank is normally aIigned to serve the residual heat removal and core spray systems.
The inspector reported this discrepancy between technical specification and actual system configuration to the plant Staff.
Plant staff determined that the PSC head. tank system has not been operational since being initially installed several years ago; The condensate backup system-is the primary means for ensure the RHR and CS discharge piping is maintained full of water.
Plant staff stated that they were not, sure whether the-PSC head tank system had ever been operationally tested.
The licensee-agreed to." evaluate the PSC'ystem operability and the apparent discrepancies between technical specifications and actual plant conditions.
(Unresolved 259/81-09-01).
Reportabl e. Occurrences The below listed. licensee event reports (LER's) were reviewed to determine if'he information provided. met NRC reporting. requirements.
The. determi-nation included adequacy of event description and corrective action taken or planned, existence of potential generic problems and the relative safety significance of each event.
Additional inplant reviews and discussion with plant personnel as appropriate were conducted for those reported indicated by an asterisk, Lfg 259-8101
"259-8106
"259-8108 260-8039
"260-8047
"260"8053"
"260-8055
"260"8056 260-8057 Oate 1/28/81 2/11'/81 2/26/81 2/20/81 11/20/80 I/2/81 1/6/81 1/6/81 1/13/81 Event Core spray differential switch setting outside of I imits RHR area, cooling fan temperature switch failed to operate.
Violation of-Secondary Containment Orywell control air valve would not close.
Condenser'circulating water pumps did not-provide dilution of water to environment Leak on ZC RHR heat exchanger Core spray discharge pressure switch was found set outside-of technical specification limit.
Reactor water level switch was found set above technical specification limit.
Hain steam line low pressure switch was found inoperabl "260"8108
"296-8107 296-8108
"296-'8109 296-8110 3/10/81 2/20/81 2/24/81 2/27/81 3/6/81 RCIC steam throttle valve would not open MSIV limit switch found set out of technical specification limit RHR pump inoperable Oiesel generator 3C would not stop Orywell high pressure switch set point outside of 1 fmft Within the areas reviewed, no violations or deviations were identified.
Reactor-Trips The inspectors-reviewed activities associated with the below listed reactor trips during this report period.
The review included determination of cause, safety significance, performance of personnel and systems, and corrective action.
The inspectors examined instrument, recordings, computer.
printouts,. operations journal entries, scram reports and had discussions with operations, maintenance and-engineering support personnel as appro-,
priate.
Of February 21, 1981, Unit 1 was manually tripped at 1:45. a.m.
from 35K, power to swap a main transformer inorder to have a spare for use on Unit, 3.
Systems involved. in the shutdown performed as designed.
On February 23, 1981, Unit, 3 was manually tripped at 10:01 p.m.
from 39K power to swap a main transformer.
While shutdown maintenance was performed on an.MG-set which had. developed excessive vibration.
The MG set supplied power to LPCI valve. motors.
The LPCI system involved performed as designed during the shutdown.
On February 25, 1981, Unit 2 was manually-tripped at 1:48 a.m.
in order to swap main transformers.
Systems involved fn the.
shutdown performed as designed.
On February 28, 1981, Unit 2 tripped at 1:58 p.m.
from 87%%u~ power during surveillance testing on the Average'ower Range Monitors (APRM').
One APRM
'as removed from bypass before it had been returned to the operate mode which introduced a half-scram.
When the operator reached to return this channel to bypass, he, unintentionally brushed, against another switch which was in bypass for another channel which was out of the operate mode.
This resulted in a
second half-scram, completing actfon for a, full scram and reactor trip.
Systems involved in the. shutdown performed as intended.
On March 7; 1981, Unit. 2 tripped at 4:OS a.-m.
from 99K power during sur veill'ance testing of the-main steam isolation valves (MSIV).
The test involved 10K closure, but when the test pushbutton for B inboard MSIV was released, the valve contfnued to close.
High steam flow in the other three lines initiated. a group I. isolation and a reactor trip..
Both RCIC and HPIC were manually started to contro1 reactor 'water level.
Relief'alves were, manually operated to control pressure.
Five different relief valves were used.
All systems performed satisfactorily.
Closure time on "B" inboar MSIV using the test button was found to be faster than the other valves; however, the normal closure time on an isolation signal was not affected.
The test procedure was revised to reduce power before test operating
"8" inboard MSIV until a drywell entry can be made to correct the test timing.
On March 8, 1981, Unit 3 tripped at 8:11 p.m.
from 1005 power from high flux due to failure of the steam pressure sensing line in the
"8" pressure regulator.
Sensing a drop in pressure, the control system initiated turbine control valve closure which produced a pressure spike and the resulting flux spike.
The turbine bypass valves did not operate automatically because of the erroneous signal from the "8" pressure regulator and five main steam relief (MSRV's) valves operated to control reactor pressure.
No problems-developed with the MSRV's..
No emergency core cooling systems were initiated.
Required systems performed as designed.
On March 13, 1981, Unit 2 tripped at 10:59 p.m.
from 54K power which was initiated. by a main turbine-trip.
The turbine trip was initiated by the feedwater control system which responded to a high water level indication in the reactor vessel.
The. reactor vessel water level was actually normal and the'roblem was associated with reactor water level instrumentation.
Paragraph ll describes the instrumentation problem.
Reactor safety functions performed as designed although some redundancy in initiating channels were lost.
No violations or deficiencies were identified.
Plant Physical Protection During the course of routine inspection activities, the inspectors made observations of certain plant physical protection activities.
These included. personnel badging, personnel search and, escort, vehicle. search and escort, communications and vital area access. control.
No violations or deviations were identified within the areas inspected.
Surveillance Testing Observation The. inspector observed the performance of the below listed surveillance procedures.
The inspection consisted of a
review of the procedure for technical adequacy, conformance to technical specifications., verification of test,. instrument calibration, observation on the conduct, of the test, the removal and the return of the system to. service'nd a review of the test data.
Surveillance:-Number Title SI; 4.7.0. 1. S-1(A)
Hydrogen-Oxygen System Isolation Valve Operability SI 4.T.C.
Secondary Containment Within the areas inspected no violations or-deviation were identifie Review of Unit 2 Startup Report The inspector reviewed Startup Report for Unit 2 i!uel cycle 4, dated February 13, 1981.
This review Has,conduci;ed to verT'fy that the in-formation repor ted was technically adequate; and that the reporting requirements of Technical Specification 6.7.1.a.
were satisfied..
Mithin the areas inspected no violations or deviations were identified.
Reactor Mater Level Instrumentation Following the trip of Unit 2 reactor on Parch 13, 1981, one of the two instruments which measures reactor vessel level inside the shroud for low pressure coolant injection (LPCI) operation (LITS 3-52) was found with its
. equalizer valve partially open.
A review of the events associated with the scram revealed that with the equalizing valve open, several safety-related channels may not. perform their required function under different rea'ctor flow condition.
As. reactor power was reduced by-reducing recirculation pump speed invalid inputs were se'nt to one channel of the associated. logic system for the following safety functions; low.level scram protection, groups 2,3, and 6'rimary conta'i'nment isolation system (PCIS) automatic depresgurizatfon system (AOS)'confirmatory low level, and 'trip signals for HPCI and RCIC systems.
In addition, a false high reactor water level signal was sent. to the feedwater control circuits and the turbine trip circuits and a main turbine trip was initiated as LITS 3-52'ffected two inputs to the feedwater control circuit and the two out. of three.logic was satisfied.
The turbine trip caused a reactor scram, initiated by stop valve closure.
The actual reactor vessel water level was normal.
Reactor power was 54%%uo. at the time of the scram, however, power had been reduced from 100%%u,'ver the preceeding
minute period to perform unrelated maintenance on a recirculation pump M-G set.
The reactor water level indicators connected to the reference leg shared. by LITS 3-52 pegged high when the plant was at 54% power.
An analysis of the scram details, provided the following information.
The shroud range Yarway, LITS 3-52, is directly affected by recirculation pump flow as the variable leg level is sensed at the diffuser section of an-instrumented jet pump.
During normal power operation, the shroud range Yarway is pegged high at 200 plus, inches.
Special measurements taken at full power indicated a -1.3 inch -pressure difference between the variable and.reference.
legs with a slightly higher pressure on the variable leg side.
This results in a nearly stable pressure difference and when the Yarway equalizer valve is open a slight flow would exist from variable to reference leg.
This would not affect any level'nstrumentation as the small flow.
would overflow the condensing'ot of the reference-leg and return to the reactor vessel.
In addition, the operator would not be aware of the abnormal position of the equalizer valve for the shroud Yarway LITS 3-52 since the control room instrumentation would look hormal.
Thus, if the equalizer valve were cracked open or leaking by its seat at 100/o power the operator would not be aware that a. potentially unsafe condition exist When reactor power level is reduced by decreasing recirculation pump speed the pressure characteristics change across the jet pump diffuser.
At minimum recirculation pump speed the measured pressure difference with LITS 3-52 equalizer valve cracked open indicated 2 psi with the higher pressure on the reference leg.
This delta presssure acts to reduce the reference leg level which results in the level instruments indicating a
higher than normal level. Therefore, as reactor level power was reduced to 545 by decreasing recirculation pump speed, the reference leg level decreased enough to make the reactor vessel level instruments (GEMACS) read a false 54 inches (the Ilevel for turbine trip).
!
The following reactor vessel level instruments share the reference leg with the shroud Yarway:
LT3-53 and LT3-206 (feedwater Control),
LIS 3-184 (AOS interl ock),
LIS 3.-203A
LIS 3-203A/B (Low Level Scram)
and group 2, 3,
and
PCIS.'he shroud Yarway supplies containment spray permissive signals,to the containment spray logic.
Thirty-six hours prior to the scram, a
surveillance instruction was conducted.
on LITS 3-52 for calibration.
The licensee is evaluating the possible causes and corrective action to be taken with respects to this.
occurrence.
General Electric, Service Information letter (SIL) No.
335, da~ed July '1980, delineates. the-'passible safety related significance of leaving an equalizer valve on a differential pressure sensor partially open.
Plant Staff did not have a. copy. of'SIL 335 prior to the scram and were not.
aware of its contents.
The inspector discussed with plant management the importance of having assurance.
that the equalizer valves are closed or not leaking through 'on the two instruments which measure water level'n the shroud because of the unique situation where such a condition is not readily apparent when operating, at full power recirculation flow rates.
The plant superintendent was aware. that this was an undesirable situation and had instructed plant personnel to devise a method to lock these valves in the closed position.
This item wil) remainopen-until 1icen'see reviews and corrective actions have been comp 1 eted.
(260/81-09-01 ).
12.
Radiation Protection On Parch 12, 1981, the licensee informed the resident inspectors that high radiation door ¹178 located-in the Rad'aste building was found unlocked and unattended by health physics personnel.
This was discovered during the periodic check,.of high.radiqtjqn doqr~.on parch 11, 1981, at Opppogimately
.1700.
Jnyeqtjgati'on by the licensee djd not reveal when oI who left door
¹178 'un1ocked.
An inspector determined that, the control of high radiation areas (Areas greater than 1000 mrem/hr) were not being performed in accordance with Technical, Specification 6.3.D.2 The key to the high radiation doors is
'
'
.13.
being maintained by health physics and health physic technicians may utilize the key without any formal permission (i,e, checkout sheet)
from the shift engineer, The resident inspector identified to the Plant Superintendent on March 24, 1981, that failure to adequately control'he high radiation door key was an apparent violation of Technical Specification 6.3.0.2 which requires the shift engineer on duty to administratively control the key.
The licensee committed to developing a
procedure for controllin'g the key.
(259/81-09"02).
On March 5, 1981, short-lived airborne radioactivity was detected by a
continuous air monitor (CAM's) in the turbine building.
Portions of the turbine building were evacuated and a total of 81'ersonnel received con-tamination on the body.
The contamination quickly decayed to nondetectable levels.
The principal nuclides were Rb-88 and Cs-138.
The airborne activity Was legs than the maximum permissible concentration (MPC).
The maximum body contaimination was: received by. a pipe fitter which measured.
250,000 OPM.
Whole body counts were made on three individuals, including the pipe fitter, with negative results.
TVA concluded there were no signi-ficant exposures as a result of thi;s; occurrence.
The licensee informed the inspectors: that the source of the activity was from a. small off-gas sample line-in Unit 1.
A new piping "T" was being installed at the off-gas station without having the line isolated.
The licensee stated the-work was performed without proper authorization or wor k instructions.
Corrective action was promptly initiated with personnel involved in.that the individuals received additional instructions in this area.
The peportabiljty requirement of this item was determined not to be required.
~ The inspectors had no further questions and were sat'isfied with the corrective action taken by the licensee.
Maintenance The inspectors held discussions with maj.ntenance personnel, engineers and plant management concerning the repair weld for the (EHC) electro-hydraulic system pressur e transmitter in Unit 2. and the modification of the off gas sample line in Unit. 3.
A, review of the documents involved in the job was performed and discussions held at the job site.
No violations or deviations were=identified in the above are ~
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UNITED STATES NUCLEAR RfGULATORYCOMMISSION
REGION II
10'I MARIETTAST., N.W., SUITE 3100 ATLANTA,GEORGIA 30303 gpR 0 9 f981 Tennessee Val 1 ey Authority ATTN:
H. G. Parris Manager of Power 500A Chestnut Street Tower II Chattanooga, TN 37401 Gentlemen:
Subject:
Report Nos. 50-259/81-09, 50-260/81-09 and 50-296/81-09 This refers to the routine safety inspection conducted by R.
F. Sullivan of this office on February 26, to March 25',
1981, of activities authorized by NRC Operating License Nos.
DPR-33, DPR-52, and DPR-68 for the Browns Ferry facility.
Our preliminary findings were discussed with Herb Abercrombie at the conclusion of the inspection.
Areas examined during the inspection and our findings are discussed in the enclosed inspection report.
Within these areas, the inspection consisted of selective examinations of procedures and representative records, interviews with personnel, and observations by the inspectors.
During the inspection, it was found that certain activities under your license appear to violate NRC requirements.
These items and references to pertinent requirements are listed in the Notice of Violation enclosed herewith as Appendix A.
Elements to be included in your response are delineated in Appendix A.
One new unresolved item is identified in the enclosed inspection report.
This item will be examined during subsequent inspections.
In accordance with Section 2.790 of the NRC "Rules of Practice,"
Part 2, Title 10, Code of Federal Regulations, a
copy of this letter and the enclosed inspection report will be placed in the NRC Public Document Room. If this report contains any information that you believe to be proprietary, it is necessary that you make a written application within 20 days to this office to withhold such information from public disclosure.
Any such application must include the basis for claiming that the information is proprietary and the proprietary information should be contained in a separate part of the document.
If we do not hear from you in this regard within the specified period, the report will be placed in the Public Document Roo J I
~'
I
Tennessee Valley Authority Apo " g 1981 Should you have any questions concerning this letter, we will be glad to discuss them with you.
Sincerely, gc; R. C.
ewss, Acting Director Division of Resident and Reactor Project Inspection Enclosures:
l.
Appendix A, Notice of Violation
Inspection Report Nos. 50-259/81-09, 50-260/81-09, and 50-296/81-09 cc w/encl:
H. J.
Green, Division Director H. L. Abercrombie, Plant Superintendent R. E. Rogers, Project Engineer H. N. Culver, Chief, Nuclear Safety Review Staff
e APPENDiX A NOTICE OF VIOLATION Tennessee Val 1 ey Authority Browns Ferry 1, 2, and 3 Docket Nos. 50-259, 260, 4 296 License Nos. DPR"33, 52, & 68 As a result of the inspection conducted on February Z6, to March 25, 1981, and in accordance with the Interim Enforcement Policy,
FR 66754 (October 7, 1980),
the following violations were identified.
A.
Technical Specification 6.3.A.5 requires detailed.written procedures shall be prepared and adhered to for preventive or corrective maintenance operations which could have an effect on the safety of the reactor.
'Browns Ferry Standard Practice 6. 1, Performance of Maintenance, requires that routine maintenance will be documented by use of a trouble-report.
Contrary to the above, the documentation of maintenance by use of a trouble report was not met in that on February 20, and 21, 1981, maintenance was performed on the scram discharge volume continuous monitoring system without documentation of the work performed.
This is a Severity Level V Violation (Supplement I.E.) applicable Co'nit-1.
B.
Technical Specification 6.3.D.2 requires that each high radiation area in which the intensity of radiation is greater than 1000 mrem/hr shall be locked and the keys maintained under administrative control of the shift engineers on duty.
Contrary to the above, on March Z4, 1981, the keys to these areas were being maintained and used by health physics personnel and radwaste operators without being under the administrative control of the shift engineers on duty.
This is a Severity Level V Violation (Supplement I.E.).
Pursuant to the provisions of 10 CFR 2.201, you are hereby required to submit to this office within twenty-five days of the date of this Notice, a written state-ment or explanation in reply, including:
(1) admission or denial of the alleged violations; (2) the reasons for the violations if admitted; (3) the corrective steps which have been taken and the results achieved; (4) corrective steps which will be taken to avoid further violations; and (5) the date when full compliance will be achieved.
Under the authority of Section 182 of the Atomic Energy Act of 1954, as amended, this response shall be submitted under oath or affirmation.
Date:APR 0 9398>
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gd UNITED STATES NUCLEAR REGULATORY COMMISSION RMION II 101 MARIETTAST., N.W suITE 3100 ATLANTA,QEQRQIA 00303 Report Nos.
50-259/81-09, 50-Z60/81-09 and 50-296/81-09 Licensee:
.Valley Authorfty
.
500A Chestnut Street. Tower II Chattanooga, TN 37401 Facflity:
Browns Ferry Nuclear Plant Docket Nos.
50-Z59, 50-Z60 and 50-Z96 License Nos.
DPR-33, DPR-52, and DPR-68 Inspection at. Browns Ferry site near Athens, Alabama.
Inspectors:
R. F. Sul ivan Senior Resident Inspector Date Signed
. M. Cha e, esident Inspector G. L. Pau k, R sident Inspector Approved by:
F. S..Cantrell, Secti n
f, Division of Resident and React 'roject Inspection SUMMARY Date Signed
+'-
Date Signed te Si ned Inspectf'on on February 26 - March 25, 1981 Areas Inspected.
This routine inspection involved 140 resident inspector hours on site in the areas of operational safety, reportable occurrences, plant physical protection, radiation protection, reactor trips, surveillance testing, maintenance, startup report:review and reactor water level instrumentation.
Results Of the 9 areas inspected, no violations or deviations were found in 7 areas,
violations were found in 2 areas; (Maintenance performed on safety-related equip-ment without proper documentation, paragraph 6;
Inadequate administrative controls on keys. to high radiation doors, paragraph 12).
OETAL'LS 1.
Persons Contacted Licensee Employees H. L. Abercrombfe, Power Plant Superintendent J.
R. Bynum, Assistant Power Plant Superintendent J. L. Harness, Assfstant Power Plant Superintendent R. T. Smfth, gualfty Assurance.Supervisor R. G. Metke, Engineering Section Supervisor A. L. Clement, Chemical Unit Supervisor 0. C. Mfms, Engineerfng and Test Unit Supervisor R. G. Cockrell, Reactor Engfneerfng Unit Supervisor J. B. Studdard, Operations Sectfon Supervisor
'A'. L. Burnette, Assfstant Operations Supervisor Ray Hunkapillar, Assistant Operations Supervisor T.L. Chfnn, Plant Compliance Supervisor.
M. W. Haney, Mechanical Maintenance Section Superviso~
J. A. Teague, Electrical Maintenance Section Supervisor J.
R. Pittman, Instrument Maintenance Section Supervisor J.
E. Swindell, Outage Ofrector B. Howard, Plant. Health Physics R. E. Jackson, Chief, Public Safety R. Cole, gA Site Representative Ofi'icer of Power Other licensee employees contacted included licensed senior reactor operators and. reactor operators, auxiliary operators, craftsmen, techni-cians., public safety officers, gA'ersonnel and engineering personnel.
Z.
Management Interviews Site management interviews were conducted on February 27, March 6, 13, and 20, 1981, with the Power Plant Superfntendent and/or hfs Assistant Superintendents and other selected members.of his staff.
The inspectors summarized the scope and i'fndfngs of their inspection activities.
The licensee was informed of two violations identified during this report period.
3.
Licensee Action on Previous inspection Findings Not inspected.
Unresolved Etems Unresolved ftems are matters about which information i s required to determine whether they are acceptable or may involve violations or deviations.
An unresolved ftem fdentfffed by this inspection is discussed in paragraph ' Operational Safety The inspectors kept informed on a daily basfs of the overall plant status and any sfgnfficant safety matters related to plant operations.
Daily discussions were held each morning with plant management and various members of the plant operating staff.
The inspectors made frequent visits to the control room such that each was visited at least daily when an inspector was on site.
Observations included instrument readings, setpoints and recordings; status of operating systems; status and, alignments of emergency standby systems; purpose of temporary tags on equipment controls and switches; annuncfator alarms; adherenca to procedures; adherence to Ifmftfng conditions for operations; temporary alterations fn effect; daily journals and data sheet entries; and control room manning.
This inspection activity also included numerous informal discussions with operators and their supervisors.
General plant tours were conducted on at least.
a weekly basis.
Portions of'he turbine building, each reactor building and outside areas were visited.
Observations included valve positions and system alignment; snubber and hanger conditfons; instrument readings; housekeeping; radiation area controls; tag controls on equipment; work activities in progress; vital area controls; personnel badging, personnel search and escort; and vehicle search and escort.
Informal discussions were held with selected plant personnel in their functional areas during these tours.
On February Zl, 1981, after shutting down Unit 1 for routine maintenance; the. control room operator reported erratic operation of the scram discharge volume (SDV) continuous water monitoring system (CNS)
west: header.
The operator reported that the CPS alarmed, but cleared almost immediately.
The alarm should have stayed in for five to tan minutes.
Investigation by the licensee concluded that the possible problem with the CMS was, a broken center conductor in the: cable from-the transducer to the ultrasonic instru-ment.
The cable was replaced, and tested.
No further problems have been identified..
The inspector determined through interviews with quality assurance (gA)
personnel, that the defective cable was replaced on February 20, 1981, to reduce intermittent al'arms caused by welding in the area.
gA personnel stated that no work authorizatfon (trouble report (TR)) was issued for the replacement of the cable,* nor was there a postmafntananca test document filled out, although the test was reported to have been performed.
In addition, no. TR was ff,Ued out. for the replacement of a defective cable on February 21, 1981, but a. post~aintananca test, was performed and properly documented.
The failure to document work performed on safety-related equipment was identified to the Power Plant Superintendent on March ZO, 1981, as an apparent violation of Technfcal Specification 6.3.A which requfres written procedures be approved and. adhered to for maintenance operations which could
6.
have an affect on the safety of the reactor.
The Power Plant Superintendent accepted the apparent violation with no comment. (Z59/81-09-01)
An in'spector conducted a plant tour on Harch 18, 1981, and noticed that the pressure suppression chamber head tank (PSC)
was isolated for all three units.
The PSC system is used to maintain the applicable emergency core cool.ing system (ECCS) filled with water from the pump discharge check valves to the last normally closed valve in the i'njectfon path.
Majntaining the ECCS~<discharge lines full of water is necessary to prevent water hammer upon system.m, initiation and reduces the injection time.
The condensate system is
'sed
',to backup the function of the PSC system.
Technical Specification 3.5.H states that the PSC head tank is normally aligned to serve the residual heat removal and core spray systems.
The inspector reported this disci epancy between technical specification and actual system configuration to the plant Staff.
Plant staff determined that the PSC head tank system has not been operational since being initially installed several years ago; The condensate backup system is the primary means for ensure the RHR and CS discharge piping is mai'ntained full of water.
Plant staff stated that they were not sure whether the PSC head tank system had ever been operationally tested.
The licen'see agreed to evaluate the PSC system operability and the apparent discrepancies between technical specifications and actual plant condi tions. (Unresol ved 259/81-09-01).
Repoi table Occurrences The below listed licensee event reports (LER's) were reviewed to determine if the information provided met NRC reporting requirements.
The. determi-nation included adequacy of event description and corrective action taken or planned, existence of potential generic problems and the relative safety significance of each event.
Additional inplant reviews and discussion with plant> personnel as appropriate were conducted for those reported indicated by an',asterisk, 259-8101
"Z59-8106
"259-8108 Z60-8039
"260-8047
"260-8053'260-8055
"260-8056 260-8057 Oate 1/28/81 Z/11/81 Z/26/81 2/20/81 11/20/80 I/2/81 1/6/81 1/6/81 1/13/81 Event Core spray differential switch setting outside of I imits RHR area cooling fan temperature switch failed to operate.
Violation of'econdary Containment.
Orywe11 control air valve would not close Condenser circulating water pumps did not provide dilution of water to environment Leak on 2C RHR heat exchanger Core spray discharge pressure switch was found set outside of technical specification limit.
Reactor water level switch was found set above technical specification limit.
Hain steam line low pressure switch was found inoperable
Ol
e
"260-8108
"296-8107 296-8108
"296-8109 Z96-8110 3/10/81 RCIC steam throttle valve would not open Z/ZO/81 MSIV limit switch found set out of technical specification limit Z/Z4/81 RHR pump, inoperable 2/27/81 Oiesel generator 3C would not stop 3/6/81 OryweI1 high pressure switch set poiat outside of limit Within the areas reviewed, no vfolatfoas or deviatfons were fdentfffed.
Reactor Trips The fnspectors reviewed actfvities associated with the below listed reactor trips during this report period.
The review included determination of cause, safety significance, performance of personnel and systems, and corrective action.
The inspectors examined instrument recordings, computer.
printouts,. operatioas journal entries, scram reports and had discussioas with operations, maintenance and engineering support personnel as appro-pr iate.
Of February 21, 1981, Unit
was manually tripped at 1:45 a.m.
from 35K.
power to swap a main transformer inorder to have a spare for use on Unit 3.
Systems involved in the shutdown performed as designed.
On February 23, 1981, Unit 3 was manually tripped at-10:01 p.m.
from 395 power to swap a main transformer.
While shutdown maintenaace was performed on an.MG-set which had. developed excessive vibration.
The MG set supplied power to LPCI valve. motors.
The LPCI system involved performed as designed during the shutdown.
On February ZS, 1981, Unit 2 was-manually tripped at 1:48 a.m. in order to swap main transformers.
Systems i'avolved in the shutdown performed as designed.
On February Z8, 1981, Unit 2 tripped at 1:58 p.m.
from 875 power during surveillance testiag on the Average'ower Range Monitors {APRM's).
One APRM
'as removed from bypass before it had been returned to the operate mode which introduced a half-scram.
When the operator reached to return this channel to bypass, he uafntentfonally brushed against another switch which was in bypass for another channel which was out of the operate mode.
This resulted in a
second half-scram, completing action for a full sc~am and reactor trip.
Systems involved fn the shutdown performed as intended.
On March 7, 1981, Unit. 2 trfpped at 4:08 a.'m.
from 9'ower during sur-veillance testing of the main steam isolation valves (MSIV).
The test involved 105 closure, but when the test pushbutton for 8 inboard MSIV was released, the valve coatinued to close.
High steam flow fn the other three lines initiated a group I isolation and a reactor trip.
8oth RCIC and HPIC were manually started to control reactor water level.
Relief valves were manually operated to control pressure.
Five different relief valves were used.
All systems. performed satisfac orily.
Closure time on "8" inboard
MSIV using the test button was found to be faster than the other valves; however, the normal closure time on an isolation signal was not affected.
The test procedure was revised to reduce power before test operating
"8" inboard MSIV until a drywall entry can be made to correct the test timing.
On March 8, 1981, Unit 3 tripped at 8:11 p.m.
from 100K power from high flux due to failure of the steam pressure sensing,.line in the
"8" pressure regulator.
Sensing a drop in pressure, the control system initiated turbine control valve closure which produced a pressure spike and the resultfng flux spike.
The turbfne bypass valves did not operate automatically because of the erroneous signal from the "8" pressure regulator and. five main steam relief (MSRV's) valves operated to control reactor pressure.
No problems-.
developed with the MSRV's..
No emergency core cooling systems were initiated.
Required systems performed as designed.
On March 13, 1981, Un'it 2 tripped at 10:59 p.m.
from 54~ power which was initiated by a'ain turbine trip.
The turbine trip was initiated by the feedwatar control system which responded to a high water level indication in the reactor vessel.
The reactor vessel water level was actually normal and the problem was associated
'with reactor water level fnstrumentation.
Paragraph
describes the instrumentation problem.
Reactor safety functions performed as designed although some redundancy in initiating channels were lost;.
No violations or deffcfencies were identified.
Plant Physical Protection Ouring the course of routine inspection activities, the fnspectors made observations-of certain plant physical protection activities.
These included. personnel badging, personnel search and escort, vehicle search and escort, communications and vital area 'access control.
No violations or deviations were identified within the are'as inspected.
9.
Surveillance Testing. Observation The. fnspector observed the performance of the below listed surveillance procedures.
The inspection consisted of a
review: of the procedure for technical adequacy, conformance to technical specifications.,
ver ificatfon of test; instrument calibration, observation on the conduct of the test, the removal and the return of the system to service'nd a review of f:he test data.
Surveillance Number SE'. 4.7.0. l.8-L(A)
SI 4.7.C Title Hydrogen-Oxygen System Esolatf on Yal ve Operabi1 fty Secondary Containment Wfthfn the areas inspected no violations or deviation were identifie Review of Unit 2 Startup Report The inspector reviewed Startup Report for Unit 2 fuel cycl'e 4, dated February 13,,1981, This review was, conducted to verTfy that the in-formation reported was technically adequate, and that the reporting requirements of Technical Specification 6.7.1.a.
were satisfied.
Within the areas inspected no violations or deviations were identified.
Reactor Water Level Instrumentation Following the trip of Unit 2 reactor on March 13, 1981, one of the two instruments which measures reactor vessel level inside the shroud for low pressure coolant injection (LPCI) operation (LITS 3-52) was found with fts equalizer valve partially open.
A review of the events associated with the scram revealed that with the equalizing valve open, several safety"related channels may not perform their requfred function under different rea'ctor flow condition.
As reactor power was reduced by reducing recirculation pump speed invalid inputs were se'nt to one channel of the associated logic system for the following safety functions; lnw.level scram protection, groups 2,3, and '6 primary containment isolatf on system(PCI@automatjg depresgurfzatfon system (AOS) confirmatory low level, and trip signals for Hf CI and NCIC systems.
In addition, a false high reactor water level signal was sent to the feedwater control circuits and the turbine trip circuits and a main turbine trip was initiated as LETS 3-52 affected two inputs to the feedwater control circuit and the two out of three logic was satisfied.
The turbine trip caused a reactor scram, initiated by stop valve closure.
The actual reactor vessel water level was normal.
Reactor power was 54% at the time of the scram, however, power had been reduced from 1005 over the preceeding
minute period to perform unrelated maintenance on a recirculation pump M-G set.
The reactor water level indicators connected to the reference leg shared. by LETS 3-52 pegged high when the plant was at 545 power.
An analysis of the scram details provided the following information.
The shroud range Yarway, LITS 3-52,,is directly affected by recirculation pump flow as the variable leg level is sensed at the diffuser 'section of an.
instrumented jet pump.
Ouring normal power operation, the shroud range Yarway is pegged high at 200 plus inches.
Special measurements taken at full power indicated a -1.3 inch pressure difference between the variable and. reference legs with a slightly higher pressure on the variable leg side.
This results in a nearly stable pressure difference and when th'e Yarway equalizer valve fs open a slight flow would exist from variable to reference leg.
This would not affect any level instrumentation as the small flow would overflow the condensing pot, of the reference leg and return to the reactor vessel.
In addition, the operator would not be aware of the abnormal position.of the equalizer valve for the shroud Yarway LITS 3-52 since the control room instrumentation would look normal.
Thus, if the equalizer valve were cracked open or leaking by fts seat at 1005 power the operator would not be aware that a potentially unsafe condition exist When reactor power level is reduced by decreasing recirculation pump speed the pressure characteristics change across the jet pump dfffuser.
At minimum recirculation pump speed the measured pressure difference with LITS 3"52 equalizer valve cracked open indfcated Z psi with the higher pressure on the reference leg.
This delta presssure acts to reduce the reference leg level which results in the level instruments indicating a
higher than normal level. Therefore, as reactor level power was reduced to 545 by decreasing recirculation pump speed, the reference leg level decreased enough to make the reactor vessel level instruments (GEMACS) read a false 54 inches (the level for turbine trip).
The following reactor vessel level instruments share the reference leg with the shroud Yarway:
LT3-53 and LT3-206 (Feedwater Control),
LIS 3-184 (AOS interlock),
LIS 3."Z03A (RCIC), LIS. 3-Z08A (RCIC),
LIS 3"203A/8 (Low Level Scram)
and group 2,
3, and
PCIS.
The shroud Yarway supplies containment spray permissive signals to the containment spray logic.
Thirty-six hours prior to the scram, a
surveillance instruction was conducted on LITS 3-52 for calibration.
The licensee is evaluating the possible causes and corrective action to be taken with respects to this occurrence.
General Electric, Service Information letter (SIL) No.
335, da~ed July '1980, delineates. the-possible safety-related significance of leaving an equalizer valve on a differential pressure sensor partially open.
Plant Staff did not have a copy of SIL 335 prior to the scram and were not aware of its contents.
The inspector discussed with plant management the importance of having assurance that the equalizer valves are closed or not leaking through 'on the two instruments which measure water level in the shroud because of the unique situation where such a condition is not readily apparent when operating at full power recirculation flow rates.
The plant superintendent was aware that this was an undesirable situation and had instructed plant personnel to devise a method to lock these valves in the closed position.
This item wfl) remain.'open until licen'see reviews and corrective actions have been completed.
(260/81-09-01).
12. 'adiation Protection On March 12, 1981, the licensee informed the resident inspectors that hfgh radiation door 8178 located in the Rad Waste building was found unlocked and unattended by health physics personnel.
This was discovered during the periodic check..of high.podiotiqn doqr~ on Parch 11, 1981, at approximately
.1700.
jnyestjgat~on by ttje lfcenqee did not reveal when or who ]eft door 8178 unlocked.
An inspector determined that the control of high radiation areas (Areas greater than 1000 mrem/hr) were not being performed in accordance with Technical.Specification 6.3.D.Z The key to the high radiation doors is
being maintained by health physics and hea1th physic technicians may utilfze the key without any formal permission (i,e, checkout sheet)
from the shift engineer, The resident inspector identified to the Plant Superintendent on March 24, 1981, that failure to adequately control'he high radiation. door key was an apparent violation of 7echnical Specification 6.3.0.2 which requires the shift engineer on duty to administratively control the key.
The licensee committed to developing a
procedure for controlling the key.
(259/81-09-02).
On March 5, 1981, short-lived airborne radioactivity was detected by a
continuous air monitor (CAM's) in the turbine building.
Portions of the turbine building were evacuated and a total of 81 personnel received con-tamination on the body.
The contamination quickly decayed to nondetectable levels.
The principal nuclides were Rb"88 and Cs"138.
The airborne activity was legs than the maximum permissible concentration (MPC).
The maximum body contaimination was: received by. a pipe fitter which measured 250,000 OPM.
Whole body counts were made on three individuals, including the pipe fitter, with negative results.
TVA concluded there were no signi-ficant exposures as a result of thfs. occurrence.
The licensee informed the inspectors that the source of the activity was from a, small off-gas sample line in Unit 1.
A new piping "T" was being installed at the off-gas station without having the line isolated.
The licensee stated the work was performed without proper authorization or wor k instructions.
Corrective actfon was promptly initiated with personnel involved in 'that the individuals received additional instructions in this area.
The reportability requirement of this item was determined not to be required.
The inspectors had no further questions and', were sat'isfied with the corrective action taken by the licensee.
Maintenance The inspectors held discussions with maintenance personnel, engineers and plant management concerning the repair weld for the (EHC) electro-hydraulic system pressure transmitter in Unit 2 and the modification of the off gas sample line in Unit 3.
A review of the documents involved in the job was performed and discussions held at the job site.
No violations or devfations were identified fn the above area.