IR 05000277/1987025
| ML20236K213 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 11/04/1987 |
| From: | Linville J, Williams J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20236K203 | List: |
| References | |
| 50-277-87-25, 50-278-87-25, NUDOCS 8711090147 | |
| Download: ML20236K213 (32) | |
Text
{{#Wiki_filter:-___ _ - _ _ _ _. / l . . L l U. S. NUCLEAR REGULATORY COMMISSION
REGION I
l l Report No. 50-277/87-25 & 50'-278/87-25 ' Docket No. 50-277 & 50-278 l
License No. DPR-44 & DPR-56 Licensee: Philadelphia Electric Company 2301-Market Street Philadelphia, Pennsylvania 19101 Facility Name: Peach Bottom Atomic Power Station Units 2 and 3 l \\ Inspection At: Delta, Pennsylvania Inspection Conducted: August 29, 1987 - October 16, 1987 l
Inspectors: T. P. A hnson, Senior Resident Inspector i R. J. Urban, Resident Inspector l L. E. Myers, Resident Inspector I J. H. Willi s, Project Engineer-
Reviewed By: N/9' ff ' . H.4Kiliam, roject Engineer date ' , //[/[9 Approved By: _ [)/ ' lef, / 'date J );7CMinvi11e ()DivisionofR 'eactor Proje Section 2A, ctor Projects Inspection Summary: Routine, on-site regular and backshift resident inspection (237 hours Unit 2; 234 hours Unit 3) of accessible portions of Unit 2 and 3, operational safety, radiation protection, physical security, control room activities, licensee events, surveillance testing, refueling and outage activities, maintenance, and outstanding items.
Results: Four shutdown cooling isolations occurred (section 4.2).
An unresolved item (section 4.5.3) was identified involving the failure of a single heat sensor which could render a diesel generator inoperable.
The logic shown on the design drawings for this circuit was not consistent with I the as-built condition.
Response to a fire and cardox injection alarm in I diesel generator rooms was good. A review of the scram discharge volume capability determined that licensee implementation of Technical Specifications for a functional test is unresolved (section 5.12).
l l 8711090147 871105 l PDR ADOCK 05000277 i G PDR ' l _ _ _ _ _ _ _
_--- . _ __ _-- - - - - -. - _ --- _ _ - _ - - . . ., DETAILS 1.0 Persons Contacted ' B.- L. Clark,' Administrative Engineer G. F. Dawson, Maintenance Engineer A.A..Fulvio, Technical Engineer l . J. F. Mitman, Radwaste Engineer D. L. Oltmans, Senior Chemist F, W. Polaski, Operations Engineer D. P. Potocik, Senior Health Physicist G. R. Rainey, Superintendent Plant Services M. B. Ryan, Outage Engineer D. C. Smith, Superintendent Operations
- D. M.~ Smith,_ Manager, Peach Bottom Atomic Power Station
{ J. E. Winzenried, Staff Engineer Other licensee employees were also contacted.
- Present at exit interview on site and for summation of preliminary
findings.
2.0 Plant Status 2.1 Unit 2 .The unit remained in cold shutdown during the inspection period.
Refueling outage recovery efforts and reactor vessel hydrostatic testing preparations continued during the period.
, 2.2 Unit 3 , ) The unit remained in a cold condition during the period.
The pipe f replacement outage began on October 1, 1987.
By the end of the ) inspection period, the reactor vessel had been disassembled and - core offload activities were underway.
' 3.0 Previous Inspection Item Update
3.1 (Closed) Unresolved Item (277/85-08-05; 278/85-08-05).
FSAR l discrepancies and lack of safety or engineering reviews for the l safety relief valve (SRV) blowdown cycles exceeding the design l basis.
I Updated FSAR table 4.2.4 " Reactor Design Cycles (40-Year Life)" i ! contained discrepancies in the number of design cycles for the following: (1) the loss of feedwater heaters; (2) reactor overpressure with delayed scram (feedwater stays on and isolation
valves. stay open); (3) safety relief valve or code safety valve , l blowdown; and, (4) hydrostatic test at 1,563 psig.
FSAR Rev. 3, l - ! y o-
-- _- - _.
-_ . _ - _ _ _ _ _ __-- _ _ _ - _ _ _ _ - _ _ - p
... ',5 "
- .
Yi' \\ < [M; i Q-3-m q i i ' ' dated-January 1985 corrected these errors.
Revised FSAR. table ' 4.2.4 Land FSAR section C.5.3.1.1 " Vessel Fatigue-An'alysis" now: agree with the'originaltdesign; however, the.recent study,TSASR 85-54LRev. 1, dated January 20,'1986,Twhich analyzed additional; SRV blowdownsi, should~be incorporated'into the FSAR since the actual n_ umber of SRV:blowdowns exceeds, the design number of two: described in the.FSAR. The licensee' indicated that the FSAR would ,,, be: revised to includeLthe new fatigue assessment.
The inspector. reviewed the engineering analysis, !' Peach Bottom Units.2 and/3=Raactor Vessel Thermal Cycle Fatigue Assessment," SASR-84-54, Rev. 1 dated January 20,L1986' performed by General , Electric.
This analysis determined that an additional 114 SRV' .i blowdown cycles (new total of.116) could:be accommodated before reaching the ASME' Code' limit forLthe cumulative usage. factor of ' one.
A. Philadelphia Electric Company-letter'to NRC Region I dated July 5, 1985, acknowledged the lack of timeliness.in analyzing the i ' consequences of the. increased SRV blowdown cycles 'and revising the-number of. design-cycles. With the currently described. activity'to-monitor component fatigue parameters, this problem should be resolved.
'
The inspectors concern over the use of GE SIL No. 318 "BWR Reactor Vessel' Cyclic Duty Monitoring" for the number of design'SRVL . blowdown cycles' in ST 12.4 " Reactor-Pressure Vessel Transients '- . Cycles: Record" has been resolved. This number from the SIL, which . is not a Peach Bottom design number, is no longer used in ST 12.4.
Based upon.a" review of the FSAR, the GE Design Report, the licensee's response to the unresolved item, ST 12.4, and discussions with the licensee, the unresolved item is closed.. , 3.2 (Closed) Unresolved Item (277/85-08-06; 278/85-08-06).
Resolution of Technical Specification (TS) paragraph 6.10.2.f requirements and the data collected in ST 12.4 " Reactor Pressure Vessel . Transients - Cycles Record." Paragraph 6.10.2.f of the Peach . Bottom' Technical Specifications requires records to be kept of I operational cycles for components designed for a limited number of transients or cycles. A PECo letter dated July 5,1985, addressed this concern stating that ST 12.4 was not intended to meet the requirements of.TS 6,10.2.f.
It was stated that operating records are available that meet.the technical specification requirements.
~' The inspector will review transient history for selected components designed for a limited number of cycles at a future date.
The inspector reviewed ST 12.4 performed on June 23, 1987.
The licensee revised the analyzed number of SRV blowdowns to 116, based upon the GE fatigue assessment.
Since the number 116 is based upon a cumulatJve usage factor of 1, the number of cycles ,
,7.. .
j ' a
- l
y ' u _4_ b ~r' .t ' , e., .for other transients does'not have'the margincindicated in the -) ' , -FSAR (which.will-be revised) Land the ST 12.4: discussion.
It was', noted.thatLthel experience with several types of transients was-Approaching the design. cycles _ faster than expected based upon'a'4C year life. These observations were discussed with the' licensee.
JThe Linspector' was. satisfied with the licensee's' actions.which-l iricluded examining individual usage ' factors to accommodate actual-experience'and monitoring other components withtfatigu'e cycle design bases.
Based upon.this review,,the unresolved item.is closed.
l 13'.3.(Closed) Inspector Follow Item (277/86-12-06L 278/86-13-06).
Revise maintenance procedure M-65.4,~ " Hydraulic Snubber Testing.
.(With Load Cell).
The' licensee _ revised procedure M-65.4 to . address _a concern regarding the ~us_e of -the pressure values.for the - . test _ gauges installed on the snubber test machine.
These values were.10weredLto pFeyent possible snubber-overload.
The-inspector ] reviewed Revision.11.to procedure M-65.4 dated April 29, 1987,'and ! discussed this item with licensee engineers.
Based on the above, the follow item is closed.
D4.0 Plant Operations Review 4.1.1 Station ~ Tours , .The inspector observed plant operations-during daily facility tours. =Most accessible areas of the station were inspected.
- 4.1.1 Control Room and facility shift-staffing was frequently l'
checked for compliance with 10 CFR 50.54 and Technical Specifications. The presence of.~a senior licensed operator and the Nuclear Operations Monitoring Team member in the , control room was verified frequently.
4.1.2 The inspector. frequently observed that selected control .; room instrumentation confirmed that instruments were ' ' operable and indicated values were within Technical I Specification requirements and normal operating limits.
ECCS switch positioning and valve lineups were verified.
. based on control room indicators and~ plant observations.
Observations included flow setpoints,' breaker . l.
positioning, PCIS status, and radiation monitoring ' l' instruments.
4.1.3 Selected control room off-normal alarms (annunciators) , ' were discussed with control room operators and shift supervision to assure they were knowledgeable of alarm status and plant conditions, and that corrective action, , ' if required, was being taken.
In addition, the applicable alarm cards were checked for accuracy.
The operators were knowledgeable of alarm status and plant conditions.
I _- _ - - _ - - -
, m . -
y a .
M.. ' .n u ' 5- - , . . ] p .4.1.4 The. inspector checked for fluid-leaks by: observing sump' - l status,' alarms, and: pump-out rates;_'and discussed ! H ~ reactor coolant system leakage with licensee _ personnel.
j i '4.1.5 Shift relief and turnover activities were monitored daily, including backshift observations,.to ensure-m ~U compliance _with administrative procedures and regulatory ' guidance. -No-inadequacies were identified.
- i 4.1.6 The inspector, observed the main _ stack'and both reactorc
.i , ' building ventilation stack radiation monitors and-i - recorders ~,_and periodically reviewed traces from backshift periods to ve.rify that radioactive gas. release.
rates.'were within limits and that unplanned releases had' not occurred. No inadequacies were identified.
4.1.7 The' inspector observed control room indications-of fire j detection instrumentation and fire suppression systems, monitored use of fire watches and' ignition source controls, checked a sampling of fire barriers for-
integrity, and. observed fire-fighting equipment stations.
Except as noted in section 4.6, no inadequacies.were' identified.
l' 4.1.8 'The ' inspector observed. overall facility housekeeping _; conditions, including control of combustibles, loose ' trash and debris. Cleanup was spot-checked duringfand after maintenance.
The inspector made entries into selected locked high radiation areas to determine. conditions of'the areas. The' areas observed were in'the Unit 2. reactor building and included Reactor Water.- Cleanup _(RWCU) holdup pump rooms A and 8; RWCU pump A, B, and C rooms; RWCU regenerative heat-exchanger room; RWCU non-regenerative heat exchanger A and 8 rooms; RWCU valve rooms; RWCU backwash room; RWCU valve. pit and mezzanine area; north isolation valve room; outboard main steam isolation valve (MSIV) room; and TIP room.
The RWCU holdup pump rooms had been recently painted and cleaned, resulting in very low contamination levels.
The TIP Room is reasonably clean overall and has' low contamination levels.
The RWCU pump ~ rooms, valve rooms, backwash rooms, non-regenerative heat exchanger room, valve pit and mezzanine area and the north isolation valve room were highly contaminated, and required cleanup of trash and other equipment left in the areas. A lack of attention to cleanup and decontamination of equipment,- floors and walls after leaks and spills was also noted.
The RWCU regenerative heat exchange room is very highly contaminated and shows evidence of long standing water \\ , l' 2___
- ___-_~___-=:_-_ _ _ _ _
- - _ - _ - - . lN, -6- -. \\ i leaks. As a consequence, the area requires respiratory . protection to gain entry. The outboard MSIV Room had i-areas tnat were recently decontaminated; however, the areas underneath the grates were highly contamirjated. These (.t,z 4 items were discussed with licensee engineers and management.
The-licensee stated that programs underway are currently addressing these areas. The inspector will review the effect of cleanup programs for these areas in a future inspection.
The licensee has initiated a program to reduce the inventory of radioactive waste stored in drums and l boxes on site and in the plant.
The radioactive waste is
being temporarily stored in the Low Level Waste Storage Facility until compacted in overpacks, transported to a repackaging facility or transported to a disposal facility.
The program is projected to have all radioactive waste in drums or boxes stored on site removed by December 1, 1987, and is on schedule.
The inspector reviewed the program, discussed it with l licensee engineers, and inspected radioactive waste removal progress.
The inspector will continue to follow this program.
4.1.9 The inspector observed the shutdown nuclear instrumentation subsystems (source range and intermediate range monitors) and the reactor protection system to verify that the required channels were ' operable.
, 4.1.10 The inspector frequently verified that the required off site electrical power startup sources and emergency on site diesel generators were operable, i 4.1.11 The inspector monitored the frequency of in plant and
control room tours by plant and corporate management.
. The tours were generally adequate.
{ 4.1.12 The inspector verified operability of selected safety I related equipment and systems by in plant checks of j valve positioning, control of locked valves, power ) supply availability, operating procedures, plant
drawings, instrumentation and breaker positioning.
! Selected major components were visually inspected for { leakage, proper lubrication, cooling water supply, j operating air supply, and general conditions. No significant piping vibration was detected.
The
inspector reviewed selected blocking permits (tagouts)
for conformance with licensee procedures.
No inadequacies were identified.
l i , . _ _ _ _ _ _ _
'..g.
- Ik. i 'M' j ' , ' '7 i r .
.; 'l - 4 '.1.13 ' 'The' inspectors performed backshift and' weekend' tours of ~l ithe; facility on the following days: Friday,' October 2, 1987; 4:30 a.m.
.6:00 a.m.
-- Saturday, October 3, 1987;.8:00 a.m; - noon --- ! Friday,.0ctober 9, 1987; 4:00 a.m. -~6:00 a.m.
--- p 4.2 Followup On Events Occurring During the Inspection , ' a, 4.2.1-Unit 2 Shutdown Cooling Isolation on September 4, 1987 E
- The Unit 2 shutdown cooling system isolated at 3:05 a.m.
on September 4, 1987, while an engineer was researching a system blocking permit.in panel 20C32. The unit was in the shutdown mode with coolant temperature at.about:
H 150 degrees F.
The' engineer moved a loose wire at a ! ~ fuse ~ terminal block to read the label causing de-ener- 'l gization of the-75 psig isolation logic signal. =This resulted in the closure of the MO-17 and M0-18-valves', and j.
a in tripping of the 20 RHR pump.
The' licensee tightened the lead,' restored shutdown cooling at 3:14 a.m., and.made an ENS call at23:58 a.m.
! The inspector reviewed control room logs,-the upset , report,-the suspected'LER and the LER (see.section-
6.2.5).
The licensee determined the root cause of the isolation to be a loose connection at the. fuse terminal.
The: inspector examined the 20C32' panel. interior for other possible loose. connections. ~None were found.
The licensee intends to. inspect the interior of control cabinets for fuse condition, terminal-and screw: . condition, tight connections, and proper placement, and-j tagging.of jumpers and 11fted leads.
The inspector will
review this in a future inspection.
The. licensee also counselled the engineer with respect to attention to detail when working inside control cabinets.
The inspector had no further questions at this time.
No violations were noted.
l 4.2.2 Unit 2 Shutdown Cooling and Group II Containment Isolations-on September 16, 1987 , Primary containment group II inboard system isolations (PCIS) occurred at 8:15 a.m. on September 16, 1987, on Unit 2 while in cold shutdown at about 150 degrees F.
The shutdown cooling system isolated and the running RHR , , pump = tripped.
Other PCIS valves closed as designed.
The licensee reset the isolations and restored shutdown -; cooling at 8:25 a.m.
An ENS call for an unplanned ESF m_______ _ _ _ _ _ _ _. _ _ _
! . p
. f actuation was made at 10:45 a.m.
The cause of the-isolations was a momentary loss of power to the 20YO33 j 120 VAC panel which provides inboard logic power for the PCIS.
This power loss' occurred during electrical . ' switching required due to plant modification work.
The licensee's investigation determined that one of, the j temporary switches was inadvertently opened and this , resulted in a momentary loss of power to the 20Y033 120 VAC panel.
Thus, an isolation occurred.. The cause of . j the inadvertent opening was undetermined; however, a mislabelled switch may have been a contributing factor, y The inspector reviewed the licensee's upset report, { control room logs, the suspected LER, the LER, and d discussed the event with operations personnel.
The
inspector had no further questions at this time.
.: No violations were identified.
. 4.2.3 Unit 3 Shutdown Cooling and Group II Containment f Isolations on October 5, 1987 i y A group II' containment isolation and loss of. shutdown cooling occurred on Unit 3 at 8:26 p.m., on October 5, ) 1987.
The apparent cause of the isolation was a ' momentary loss of power to the 30YO33 120 VAC panel which supplies logic power. A test engineer was , ! troubleshooting a problem with the 3C high pressure service water pump breaker.
He inadvertently opened the
compartment door for the 4 KV bus potential transformer.
! This caused an undervoltage signal and the E334 load s center breaker tripped, causing loss of power of the d temporary feed to the 30Y33 containment isolation logic power.
The test engineer reclosed the compartment door and the E334 load center breaker reclosed in three ' seconds as designed. The temporary feed was in place because of an E13 bus outage due to maintenance.
The ,
licensee reset the isolation and restored shutdown ) cooling at 8:37 p.m.
An ENS call was made at 9:45 p.m.
A warning label on the bus potential transformer , compartment door was missing.
The label warned of power ] loss and breaker tripping if the compartment door was a opened.
Recent painting in the area had occurred, and apparently the warning label was removed from the panel.
I ' The inspector reviewed control room logs, the licensee's investigation and suspected LER, and discussed the event with operators and engineers. An LER will be submitted, and the inspector will review it in a future inspection.
j No violations were noted.
i
. _ - - _ _ _ _ - -
@ m ~ ' .Mi 4 %;" ,a w m , % o- ,, , ' '~j: .j[ 7 y( )t %l ^ .' [ ' 't n ' ,p ,,, i
, .. ,J' C %4 ' < a 'l < lQVf f [[, l
, vs .. 'a ,' - ,#~ -# .~ w.
y; y#a' T T,
' 9- - - n , , < e gt ,, ; . p,. , , E j,: ,;,' "/c, p, . ). ' $ [/
9' f j.~ , ., . ,
-r ' n.
'.
' , e "4(2; 4 :. Unit 3 ShutSwn Cooling and Group II Containment' ' L 'i ' !" 4l, f % . . . .. .. Isolations on 0ctober 12, 1987-E ~M ' , '.
- 0n-October 12,- 1987, at J12
- 00 : aim...a partial group II.
, a- , ' P containe ntoisolation occurredfon Unit 3 during.
L L - , W 'surveiTlince testing (ST).. ST-1.393, "PCIS - Logic. System. il ' , i Funstional Test" was being performed to verify standby. P,;r . 'y m ' . ' jgan' treatment system' operability.
This included.an L j .?; ' a .H ' 'unplarned 'shutdcwn cooling. (SDC) isolation' when.- the.SDC i . , L jnjection valve (MO-3.10-25B) cic. sed durir.g ST-1p3.
"'
h , performance by test engineers. An apparent inal.taquate
' .,4
- ,.
L3-temporary procedure change,(TPC) resulted in the valve 7 q': %* closure;' The, licensed operator' tripped th'e 30 RHR pump T -' upon. val.ve closure.
The vaine was reopened, 'the - 'f ' isolation was reset, and.' SIT Vas retprned to service a; % j
12:0} a.m' The licensee wh an ENS tall at.1:30 a.m / , Unit'3 D s in a pipe. replacement:rerueling outage. 3The4 ,.i ,
reactor vessel had'been disassembled, the fuel pool and ' equipsent pit gates NeFe ramoved, 'and itbe. cavities were .!q? , '
L4 ' i j ' l! %
' -flooded at the time p/f and was being maintained by the feelReact f,the isolation.
. as.abouts70 degrees ~ @} w pool cooling'and' SOC systems. The licensee determined that the TPC to ST 1.3-3 was inadequate.in that valve'M0-3-10-258 lfd ." f!
- was missed durtiig the TPClrevitw.
In' addition,.the valve" ' r 'is:not listed in procedure ST 1.3-3 and this apparently . ! - i
ccmounded the error; The insrpct;or reviewed ST-1.3-3, the.
TPC, the: licensee's. upset reportpand suspected LER, and ' discussed the event with operadorp and er.gineers.
The .
t inspector 9 ill review the-LER in atfuture inspection.
The inspector had no furthgr questions at'this time.
.t > 4.3 ' Logs and Recogds j t
, , ,[ .i The inspechor reviewed logs and records for accuracy, completeness, ' y' abnormal conditions, significant' operating changes and trends, required entries, equipment control and blocking status, jumper log validity, conformance with Lim 1 ting Conditions for Operattores, and proper reporting. The following logs and records were reviewed: / Nuclear Operations eionitoring Log, Shift Supervisipo Log, Reactor E,,{j Engineering Logs, Unit 2, Reactor Operator's Log, Uriit'3 Reactor >' Operator's Log, Control Operator Log Book and STA Log Book, Radiation f ' ,, Work Permits, Locked Valve Log, Maintenance Request Forms, Temporary .,. i
Circuit Modification Loghnd Ignition Source Control Checklists.
! ' f ', Control Room logs werd compared with Administrative Procedure A-7, t Shift Operations.
Frequent-in'itialing of entries by licensed operators, !
shift supervision, and licensee on site manage:nent constituted l avidence of licensee revik.
" >l[; s.
> , f '/ [ j
t , ,
,' , l
' [ . ' \\.' ' T 4 '! -- - - - - _ - - - - - - - - -
_ , . ,. -. - - j , ' l' ' - , -,
' U ,M ng [ ' 'n V4 ... q.
, ' .10- ' , , . ) $' '( , The licensee is using three licensed individuals from another BWR
ifacility to assist in the preparation of permits and blocking
. requests.
The inspector reviewdithe following administrative , ' ' procedures: A-26, '" Procedure for Corrective Maintenance," Rev.
' F5;'A-26A, " Procedure for Corrective and Preventive Maintenance < Uting' Champs," Rev. 5; and A;U, " Procedure.for Control of Safety-j,' ~ -Related Equipment." The insoector also reviewed selected permits M and discussed these items with licensed operators, permit writers and licensee managements The inspector questioned th6. extent of
i training given to the three individuals from the other facility.
The licensee stated that they successfully attended a three day , " Permit & Blocking"' class in August 1987.
The inspector verified ' thi.s'by reviewing.a memo dated September 29, 1987, and the class - attendance list. The inspector ",tated that procedures A-26,
A-26A, and A-41 are worded,such that there is confusion as to who , prepares the permit.
In some cases, it is the Chief Operator or . J) the Plant Operator - Nuclear or other qualified " Permit Freparer."
The ligepsue: agreed that there may be confusion, stated that the procedures are currently being revised, and indicated that this ~ item wculd M addressed.
The inspector will review these revised procedurret,Mn a; future inspection. The inspector had no further questions npt:cduferns, and no violations were noted, j 4.4 Enginee_redSafehuardsFeatures(ESF)AstemWalkdown ) i The inspector performed a detailed walkdown of portions of the l core spray system in order to independently verify the operability of the Unit 2 and 3 systems. The system walkdown included
EM verification of the'following items.
' y
Review of documents listed in Attachment 1.
-- ( Inspection of system equipment conditions.
-- ' Confirmation that the system check-off-list (COL) and ) a -- operating procedures are consistent witn plant drawings.
{ .3
Verification that system valves, breakers, arld switches are ! -- i propedy aligned.
I i Verification that instrumentation is properly valved in and !
-- operable.
Verification that valves required to be locked have -- appropriate locking devices , ( !-- ' Verification that control room switches, indications and l controls are satisfactory.
- A l
\\ ) ' , i
, i? ' . ' t
- (q
' ,q., ,, l ' , e ' ., . ! i-11_ . , .' ; . a ' ., , Verificationth$tsurveillancetestproceduresproperly.
. - - - implement the TechnictQ ypecificaticos surveillance /0 '
requirem9ats.
/> ' ' . ! No unacceptable 'conditf oris were identif hd.
l 4.5' Fire Prckection Systen Eyonts .H
< , , 4.5.1 i E-4 Diesel Gerierator (DG) Cardox Block"Valvt ) i On SepteAber 26, 1987, while applsing a ' system blocking ' permit nr. the lE-4 DG cardox systgrQ the licensee found l the electrormanual pilot cabinet (yMPG) block valve stem { to be.separ'ated.
The EMPC block ulve is a 3/4 inch ' Juiescury manufactured valve. Withthevalv/ operator I stemspparatedfromthevalvedisc,thevalvecouldnot . v be c ned. Jhfs prevented auto and manua? actuation of j ', the'esrdoxfiystem for the E-4 DG room. Goth Units 2 and i I 3 wer% in cold shutdown and the E-4 DG (and th_e other 3 DGs) were not required to be operable.
However, the ' licensee made an ENS call at 2:10 p.m., on September 26, 1987.
The licensee initiated a fire watch for the E-4 DG roo.n and repaired the block valve.. . ?,. The inspector walked down the cardox system in the DG rooms, revicsed P&ID M-318, and discussed this item with .. licensee operations and fire protection personnel.
The l inspector had'no further questions and no violations . were noted.
The licenser.r intends to ;ubmit an LER and the inspector will review it in a futdre inspection.
Within.the scope of this review, no inadequacies were ' identi fied. !I ' J 4.5.2 i E 2 Diesel Generator 1DG) Room Fire ) \\ At 10:55 a.ms, on October 7, 1987, a small fire occurred in the E-2 DG, room.
The DG was being run at the time ,, j for the weekly surveillance test. The cardox system was ' Ddisabled due to work and painting in the room.
A fire watch was posted.
Thesp'lant operator and the continuous ! fire watch extinguished %p fire in several minutes . s , using a. portable CO2 and absul ext:inguisher.
The DG was . tripped from the control room.
The. fire brigade was ' ' ( sent to the DG room; however, the. fire was out when they j ' arrived.
The apparent cause of the fire was a loose S, lube oil sensing line wnich sprayed oil on the "" ' exhaust manifold.
, t
i < , , _
l
> , , f
, < @, b : * " ' -12- < <g . J Mf The. inspector examined the DG and room with the system ! " V engineer, operator;, and a vendor representative on
October 7, 1987. No damage was noted in the area yhere l ti the fire occurred.
There was so,ae residue of the ansul , ' f//( j extinguishant.
The vendor representative was on site .
f 4 y' supervising the.E-1 OG annual _ over%ul and he inspected ' ' the E-2 DG.
m . The E-2 DG was started on October 8, 1987, )
and declared operable.
The, inspector.' observed this DG l
testing from the OG room.
In addition,.the inspector j interviewed the fire watch and operator who,were present
in the E-2 DG room when the fire occurred.. The ' . inspector determined that the overall. response to the 1, -, ' fire was timely and adequate.
No violations were noted, j D 4.5.3 E-4 Diesel Generator (DG) Room Cardox Alarm q On October 9, 1987, at approximately 7:30 a.m.. the j control room received a'cardox initiation alarm for th'e
.E-4 diesel generator. (DG) room.
Licensed operators in - the control room used the plant page system to warn > personnel to evacuate the E-4 OG room. Within five ! minutes, the. resident inspectors had arrived at the E-4 j j. ' DG room and the fire brigade was exiting wearing Scott ) 0-air packs.
They had just checked the entire DG building to make sure all personnel were accounted for and were safely
outside.
The evacuation alarm was audible, but the , '% inspectors were told that actual cardox i.njection had not. occurred.
Ten minutes earlier, an auxiliary operator had blocked the cardox injection path to the ) E-4 DG room because of cleaning and painting activities.
i Had it not been for the block, cardox would have l %, ' ' njected into the room.
i .The auxiliary operators, wearing Scott air packs, j entered the cardox room and determined that the tank was ,' , still full. Next, using an oxygen analyzer, the auxiliary operators toured the entire DG building and determined that a cardox injection had not occurred.
The alarm was reset and normal work activities resumed.
The inspectors determined that licensee response (security and fire brigade) was excellent and was very s- ' thorough.
s , Licensee investigation determined the cause of the cardox alarm to be a painter bumping his hard hat on one of the ) ,s . four heat sensors on the E-4 DG room ceiling.
The heat @q.
sensors are rapid rise (15 degrees F per minute) or.
, L6 sustained high (190 degrees F) temperature detectors.
j Bumping the detector could cause a pressure spike inside C the detector causing a false rapid rise temperature signal.
(.,- - i \\ \\, f; o h
, l .
_ _. , _ ._.
,' k , . j l-13- , i , The licensee examined the detector and fcund a dent in its shell;.they also bumped the detector again and were able to reproduce the alarm. The detector was replaced with a spare.
The inspectors examined electrical schematic diagrams i and datermined that two of the four heat sensors were needed for cardox. injection.
This fact was not in agreement with tiie event and surveillance tests that are performed on the heat sensors.
The inspectors accompanied the test engineers to the local cardox control cabinet. As found wiring indicated that a one-out-of-four logic was present rather than a two-out-of-four coincidence logic as depicted in . , elementary line and connecting diagram 6280-M46-24-18.
Since a cardox injection signal trips the diesel generatue (except during a LOCA MCA start), a single heat sensor failure or actuation cculd disable the , respective diesel generator.
This singl.e failure issue } for each OG, and the differences in logic for the as designed and a's built conditions are unresolved (277/87-05-01; 278/87-25-01).
5.0 IE Bulletin 80-14 and 17 Regarding Scram Discharge Volume Capability {MPA-B-58) e 5.1 Background On June 28, 1980, during a routine shutdown at a BWR facility, a manual scram from about 36% power failed to insert approximately 40% of the control rods.
The root cause was isolated to a problem with the scram discharge volume (SOV) system.
Followup inspections at other BWR facilities revealed a number of other deficiencies with the SDV system.
Multiplant Action (MPA) Item B-58, " Scram Discharge Volume Capability," was assigned to track , this issue.
C-orrective measures for MPA B-58 were divided into a short-term prcgram and a long-term program.
Short-term actions were implemented by IE Bulletins 80-14 and 80-17 (plus supplements).
Inspection reports (50-277/50-278) in which these bulletins were reviewed were: 80-24/18; 80-28/20; 80-29/21; 80-33/26; 80-35/28; 81-03/03; 81-09/10; 81-27/30, 81-XX/32; and 83-34/32.
Specifically, Bulletin 80-14 was closed in 81-27/30 and Bulletin l 80-17 was closed in 84-34/32.
The objective of the long-term L program was to improve the SDV design.
To verify satisfactory a completion of long-term licensee actions for MPA B-58, this review a was conducted in accordance with Temporary Instruction TI 2515/90.
l
, V
- _ _ - _ _ _ _
. -._ _ - _.
_ . . j' i l l . f l-14- . TI 2515/90 lists eleven criteria along with actions to be taken to ' audit compliance with the generic Safety Evaluation Report (SER), "BWR Scram Discharge System," dated December 1, 1980.
The eleven criteria selected cover objectives of the SER not yet adequately.
j addressed by Bulletins 80-14 and 80-17 (plus supplements).- The ( following eleven sections discuss each criterion and its disposition j in detail.
. 5.2 Scram Discharge Header Size (Criterion'D ] 5,2.1 Criterion < j i l The scram discharge headers shall be sized in.accordance l with GE Operating Experience Report (0ER)-54 and shall be hydraulically coupled to the instrumented volume (s) in a manner to permit operability of--the scram level-instrumen-tation before loss of a system function.
! 5.2.2 Disposition ' To meet the SDV header size criterion as stated in GE 0ER-54, a minimum size of 3.34 gallons per drive is required.
For Peach Bottom, the minimum header size equates to 618 gallons.
The' licensee's response to IE Bulletin 80-17 and 80-17 Supplement 1 (Gallagher, PEco to Grier, NRC Region I, 8/7/80) stated that the SDV f header volumes were 628 gallons'and 689 gallons for Units 2 and 3, respectively. These volumes do not include any volumes from the vent lines, drain lines, instrument volumes (IVs) or connecting piping.
. The inspector obtained isometric as-built drawings and.
walked the system down in both units.
The as-built conditions reflected the isometric drawings. The inspector performed an independent calculation of the SDV header volumes and found them to be sized as stated ! in the licensee's August 7, 1980 response.
The Unit 2 scram discharge system originally consisted of two 6" diameter SOV headers connected to a 12" diameter IV by 2" piping.
Unit 3 was similar, except that two 8" diameter SDV headers were installed.
The licensee modified both units' scram discharge systems by ' replacing the 2" piping with 6" and 8" piping for Units 2 and 3 respectively.
The connecting piping is now equal in cross sectional area to that used for the , existing SDV headers.
Therefore, proper hydraulic I coupling is ensured (drainage) and additional scram discharge system volume is gained.
) -
- _ _ ' . ' - ,, -15- , The inspector concluded that the SDV headers are adequately sized and are hydraulically coupled to the IV. Criterion 1 is closed.
..,. 5.3'AutomaticScramonHighSDVLevel(Criterion 21 5.3.1-Criterion.
i Level instrumentation shall be provided for automatic scram initiation while sufficient volume exists in the ' ' SDV.
,
h 5.3.2 . Disposition ' < The inspector reviewed F5AR Section 7.2, Technical Specification Table 3.1.1, and various electrical and . piping drawings.
There are four level switches that input to the reactor protection system (RPS) on each
!, , unit.
Unit 2 utilizes two Robertshaw float switches and ' two Magnetrol float switches.
Each of the two RPS channels uses one of each type for added diversification.
Unit 3 utilizes two Magnetrol float switches and two Fluid Components thermal dispersion level elements.
As in Unit 2, each RPS channel uses one of each type for diversification. When any one of.the two different level sensors in each RPb channel detects ' 50 gallons of water in the IV (one out 'of two twice logic) an automatic scram occars.
At this time, the IV - is nearly full, but as stated in detail 5.2.2 of this ! report, adequate scram vo',ume exists in the SDV headers alone.
Vent lines, drain lines, and the connecting piping are additional volumes.
The inspector concluded that an automatic scram occurs
while there is still sufficient volume remaining in the SDVs.
Criterion 2 is closed.
' 5.4 Instrument Taps Not On Connected Piping (Criterion 3) , 5.4.1 Criterion.
Instrumentation taps shall be provided on the vertical IV and not on the connected piping.
i
5.4.2 Disposition The inspector visually verified that safety related IV
level instrument taps were on the IVs and not on
connected piping above or below the IVs.
i l l l I t \\ __ -- _ _ _ _ _ - _ _ _ _ _. _ _ _ _ _ _ _ _. . -16-The inspector concluded that the as-built IVs fulfill the requirement in Criterion 3.
Criterion 3 is closed.
5.5 Detection of Water in the IV (Criterion 4] l 5.5.1 Criterion , l Scram instrumentation shall be capable of detecting water accumulation in the IVs assuming a single active failure in the instrumentation system or the plugging of an instrument line.
5.5.2 Disposition The physical arrangement of IV scram level instrumentation is slightly different at each unit. At Unit 2, level sensors 231 A and B (Al and 81 inputs to the RPS, respectively) are float level switches manufactured by Magnetrol.
These level sensors share a common upper instrumentation tap but have separate lower instrumer.t taps. The two remaining level sensors 231 C and 0, (A2 and 82 inputs to the RPS) are also float level switches, but are manufactured by Robertshaw.
These level sensors share a common upper instrument tap but have separate lower instrument taps.
At Unit 3, level sensors 231 A and B (Al and B1 inputs to the RPS, respectively) are float level switches manufactured by Magnetrol.
These level switches share a common upper instrument tap but have separate lower instrument taps.
The two remaining level sensors, 231 C and D (A2 and B2 inputs to the RPS, respectively) are thermal dispersion level elements manufactured by Fluid Components.
These level elements share a common upper and lower instrument tap.
For a single instrument tap plugging incident, the worst case scenario would be the loss of two of the four level sensors at each unit.
However, due to the arrangement of the logic, a scram could still occur since the two remaining level sensors are in different RPS scram channels.
For a common mode failure of one level sensor design, the worst case scenario would be the loss of two of the four level sensors. However, as in the prior case, due to the arrangement of the RPS scram logic, a scram could still occur with the two remaining level sensors.
In addition, the level sensors are fail safe such that component failure on a loss of power will result in a trip signal.
,. .. - - - - - - - - - - - - -. _ -. - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -.--- - --- . -17- , Power to the level sensors is from the RPS.
Each
channel of RPS (A and B) for each unit is powered from an independent power supply.
Each RPS channei receives its power from a different emergency load center.
Backup power is also available to the RPS channels, l either from a third emergency load center or from 250 l VDC. An interlock exists such that backup power cannot be fed to both RPS channels simultaneously.
If power is lost to any of the level sensors, they fail in the safe direction and insert a trip signal, The inspector also examined whether the level sensors on , - l each IV are diverse. According to TI 2515/90, diverse ! is defined as using different operating principles to measure water level in the IV.
Unit 3 uses two float level type switches and two-temperature sensitive probes to measure water level in the IV.
This arrangement clearly meets the diverse criterion as stated in TI 2515/90.
Unit 2 on the other hand uses two pairs of float level switches; each pair is made by a different manufacturer. This arrangement does not fit the diverse criterion as stated in TI 2515/90.
Upon further review, the inspector located Generic Letter (GL) 81-18, "BWR Scram Dis' charge System; Clarification of Diverse Instrumentation Requirement," dated March 30, 1981. GL 81-18 allows, as an alternative to using different operating principles to measure water level in the IV, substitute level sensors'of the same operating principle but made by a dif ferent manufacturer.
Therefore, the Unit 2 arrangement meets the diverse criterion.
The inspector concluded that the IV's scram instrumentation is diverse and redundant and that the IV's instrument taps are also redundant.
Criterion 4 is closed.
5.6 Vent and Orain Valves System Interfaces (Criterion 5) 5.6.1 Criterion Vent and drain functions shall not be adversely affected by other system interfaces.
The objective of this requirement is to preclude water backup in the scram IV, which could cause a spurious scram.
5.6.2 Disposition In an August 29, 1980, letter to B. H. Grier (NRC) from S. L. Daltroff (PECo), the licensee responded to question A.1 from I. E. Bulletin 80-17, Supplement 1.
s _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - - - - - - - - - - - - - - _
-_ ._.._ . -18- . This question concerned,.in part, the adequacy of the SDV vent and drain systems.
In this letter, the-licensee stated that the SDV vent and drain system design is satisfactory and no design deficiencies exist.
The licensee further stated that the vent system is not ! affected by any interconnecting systems, the drain system j is vented to the atmosphere, and the vent and drain' systems ' are not interconnected direr.tly and therefore do not interact.
In combined NRC inspection report 50-277/80-24; 50-278/80-18, NRC inspectors reviewed vent and drain path piping arrangements by utilizing system drawings ' and walking down piping in plant. No unacceptable conditions were identified.
In combined NRC inspection report 50-277/80-28; 50-278/80-20, NRC' inspectors reviewed vent and drain piping diagrams and inspected in plant piping configurations. No unacceptable conditions were identified.
Based on the August 29, 1980, submittal and on past NRC ! inspection reports, the inspector concluded that the SDV vent and drain functions would not be adversely affected by other system interfaces.
Criterion 5 is closed.
5.7 Vent and Drain Valves Close on Loss of Air (Criterion 6) 5.7.1 Criterion The power operated vent and drain valves shall close under loss of air and/or electric power.
Valve position indication shall be provided in the control room.
5.7.2 Disposition The arrangement of the SDV vent and drain valves on both units is similar.
There are a total of six valves per unit; two vent valves per SDV header (two SDV headers per unit) and two drain valves per IV (one IV per unit).
Electrical power to these six valves per unit is provided by RPS.
Channel A powers the three inboard solenoid valves (32A, 32B, 33) at each unit and channel B powers the three outboard solenoid valves (35A, 35B, 36) at each unit.
Using electrical schematic diagrams, the inspector verified that a loss of power to the solenoids would cause the vent and drain valves to ' close.
The inspector also examined GE prints and determined that all the valves were air-to-open, spring-to-close.
Therefore, on a loss of air, the l-valves will close.
The inspector also verified ,. . l l Q -_ __:_ __ O
_ _ __-____ __ . -19- , ' that there is position indication for all the vent and drain valves in 'the control room on both the COSA and C03 panels.
The inspector concluded that the SOV vent and drain valves on both units are fail safe (air / electrical) and ' position indication is available in the control room.
Criterion 6 is closed.
~ 5.8 Operator Aid (Criterion 7) 5.8.1 Criterion Instrumentation shall be provided to aid the operator in the detection of water accumulation in the IVs before scram initiation.
5.8.2 Disposition The inspector reviewed plant drawir.gs and visually verified that an alarm exists for each unit in the control room.
This alarm alerts the operator that water is present in the IVs.
Level switch 231F (Magnetrol float switch, both units) sounds an alarm in the control room when approximately five gallons of water is present in the IV.
The alarm response card informs the operator of what actions to take and also to enter procedure OT 105, " Scram Discharge Volume High Level," Rev. 1, dated 5/26/83.
This procedure instructs the operator to verify that all SDV vent and drain valves are open, and if the SDV cannot be drained, to begin a controlled plant shutdown.
The inspector concluded that alarms and procedures exist in the control room in case the SDV begins to unexpectedly fill with water.
Criterion 7 is closed.
5.9 Active Failure in Vent and Drain Lines (Criterion 8) 5.9.1 Criterion Vent and drain line valves shall be provided to contain the scram discharge water with a single active failure l and to minimize operational exposure.
5.9.2 Disposition i The inspector visually verified that there is double isolation on all the vent and drain lines for the SDVs at each unit. As described in detail 5.7.2 of this I
_ _ _ - - "r
-20-c.
report, each of tha two vent and drain valves on each vent and drain line are powered from a different RPS channel.
If a loss of. power occurred, the valves'would fail clos 3d.
Also, a loss of air would cause the vent and drain valves to close.
The inspector concluded that the vent and drain valves are, redundant, and that a single active failure will not j ' defeat isolation of the vent and drain valves.
' Criterion 8 is closed.
5.10 Periodic Testing of Vent and Drain Valves (Criterion 9) 5.10.1 Criterion-Vent-and drain valves shall be periodically tested.
5.10.2 Disposition The inspector examined surveillance test ST 9.22, " Scram Discharge Volume Drain and Vent Valve Stroking," Rev. 9, dated 7/19/84. The valves are stroked once per month and must close within fifteen seconds.
The inspector concluded that the vent and drain valves are periodically tested and timed.
Criterion 9 is closed.
5.11 Periodic Testing of Le' vel Detection Instrumentation (Criterion 10) l 5.11.1 Criterion Level detection instrumentation and verifying level detection instrumentation restoration shall be peridoically tested.
5.11.2 Disposition ' . The inspector examined the following surveillance tests to determine if they meet criterion 10: ST 9.27, " Flush, Calibrate & Functional Test SDV -- Level Switches," Rev. O, 12/31/85; ST 2.4.11 A; B; C; D, " Calibration Check of LS -- 2-3-231 A & F; B; C; D & E," Rev. 6, 8/5/86; Rev.
4, 12/9/85; Rev. 4, 12/10/85, Rev. 6, 8/5/86; ST 2.4.11 A/C/F; B/D/E, " Functional Test of LS -- 2-3-231 A/C/F; B/D/E," Rev. 10, 6/22/86; Rev 13, 7/31/86; and L __- -
_- _ __ __ . _ P q ! < , .. ' ' p.
' -21- . Similar surveillance tests for Unit 3.
-- The inspector concluded that surveillance tests exist.to ^ test SDV level alarms and trip instrumentation in' place.
These STs'also include steps.for restoration of system configurationby using double verification and . independent verification by a QC inspector. Criterion 10 is' closed, l.
5.12 Periodic Testing. Operability of the Entire System (Criterion 11) o 5.12.1 . Criterion j The' operability.of the entire system as an integrated
- whole shall be. demonstrated periodically and during each
' operating cycle by demonstrating' scram instrument response and valve. function at pressure and temperature at approximately. 50% control -rod density.
, '5.12.2 Disposition-Surveillance tests and Technical Specifications:do not-currently exist at Peach Bottom to fulfill Criterion 11.
This criterion is stated in the NRC's Generic Safety - Evaluation' Report on the BWR Scram Discharge System dated December 1, 1980, as required by-Confirmatory ' Orders issued to Peach Bottom.
Specifically, a full scram shall be initiated once per cycle from a normal control rod configuration of less than or equal-to 50% control rod density.
The licensee shall verify proper closing of the vent and drain valves within 15 seconds > ' , and proper reopening, after the scram reset.
Proper j r ,,' ' operation of other system components and instrumentation ' shall be verified. Also, after each scram from a pressurized condition, or at least every 31 days,-proper float and level sensor response shall be verified by a performance of a' channel: functional test of the SDV i scram and control' rod block level instrumentation.
' The licensee includes some of the above requirements in general plant procedures. GP-18, " Scram Review Procedure," Rev. 5, dated 2/12/87, checks that the SDV -- isolated, but doesn't check the closing times of the vent and drain valves and does not check their reopening after j scram reset. GP-18 also checks for a SDV high level trip from all four channels. GP-2A COL, " Reactor Startup Order," Rev. 70, dated 11/26/86, states that it is a good idea, but not a requirement to perform a functional test of the SDV instrumentation after every scram.
However, the licensee does perform a channel functional test of the < _. _ _ _ _ _ _ _ _ _ _ _ _. _ -. -
__ _-_;-. ._ r e !r .E , -22- , , , p L t.
I-SDV float and level' sensors by.using ST 2.4.11 A/C/F & B/D/E and ST 2.9.11 A/C/F & B/0/E, for. Unit. 2 and 3, respectively.
, J l"_ TheLinspector contacted several oth'er.BWR facilities in j L ~ Region I to. determine' how those licensees ' met criterion 11. LThe. inspector determined that eachifacility was L different-Contact was made with the NRR technical . L reviewer who has been assigned to develop a standard. set l' of requirements for all BWR' facilities to fulfill-
. criterion 11~. Until some standard. requirements are . released by NRR, criterion:11 will' remain open and , . unresolved (277/87-25-02; 278/87-28-02).
.' ' ? 6.0 Review of Licensee Event Reports (LERs)- 6~.1 LER Review v , The inspector reviewed LERsisubmitted to'the NRC to verify that- -the details were clearly reported; including the' accuracy of the.
description ~and corrective action adequacy.
The inspector determined whether further information.was required,.whether . generic implications were indicated, and whether the event warranted on-site followup.
The following LERs were reviewed: < LER No.
LER Date Event Date Subject ' 2-87-13 Unit 2 Group II B Containment Isolations
September'21, 1987 due to blown fuse ' August 20,'1987- -
- 2-87-15 Unit 2 and 3 containment isolations due to
5eptember 15, 1987 partial loss of off site power LAugust.15,-1987
- 2-87-16 Unit 2 and 3 containment isolations due to a
' ' September.21, 1987 partial loss of off site power August ~20, 1987
- 2-87-18 Unit 2 shutdown cooling isolation due to September 30, 1987 inadequate blocking August 28, 1987
!
- 2-87-19-Unit 2 shutdown cooling isolation due to October 5, 1987 loose electrical connection
" September 4, 1987 ,u e-A _ _ _ _ _ _. _. - _. _ _. _ _ _ - - - _ _ _ _ - - -
___ I I . l-23-r ,
- 2-87-21-Unit 2 shutdown cooling isolation due to loss October 16, 1987 of power to 20Y033 panel September 16, 1987 3-87-7 HPCI inoperability due to failure of September 28, 1987 power supply August 29, 1987 6.2 LER On-Site Followup For LERs selected for on site followup and review (denoted by
, asterisks above), the inspector verified that appropriate i corrective action was taken or responsibility assigned and that continued operation of the facility was conducted in accordance with Technical Specifications and aid not constitute an unreviewed safety question as defined in 10 CFR 50.59.
Report accuracy, compliance with current reporting requirements and applicability to other site systems and components were also reviewed.
6.2.1 LER 2-87-13 concerns a group IIB containment isolation (shutdown cooling) on Unit 2 on August 20, 1987.
The event was reviewed during NRC Inspection'277/87-22 and 278/87-22.
No inadequacies were noted relative to this LER.
6.2.2 LER 2-87-15 concerns a partial loss of off site power on August 16, 1987, resulting in Unit 2 and 3 containment , isolations. This event was reviewed during NRC Inspection 277/87-22 and 278/87-22.
No inadequacies were noted relative to this LER.
E 6.2.3 LER 2-87-16 concerns a partial ross of off site power on August 20, 1987, resulting in Unit 2 and 3 containment ~ isolations.
This event was reviewed in NRC Inspection 277/87-22 and 278/87-22. The cause of this event was an 80-ton mobile crane drawing an are through the boom to the ground while attempting to move a cargo unit located under the No. 2 off site power source 220 KV lines.
Corrective actions have been completed to prevent recurence which included training of maintenance foremen and riggers in electrical safety and placement of warning bars or trapeze from high voltage lines where appropiate.
No inadequacies were noted relative to this LER.
' 6.2.4 LER 2-87-18 concerns a Unit 2 containment isolation resulting in a shutdown cooling isolation on August 28, 1987.
This event was reviewed in NRC Inspection 277/87-22 and 278/87-22.
An unresolved item was identified regarding the reapplication of a permit _ _ _ _ _ _ - _ _ _ _ _ -
Im I-H ' . L-24- ! . during changing plant conditions.
The LER committed to reviewing this item by January 1988.
No inadequacies were identified relative to this LER.
6.2.5 LER 2-87-19 concerns a shutdown cooling isolation on Unit 2 on September 4,-1987.
The event was reviewed in section 4.2.1 of this report. No inadequacies were noted relative to this LER.
6.2.6 LER 2-87-21 concerns a shutdown cooling and containment i isolations on Unit 2 on September 16, 1987. The event.
' was reviewed in section 4.2.2 of this report.
No inadequacies were acted relative to this LER.
7.0 Surveillance Testing The inspector observed surveillance tests to verify that testing had been properly scheduled,' approved by shift supervision, control room operators were knowledgeable regarding testing in progress, approved procedures were being used, redundant systems or components were available for service as required, test instrumentation was calibrated, work was performed by qualified personnel, and test acceptance criteria
were met.
Parts of.the following tests were observed: -- ST 2.9.11 B/D/E, " Functional Test of LS 3-3-231 8/0/E," Rev. 13, performed on Unit 3 on October 3, 1987.
ST 2.9.11 A/C/F, " Functional test of LS 3-3-231 A/C/F," Rev.12, -- performed on Unit 3 on October 3, 1987.
ST 8.1, " Diesel Generator Full Load Test," Rev. 24, performed on -- the E-2 DG on October 8, 1987.
ST 9.32-2,-2, 3; " Reactor Cold Shutdown Data Log," performed -- hourly on both Units 2 and 3 during the inspection period.
-- ST 3.1.1, "SRM Functioaal & Calibration Adjusteent Procedure," Rev. 10, performed on Unit 3 on October 15, 1987.
In addition, a review of the following completed surveillance tests was performed.
-- ST 1.3-3, "PCIS Logic System Functional Test," Rev. 7, performed on Unit 3 on October 12, 1987.
ST 3.1.2, "SRM Core Monitoring Test," Rev. 10, performed on Unit 3 -- on October 15, 1987.
No inadequacies were identified.
I E_ .
__ __ . . _ - _ _ - _ - , _ _ _ _. _ __ . - _ - _ - _ _ - _ - _ - - _ _ _ - - - - - _ _ -
- ^
' ,
2 - 'f g - r ru -, t - s G ' '
- p.
m '25- . ; - . ... . .. . 8.0 maintenance .. For the following maintenance activities lthe inspector spot-checked .J l administrative controls,1 reviewed documentation,.and observed' portions H 4., o. of the" actual maintenance: . . Maintenance r ' Procedure /f l%,'
- Document
, Equipment' Date(s): Observed
- SP-1009
' Scram Discharge ~ Header.
. September 2'3-30, -1987 ~ Decontamination _.
, _
- M52.2. E-1 DG. Annual Inspection.
October 5-8, 1987: ' M00-664 Containment Purge / Vent September 3, 1987 , ' Radiation. Monitors ,, i Administrative controls checked included maintenance request forms .'(MRFs), blocking. permits, fire watches and-ignition' source controls,
' ' item. handling; reports,-QCLinvolvement', plant conditions,.TS LCOs, ! - t . equipment turnover information, and, post maintenance testing.
' Documents reviewed included maintenance procedures,. material certifications,-RWPs, MRFs, and receipt inspections.
a
No; inadequacies were identified.
9.0 Radiation Protection" , 9.1 Routine Observations , . . ) During the. report period, the inspector ' examined work in: progress in both units,' including health physics-(HP) procedures and
controls, dosimetry.and badging, protective clothing use, adherence to radiation work permit (RWP) requirements, radiation .j' surveys, radiatien protection instruments use, and handling of _ potentially contaminated equipment and materials.
, ' . . ? = The inspector observed individuals frisking in accordance'with HP a procadures. A sampling of high' radiation doors was verified to be .. locked as required.
Compliance with RWP requirements was verified during each tour.
RWP line entries were reviewed to verify that personnel had provided the required information and people working in.RWP-areas were observed to be meeting the applicable , requirements.
No unacceptable conditions were identified, j y '9.2 Locked High Radiation' Area Controls ' As a result of numerous occasions when locked high radiation areas were found unlocked during surveillance, the licensee has i significantly changed the methods of control of these areas.
] i ^ ^ .. ... -- - - - -
_ _ - _ < l - .c L ' -26-L.
Licensee investigation had revealed that the failure to lock the , entrances to these areas were primarily due to personnel error; that is, individuals forgot to lock the areas when leaving the area.
However contributing factors were the lack of strict key control by Applied Health Physics and deficiencies in controlling procedure HP0/CO-145, " Locked High Radiation Area Access Control," to provide accountability and oversight. As a consequence, the method of controlling locked high radiation areas was reviewed by the licensee.
The result is a new procedure, HP-109, " Locked High Radiation Area Access Control," which superseded HP0/C0-145 on August 28, 1987.
) The new procedure initiates strict key control and specifies two levels of control depending on the radiation levels.
All locks of locked high radiation areas have been re-cored to one master.
Level I control is specified for areas where there exists radiation at a level so that a major portion of the whole body- , could receive in any one hour a dose in excess of 1.0 rem but less than 30 rem.
Level II control is specified for areas with dose rates in excess of 30 rems in one hour, or as designated by the Applied Health Physicist for potentially extremely high exposure rates.
Master keys have been issued to the Shift Superintendent and Applied Health Physics supervisor.
These keys are maintained in locked key lockers and are to be used only in cases of emergency. A complete set of individual keys to the areas are maintained under the control of the Applied Health Physics Supervisor in a locked key locker.
The individual keys are further segregated into Level I and II keys.
Level I key issuance is under the control and approval of the Senior HP Technician, Level II keys are under the control of the Senior HP Technician but are issued only to qualified Health Physics personnel after the approval of the Applied Health Physics Supervisor and the Shift Superintendent.
Accountability is provided by requiring each individual issued a key to sign a statement understanding the responsibilities of
entry to locked high rat::* ion areas, and upon return of the key, ~ a statement that the area was locked on exit. Oversight is provided by sending a health physics technician within one hour of _ key return to ensure that the area is locked.
Since the procedural change, there haven't been any instances of locked high radiation areas left unlocked.
The-inspector reviewed HP-109, the locked high radiation door program, and discussed this with licensee personnel.
The inspector also reviewed the licensee's health physics deficiency reports for - these recent problems. A survey of locked doors was also performed-. No unauthorized open doors were found.
Other than the licensee l identified violations of locked high radiation doors, no unacceptable conditions were noted.
The inspector will continue to follow this item.
_ _ _
_ _ - _. l- ' ({. i , Jt.
( l ' -27- , i Within the scope of'this review, no violations were identified.
d 9.3. Unit 3 Pioe Replacement Controls On October 7, 1987, representatives of the Pipe Replacement Project Group (PRPG) met in Region I to explain the organization , of.the Radiation Protection Group supporting the outage work on j Unit 3 for the pipe replacement.
The Radiation Protection Group (RPG) is supervised by a qualified radiation protection manager who is responsible to the PECo Senior Health Physicist.
The RPG J l is structured in. parallel with the plant-health physics department I and will have an Applied Health Physics Section, Technical Support - Section and an ALARA Section.
The RPG will be responsible for specified areas in the Unit 3 Reactor Building which will include the entire drywell, all of 135 foot elevation,165 foot elevation and portions of the 195. foot elevation.
In addition RPG will-maintain an access to these areas through a specially constructed access building, the Containment Access Building. All entries for PRPG will be through this access building.
The group estimates the. dose will be 1725 Rem for the PRPG outage.
10.0 Physical _ Security 10.1 Routine Observations The inspector monitored security activities for compliance with the accepted Security Plan and associated implementing procedures, including: operations of the CAS and SAS, checks of vehicles on site to verify proper control, observation of protected area h access control and badging procedures on each shift, inspection of physical barriers, checks on control of vital area access and escort procedures.
No inadequacies were identified.
10.2 _ Inattentive Unit 2 Orywell Watchperson The PECo shift security assistant and the corporal of the guards found a security watchperson inattentive at the Unit 2 drywell equipment access at 1:50 p.m., on September 6, 1987.
The watch-person was observed with her feet up and head down.
She responded _ immediately to questioning, and she had made a radio check one minute earlier (1:49 p.m.) with the SAS operator.
The watchperson was suspended pending an investigation and was escorted off site.
No work was in progress 'in the drywell and a HP control point was not established. An inspection of the drywell was performed and no abnormalities were observed.
The licensee informed the senior resident inspector and made an ENS call at 5:05 p.m.
An inves-tigation of the incident was performed. The watchperson subsequently quit the security force.
- _ - - _ _ _ - _ - . _ _ _
___ - , , . - - i "
- l
, , . - l' ! . -28-l i , < l . ,. ' .. l ' The inspector reviewed the licensee's investigation including statements, computer printouts, sequence of events and post l, instructions. The inspector discussed the event with licensee , . security management. This is the third instance of an inattentive
drywell watchperson. The other two occurrences were on June 21, 1987 l <and May 29, 1987. Unresolved item 277/87-15-02' remains open.
The { ' inspector had no further questions at this time.
j 10.3 Fitness for Duty Prograin Implementation On September 10,.1987, the licensee informed the senior resident inspector that access to. the protected area for three PECo - 1 . maintenance.. department personnel was suspended effective September 4, 1987. This action was based on suspected use and/or dealing of controlled substances. The licensee is evaluating the individuals
involved in accordance with the PECo fitness for duty program.
On September 30, 1987, a contractor employee tried to enter the protected area while intoxicated.
Licensee security guards administered sobriety-tests which the individual failed.
His superviscr was notified and his security badge was pulled to deny protected area access.
The contractor eventually. terminated the individual's employment.
The inspector reviewed the licensee's investigation in each of these two instances.
01scussions were held with licensee management.
The inspector had no further questions at this time.
The inspector will continue to follow the licensee's Fitness for Duty Program.
10.4 Bomb Threat on September 20, 1987 The control room shift clerk received two anonymous telephone bomb threats on Sunday, September 20, 1987.
The calls were received at 5:20 and 5:30 p.m. from a male caller who whispered: "there's a bomb".
No specific mention of Peach Bottom nor time / location was included in the bomb threat.
The licensee implemented the security procedure for a bomb threat.
This included additional searches of personnel entering the protected area and a search of plant buildings by security and operations personnel.
Nothing abnormal was noted during the searches.
The licensee evaluated the threat as having no supporting evidence and not to be an explicit threat.
No formal NRC notifications were made; however, the anfor resident inspector was called at home by plant management.
The inspector reviewed the licensee's bomb threat procedures and interviewed the shift clerk who received the call.
In addition, a review of the licensee's investigation was performed. No violations were noted.
. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . fb y; q ... -29- ' . -0 10;5 Drugs Found On Site l / c L On October.4,.1987, a contractor employee..found a small plastic bag containing a white powdery substance in.the parking lot' outside of the protected ~ area.
He turned the bag over to Burns ' Security lwho' then. turned it over to PECo Claims Security fors analysis. Five hours 11ater a Burns Security guard found a small pile of white powder on'the floor.in the men's' room.inside the guard house..This: area is outside.the protected' area but inside
-the protected area search equipment. -The pile.was scooped up and ' turned over to PECo Claims' Security for analysis; Licensee preliminary testing..was performed on both substances the morning ... of October 5,1987;. each tested positive. as cocaine.
The licensee - " , notified the Pennsylvania State Police and tran' sported the illegal-drug to the state laboratory for more accurate analysis and i - . - ' disposal.
bl The~1'spector: held discussions wit'h' security management and n ,
' reviewed their actions, No violations were noted.
l ' 10.6' Drug' Allegation (RI-87-A-0070) In combined inspection report 50-277/87-17; 50-278/87-17, details-of;an. allegation were discussed in section 10.3.
The allegation j co'cerned;an a'nonymous phone caller who stated he knew of a
n " contractor employee. who was a habitual drug. user. - As a result, j the contractor employee had his security badge pulled pending further " .,. licensee investigation.
' Lower level mana'gement in Nuclear Security decided that the' H individual could regain his security badge if he passed a drug test ~and a psychological evaluation. The drug test was; passed as
- discussed in combined. inspection report 87-17
.A psychological . evaluation was also passed; however..the psychologist was not . familiar with the case, and only a standard evaluation was given.
, , The contractor was allowed access to the protected area under the condition of taking random drug tests for an indefinite time l period.
The on site PECo Claims Security. Inspector saw the' individual J inside the protected area and was not aware that he had regained
access;-upper level security management was'also unaware of this occurrence, Due to the manner in which this situation was handled, a.new PECo policy was instituted. Any time' a person is denied . protected area access, the manager of PECo Claims Security must approve renewed access.
q q i Based on: discussions the inspector had with the PECo Claims ] Security Inspector; the individual passing random drug tests and a i psychological evaluation; and a commitment to random drug test the l I i i l ) - _ _ _ - _ _ _ _ _ _ - __ _ _ _ _
_ - - _ _ _ _ _ _ _ _ _ _ _ _. - _ _ _ _ _ _ _ __ .__ . _ _ _ - _ _ _ _ _ - _ _ _ _ _ . -30-Q individual for an indefinite period; this allegation is considered resolved. However, if the individual tests positive for drug use in the future, PECo will take appropriate action in accordance _ with their Fitness for Duty Program.
11.0 In-Office Review of Public and Special Reports The inspector reviewed the following: Peach Bottom Atomic Power Station Revised Monthly Operating Report -- for August 1987, dated October 2, 1987.
! No unacceptable conditions were noted.
i 12.0 Unresolved Items Unresolved items are items about which more information is required to . ascertain whether.they are acceptable violations or deviations.
Unresolved items are discussed-in section'4.5.3 and 5.12.2.
L13.0 Management Meetings 13.1 Preliminary Inspection Findings A verbal summary of preliminary findings was provided to the Manager, Peach Bottom Station at the conclusion of the inspection.
During the inspection, licensee management was periodically notified' verbally of the preliminary findings by the resident
- inspectors.
No written inspection material was provided to the-licensee during the inspection. No proprietary information is included in this' report.
13.2 Attendance at Management Meetings Conducted by Region Based a Inspectors t' Inspection Reporting Date Subject Report No.
Inspector Sep 21-25, 1987 Operator AttituJe None Linville Training Sep 28-Oct 2, 1987 Regulatory ilone Pickett Effectiveness Revieu Oct 5-7, 1987 Operator License 87-27/27 Howe ! Exams l
} I l ! ! ! I i ___ _ _ _.. _ _ _ _ _. _ _ _ _ !
, _ __ - . _ _ _ _ _. _ _ _ _ _ _ . -31-i o 13.3 Attendance at Other Management Meetings 'Date Subject Sep 9, 1987 Unit 3 Pipe Replacement Sep 14, 1987 NRC Commission Briefing on Peach Bottom Sep 24, 1987 Public Meetings in Harford County, MD and York County, PA Oct 1, 1987 SALP Management Meeting ' Oct 7, 1987 Radiological Controls for Unit 3 Pipe Replacement Outage.
i , . _ _ _. _ _ _ _ _ _ _ - -. - - _ _
_ - _ _ p j ! ! !+ I ! .o , ATTACHMENT 1 Documents Reviewed for Core Spray ESF Walkdown . ' 1.
Peach Bottom Atomic Power Station, Units 2 & 3, Technical Specifications
2.
PBAPS, Units 2 & 3, Updated Final Safety Analysis Report 3.
System Procedures S.3.4.A, " Setting Up Core Spray for Auto Operation",- I Revision 6, dated 7/24/85 ' ! 4.
System Procedure, S.3.4.E, " Normal System Setup for Automatic ' t Operation",. Revision 2, dated 4/30/85 5.
System Procedure, S.3.4.El COL (Units 2 & 3), " Core Spray System Normal Line-Up (Control Room Check), Revision 2, dated 10/22/86 , 6.
System Procedure, S.3.4.E.2 COL (Units 2 & 3), " Core Spray System Normal Valve Line-Up (Outside of Control Room) Units 2 & 3, Revision 6, dated 6/24/87 7.
System Procedures, S.3.4.F, Filling and Venting the Core Spray Loop Discharge Piping," Revision 2, dated 6/1/77 8.
P&ID M-362, Sheets 1 & 2, " Core Spray Cooling System" _ _ _...-.. ___-___m_-
----_.-----_--_ - - - - ---
}}