IR 05000277/1998006
| ML20236R766 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 07/16/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20236R728 | List: |
| References | |
| 50-277-98-06, 50-277-98-6, 50-278-98-06, 50-278-98-6, NUDOCS 9807220262 | |
| Download: ML20236R766 (66) | |
Text
_ _ _ _ _ _ _ - _ _ _ _ _ _ - _ _ _ _ _ _ _ _ - _ - _ _ _ _ _ _ _ _ - _ _
.
..
U. S. NUCLEAR REGULATORY COMMISSION REGION 1 License Nos.
DPR-44 -
DPR-56 Report Nos.
98-06 98-06 Docket Nos.
50-277 50-278 Licensee:
PECO Energy Company Correspondence Control Desk
.. P.O. Box 195 Wayne, PA 19087-0195 Facility:
Peach Bottom Atomic Power Station Units 2 and 3 Inspection Period:
May 5,1998 through June 22,1998 Ir.spectors:
A. McMurtray, Senior Resident inspector M. Buckley, Resident insp9ctor
'
B. Welling, Resident inspector Approved by:
Clifford'J. Anderson, Chief Projects Branch 4 Division of Reactor Projects
>
9907220262 980716 PDR ADOCK 05000277
,
e pan
'
I
!
_ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ - _
.
.
EXECUTIVE SUMMARY Peach Bottom Atomic Power Station NRC Inspection Report 50-277/98-06,50-278/98-06 This inspection report included aspects of licensee operationa surveillance and maintenance; engineering and technical support; and plant support areas. The report covers a seven-v eek period of resident inspection.
Onerations:
e During a Unit 2 downpower evolution on Mav 16,1998,aperators reduced speed on an incorrect reactor feed pump, resulting in a reactor level excursion and recirculation system runback. This event was indicative of poor operator performance, reflecting weaknesses in communications, self-checking, and peer / supervisory review. Following the evem, the inspectors observed increased peer checking and improved oversight by control room supervisors. (Section 01.2)
The 3A stator water cooling pump tripped during system troubleshooting efforts e
on April 28,1998, due to weaknesses both in operations review of the work and with communications regarding restrictions on the work scope. Operations personnel performed a good investigation of this issue and initiated appropriate corrective actions. (Section 01.3)
On May 15,1998, operations personnel identified that the trip relay for the Main Control Room Emergency Ventilation (MCREV) radiation monitor had not been in the tripped status for approximately 28 hours3.240741e-4 days <br />0.00778 hours <br />4.62963e-5 weeks <br />1.0654e-5 months <br /> while the 'B' channel radiation monitor was inoperable. This condition resulted in a violation of technical specification 3.3.7.1 since the 'B' channel was required to be tripped within six hours after the charinel became inoperable.
The operations personnel installing the jumper to initiate a Division 11 isolation trip of the MCREV radiation monitor did not perform, nor did the procedure instruction require, a positive verification that the trip was properly inserted. The corrective actions from the July 10,1997 event were not comprehensive enough to prevent this subsequent event. (Section O2.1)
NRC review of a number of plant status controlissues related to mis-positioned valves in safety related systems identified a common element of inconsistency between plant procedures, check-off lists, drawings, administrative guidance, and/or operations and maintenance practices, in some instances, plant personnel missed opportunities to identify the discrepancies.
I l
In one instance, engineering personnel failed to take prompt and effective corrective actions followina OA's identification of errors affecting several procedures for the residual teat removal system. Operators used one of the l
)
l ii
!
!
.
j
_ _ - -
._. _ - _ _ _ _ - _ _
s
.
Executive Summary (cont'd)
!
incorrect procedures approximately five months after QA's findings, and as a result, one valve was left out of its required position. Although system W.
. ~ operability was not affected, an engineering evaluation was necessary to reach that conclusion. This was considered a violation of 10 CFR f>0, Appendix B, Criterion XVI, Corrective Action. (Section O3.1)
On June 7,1998, the 3A recirculation pump ran back to 30% speed due to the e
unexpected loss of a 500 kV line during an electrical storm and the slow opening of 500 kV hreaker. The 38 recirculation pump remained at full speed during this
!
event. Due to the differences in pump speeds of the Unit 3 pumps, the flows in the recirculation loops were significantly mismatched. The recirculation loop flows remained mismatched outside of Technical Specification Surveillance Requirement (SR) 3.4.1.1 for over 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. This mismatched flow resulted in a violation of Technical Specification 3.4.1.
Operations and engineering personnel did not understand the e,ffects of the recirculation flows mismatch on the jet pumps and jet pump riser braces due to the excessive vibration stresses. They also failed to recognize that Unit 3 was in
. single loop operation following the runback of the 3A recirculation pump.
In addition, no information was documented in the general procedure used during this event that would alert the 9perators to the need to balance recirculation l
flows quickly to prevent high vibration stresses in the jet pump loops. Also, the abnormal procedure for single loop operation did not contain any information
. regarding the vibration concerns or technical specification bases information about being in single loop operation when the mismatch between the two recirculation loops was greater than required limits. (Section 04.1)
-o On March 23,1998, the licensee identified that they failed to properly implement i
the improved Tecnnical Specification Surveillance Requirement 3.4.9.4 for the start of the first recirculation pump. Between January 18,1996, and March 23, 1998, operations personnel were not verifying that the temperature differential between the reactor coolant in the recirculation loop being started and the reactor pressure vessel coolant was within 50*F. On October 29,1997, the 'B'
recirculation pump was started with a differential temperature of 84 'F.
Although this did not exceed design limits nor impact fuel performance, it was a violation of Technical Specification Surveillance Requirement 3.4.9.4. (Section 08.1)
Maintenance:
e Unit 2 on-line scram solenoid pilot valve replacement activities, from June 5 through June 10,1998, were particularly well-executed. Nuclear Maintenance
. Division. technicians and operations personnel displayed good procedure usage
and sound work practices. Supervisory personnel provided very good l
. coordination and oversight. (Section M1.2)
I lii
,
-
.
L
I
.-
,
Execut ie Summary (cont'd)
Enaineerina:
-e-
. Engineering personnel failed to recognize the potential for high vibration stresses on the 'A' jet pump loops due to the large recirculation flow mismatch following the 3A recirculation pump runback on June 7,1998. The potential for recirculation flow mismatch to cause excessive vibration of the jet pumps and the jet pump riser braces was described in the Peach Bottom Design Basis
!
Document (DBD) for the recirculation system. This lack of understanding of the j
effects of this mismatch contributed to the failure of engineering personnel to '
provide the necessary technical information to operations personnel. This resulted in the failure to recognize the need to expeditiously resolve the mismatch between recirculation loop flows. Since operations personnel were not
- provided these technical insights, the Unit 3 recirculation loops were operated,
with mismatched flows greater than the technical specification required limits for over 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. This condition resulted in the vibration stresses in the loop 'A'
riser braces exceeding the design endurance limit, above which vibrations measurably increase design fatigue usage.
Also, Unit 3 experienced a runback of the 3A pump in December 1993 due to '
'
the loss of power to the same relay that dropped out during this event. Part of-the corrective action for this event was to install a modification which would
..
. > provide a non-interruptible power supply to the recirculation pump runback
>
relays. This corrective action, which could have prevented the 3A runback on :
June 7, was never performed. (Section E1.1)
e.
. Some emergency diesel generator (EDG) oil leak reduction strategies were not well-implemented or well-communicated to operations personnel. These factors
!
contributed to oilleaks and flames observed on the E2 and E1 EDG exhaust manifolds in May and June,1998, respectively. Engineering personnel H
~ concluded that EDG operability was not affected by the flames, but they were
]
<
- pursuing additional leak reduction initiatives. (Section E2.1).
{
'
e Reactor engineering personnel provided good engineering support of plant operations by ensuring timely analysis and effective resolution of degraded conditions on Unit 2 scram solenoid pilot valves. (Section E2.2)-
Plant Suonort:
e 1.icensee personnel demonstrated good performance during an emergency preparedness mini-drill conducted on June 15,1998. The drill was adequately controlled, and the post-drill critique was good. (Section P1.1)
iv
-
,
" -
-
__
-__
_ _ _
_ _ _ - _ - _ _ _ _ _
______ _ _ __ _ _ _ _____ ____ - _.
l
.
l l
l
i i
TABLE OF CONTENTS EXECUTIV E S U M M ARY.............................................. ii TAB LE O F CO NTENTS -................................................ v Summary of Plant Status............................................ 1 1. Operations
....................................................1
.
. Conduct of Operations.................................... 1 01.1 General Comme nts.................................. 1 01.2 Reactor Level Excursion Caused by Mis-operation of 2A Reactor Fe e d Pum p....................................... 1
01.3 Stator Water Cooling Pump Trip (Unit 3)................... 3 I
O2 Operational Status of Facilities and Equipment.................. 4 02.1 Failure to Energize Trip Relay for Main Control Room Emergency Ventilation as Required by Technical Specifications and (Closed)
Licensee Event Report (LER) 50-277(278)/2-98-03........... 4
03 Operations Procedures and Documentation...................... 6 l
03.1 (Closed) Unresolved item 98-02-01 Plant Status Control issues.. 6 03.2 (Open) Unresolved item 50-277(278)/97-07-03 Unit 2 Cooldown Monitoring Following the Ncvember 9 Reactor Scram......... 10
Operator Knowledge and Performance........................ 12 04.1 Unit 3 Recirculation Loop Mismatch Following the Runback of.
the 3 A Recirculation Pump........................... 12
Miscellaneous Operations issues............................ 15
,
08.1 (Closed) LER 98-001 Failure to Perform Surveillance Required l
for First Recirculation Pump Start
......................15 11. M aint e n a n c e.................................................. 1 6
-!
M1 Conduct of Maintenance.................................. 16
!
M1.1 General Observations............................... 16 i
M1.2 Scram Solenoid Pilot Valve Replacement Activities (Unit 2).... 18 Ill. Engineering.................................................. 18 i
E1 Conduct of Engineering................................... 18 l
E1.1 General Observation::...............................18 i
E2 Engineering Support of Facilities and Equipment................. 19 E2.1 Emergency Diesel Generator Exhaust Flames.............. 19 j
E2.2 Scram Solenoid Pilot Valve Degradation....
.............21 i
E2.3 (Closed) Valve Operation Test and Evaluation System 10 CFR21
)
Notification, " Potential Thmst Limit Exceedance"........... 22 IV. Plant Support
.......................................-.........23 l
R1 Radiological Protection and Chemistry (RP&C) Controls............ 23 l
[
. R1.1 Locked High Radiation Door and Posting inspections During
,
)
Pl a nt Tou r s....................................... 2 3
!
!
!
I
_-______________ - _______ A
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ _ - _ _ _ _ - _ - _ _ _ - _ _ _ _ _
.
l l
. Table of Contents-(cont'd)
P1 Conduct of Emergency Preparedness Activities.................. 23 P1.1 Observation of Emergency Preparedness Drill.............. 23 S1, Conduct of Security and Safeguards Activities.................. 24
,
S1.1 Security Force Personnel Compensatory' Actions During Motion Detection System Repeirs............................ 24 -
S2 Status of Security Facilities and Equipment..................... 25 S2.1 General Integrity of Protected Area (PA) Barriers............ 25
. V. M anagement Meetings.......................................... 2 5 X1 Exit Meeting Summary................................... 25 X2 Predecisional Enforcement Conference........................ 25 -
X3 Review of Updated Final Safety Analysis Report (UFSAR) Commitments. 26 ATTACHMENTS
- Attachment 1 - List of Acronyms Used
.,
-Inspection Procedures Used
-
- Items Opened, Closed, and Discussed
.
. Attachment 2 - Licensee's Presentation Materials i
-
,
!
)
vi
,
,
._
_ _ _ _
s.
,
1.;
,
Report Details p
Summary of Plant Status
. PECO operated both units safely over the period of this report.
...
Unit 2 began this inspection penod at 100% power. On May 12,' unit load was reduced to withdraw a control rod following repairs to one of its scram solenoid pilot valves.- On ~
May 20, the unit began end-of-cycle thermal coastdown. The unit power was reduced on
. May 22 for condenser waterbox cleaning." On June 1, unit load was reduced following a
"
scram of a control rod during reactor protection system testing. The control rod had a -
leaking scram solenoid pilot valve. The unit power was reduced on June 5 to facilitate -
, control rod hydraulic control unit (HCU) on-line maintenance to replace several scram solenoid pilot valves. On ' une 7, unit power dropped to approximately 37% following a J
runback of both r' circulation pumps due to a degraded voltage condition from the -
i, unexpected loss of a 500 kilovolt line during an electrical storm. Unit 2 power was at 92% at the end of the inspection period, b
-
Unit 3 began this inspection period at 100% power. 'On May 29, unit load was reduced to
' clean condenser waterboxes. On June 7, unit power dropped to approximately 68%
"
following the runback of the 3 'A' recirculation pump due to a degraded voltage condition
.from the unexpected loss of a 500 kilovolt line during an electrical storm. On June 9, the.
unit was returned to 100% power, where it remained for the rest of the period. PECO..
L
.
- 4 l
m.
-
-occasionally reduced unit load for control rod pattern adjustments and other activities.'
-
1. Operations 01'
Conduct of Operations'
01.1 General Comments (71707)
,
The inspectors observed increased peer-checking for most evolutions performed in the control room, following the 2A reactor feed pump mis-operation event (discussed in Section 01.2). More deliberate self-checking and increased control
- room supervisor oversight of routine control room operations were also observed following this event through most of the inspection period.-
01.2 L Reactor Level Excursion Caused by Mis-ooeration of 2A Reactor Feed Pumo
'
a.
Inspection Scone (71707)
,
- *
On May 16,1998, control room operators were preparing to remove the 2C reactor a:
sfeed pump turbine (RFPT) from service as part of a down-power evolution. The 2C ho2 r. \\
Jv:
-1 Top 6 cal headings such as 01, Ms, etc are used in accordonne with the NRC standardized reactor irspection report outline. Individual soporte are
{
< not espected to address all cuttlne tapece.
,
_
.
-
l
p.
L+
>
RFPT was in standby, and the operators were proceeding to fully shut it down.
However, the Unit 2 reactor operator inadvertently selected the 2A RFPT and -
manually. reduced speed. After about 25 seconds, operators received alarms gyo
,,
!'
indicating a low reactor water level and recirculation system runback to approximately 45% flow.
The inspectors reviewed documentation associated with this event and discussed the event, including corrective actions, with operations staff and management.
b.
Observations and Findingg Operator response to the transient was good. Operators entered off-normal procedures and took appropriate actions to restore reactor water level to normal.
l
. Operations personnel investigated this event and found that the operator had the correct procedure in hand and was following it step-by-step. The evolution was being peer-checked by another operator, but just before the event, the peer-checker stepped away to answer the telephone. Operations personnel noted that a similar uincident of selecting the incorrect reactor feed pump had recently occurred in
> simulator practice for another shift crew. This incident was actually discussed during the shift briefing prior to the evolution in which the event occurred.
Although the licensee's investigation was still in progress at the end of the
,
i-m.
Inspection period,' operations personnel preliminarily identified two root causes for this event:
Communication failure. - The reactor operator communicated his intentions
regarding the feed pump operations, but did not recall any acknowledgment from the control room supervisor or other reactor operators.
Failure to peer check. The reactor operator assigned to perform a peer check had stepped away from the panels just prior to the event.
Following thih svent, operations management revised its expectations regarding peer checking. Operation management directed that, with a few exceptions, all operations in the control room be performed with a second operator as a peer checker. Management was also concerned about the fact that although the pre job briefing discussed peer checking and the possibility of selecting the incorrect RFPT, operator awareness / sensitivity.was not raised. Therefore, operations management was considering methods to enhance briefing techniques.
The inspectors observed that, in response to the revised expectations for peer checking, shift management implemented high standards for peer checks. The
- _ inspectors also observed that. control room supervisors. increased their oversight of
-.,
routine evolutions, in some instances, operators performed self-checking in a more i
deliberate manner following the event.
_ _ -. - -_ _____________ - -_ - - _
i
.
l
!
!
l Some operators stated that the expectations rngarding peer checking were not clearly defined. The inspectors discussed this with operations management and
.
.
_ learned that written guidance and other materials were under development and would be issued promptly.
The inspectors reviewed the event for procedure adequacy issues and identified no concems. The inspectors determined that this event was an operator performance
issue, involving weaknesses in communications, self-checking, and peer / supervisory review.
l c.
Conclusions
!
During a Unit 2 downpower evolution on May 16,1998, operators reduced speed on an incorrect reactor feed pump, resulting in a reactor level excursion and i
recirculation system runbeck. This event was indicative of poor operator i
performance, reflecting weaknesses in communications, self-checking, and peer / supervisory review. Following the event, the inspectors observed increased peer checking and improved oversight by control room supervisors.
01.3 Stator Water Coolina Pumo Trio (Unit 3)
a.
Insoection Scone (71707)
"The inspectors reviewed the circumstances that lead to the 3A stator water cooling pump tripping during troubleshooting on April 28,1998.
b.
Observations and Findinas I
i i
The 3A stator water cooling pump tripped when its control switch was repositioned l
from auto to manual, through tne off position. Operations and l&C personnel involved in a stator cooling water system troubleshooting effort incorrectly expected
~
the pump to remain in operation when the pump switch was repositioned. This issue had the potential to initiate a turbine trip due to a loss of stator cooling.
Operations personnel performed a good investigation and identified important issues related to operations review of planned work and communications. They noted that the shift crew that reviewed the work did not request assistance from other plant personnel who were more knowledgeable of the switch circuitry. The operations technical review was performed only by the work control supervisor, who typically l
oversees numerous jobs. Operations management determined that a review by an off-shift operator before submittal to the work control supervisor would have been j
more appropriate. Also, operations personnel found that restrictions on lifting leads
'
during troubleshooting, which operations management placed on the job, were not
.
. communicated to either the veork control supervisor or l&C personnel.
The inspectors noted that the lessons learned from these issues were discussed with operations personnel. The inspectors interviewed selected l&C personnel l
following this event and learned that operators had begun questioning l&C work l
I i
!
!
--_
_ _ _ _ _ _ _ _ _ _ - - - _ - _ _,
.
..
more thoroughly, particularly when the work involved non-routine items or had the potential to initiate a plant transient. The inspectors also noted that some testing
. plans were being routed through operations support personnel prior to being submitted to shift crews, providing an additional review opportunity.
c.
Conclusions The 3A stator water cooling pump tripped during system troubleshooting efforts on April 28,1998, due to weaknesses both in operations review of the work and with communications regarding restrictions on the work scope. Operations personnel performed a good investigation of this issue and initiated appropriate corrective actions.
Operational Status of Facilities and Equipment O2.1 Failure to Eneraize Trio Relav for Main Control Room Emeraency Ventilation as Reauired by Technical Specifications and (Closed) Licensee Event Reoort (LER) 50-277(278)/2-98-03 a.
Insoection Scooe (71707 & 62707)
On May 15,1998, operations personnel identified that the trip relay for the Main Control Room Emergency Ventilation (MCREV) radiation monitor was not in the tripped status while the 'B' channel radiation monitor was inoperable. Because the r
trip relay was not tripped, the MCREV had inadvertently been inoperable for
{
approximately 28 hours3.240741e-4 days <br />0.00778 hours <br />4.62963e-5 weeks <br />1.0654e-5 months <br />. Technical specification actions required the trip to be l
installed within six hours after either channel of MCREV is inoperable. The inspectors discussed this event with operations and instrument and control personnel and reviewed applicable documentation.
I b.
Observations and Findinas i
While removing a jumper installed on the Main Control Room Emergency Ventilation (MCREV) radiation monitor, it was discovered that the trip relay was not in the tripped status. The jumper had been installed by operations personnel to place a short circuit across the 'B' channel radiation monitoring instrumentation and initiate a Division 11 ! solation of the MCREV system. This short circuit should have provided a trip in the logic that was associated with the 'B' monitor. Subsequent investigation by the licensee revealed that the trip relay coil was not energized due to a loose terminal screw on the trip relay coil. The lead was retightened at the terminal and other leads for the MCREV relays were checked and tightened as necessary.
l During the investigation of this event, the licensee determined that. operations personnel did not verify that the relay had tripped when they installed the jumper.
Also, the procedure used to install and remove the jumper, General Procedure (GP)-
25, Appendix 14, Revision b, " Main Control Room Ventilation isolation, Division 11,"
!
!
i
_ _ _ _ _ _.
_ _ _ __
-,
...
.
'
did not require positive verification that the installed jumper actually resulted in the desired trip.
On July 10,1997, a similar event occurred when the 'A' channel of the MCREV was not maintained in the tripped condition for greater than six hours while the channel was removed from service for repair and testing. As documented in LER
.
50-277(278)/2-97-04,a keylock switch that was used for installing the trip was not adequately controlled.
Part of the corrective actions for this event included revising GP-25, Appendix 13, Revision 6,1" Main Control Room Ventilation isolation, Division I" and GP-25, Appendix 14 to include a positive verification that a trip was inserted when a jumper is installed.
The inspectors learned during discussions with instrument and control personnel that maintenance personnel were reviewing their work practice standards and'
procedures for electrical terminations as part of the corre':tive actions for this event.
This review was being performed to ensure that maintenance personnel were-m. verifying that the expected action took place when a jumper was installed or othere
- electrical termination or lifting was performed. The inspectors learned during
discussions with operations personnel that all GP-25 procedures which install trip functions will be reviewed to ensure that the procedures require positive verification that the desired action did occur after the trip was installed.
,
.The inspectors noted during the review of Operations Manual, OM-P-7.7,
~
Revision 1,." Boots and Jumpers" that this procedure also did not contain any
. instructions to verify that an expected action occurred when a jumper was installed
- to initiate a trip function. The inspectors also noted that Administrative Procedure A-C-OO1, Revision 0, " Procedure Writer's Guide" stated that " quantitative and qualitative acceptance criteria, signoffs, and inspection points shall be included to -
- W+
- assure that important activities are satisfactorily completed." The inspectors were concerned that the corrective actions for the July 10,1997 event, including the procedure revisions to GP-25, Appendix 13 and 14, were not comprehensive enough to prevent this subsequent event.
Technical specification 3.3.7.1 requires that when a channel of the MCREV system L
. instrumentation is inoperable that the channel be placed in trip within six hours.
Contrary to the above, on May 15,1998, operations personnel identified that the
'B' channel had not been placed in the tripped condition for approximately 28 hours3.240741e-4 days <br />0.00778 hours <br />4.62963e-5 weeks <br />1.0654e-5 months <br />.
'
L This channel was required to be tripped since an inoperable radiation monitor (RIS-07608) rendered this channel of MCREV instrumentation inoperable. (VIO 50-277(278)/98-06-01)
. _ _ _ _ _ - - _ _ _ _ _ _ _ _
_
- - - _ _ _ _
_ - _
.
l
c.
Conclusions H
.
., On May 15,_1998, operations personnel identified that the trip relay.for_the Main
. Control Room Emergency Ventilation (MCREV) radiation monitor had not been in the tripped status for approximately 28 hours3.240741e-4 days <br />0.00778 hours <br />4.62963e-5 weeks <br />1.0654e-5 months <br /> while the 'B' channel radiation monitor was inoperable. This condition resulted in a violation of technical specification
' 3.3.7.1 since the 'B' channel was required to be tripped within six hours after the channel became inoperable.
'
The operations personnel installing the jumper to initiate a Division ll isolation trip of the MCREV radiation monitor did not perform, nor did the procedure instructions require, a positive verification that the trip was properly inserted. The corrective actions from the July 10,1997 event were not comprehensive enough to prevent L
this subsequent event.
' Operations Procedures and Documentation
<
O3.1 (Closed) Unresolved item 98-02-01 Plant Status Control Issues
!
a; inspection Scone (71707,62707 & 375511'-
L
'
- The _ inspectors completed an inspection of plant status control / configuration.
-
control issues that was begun during inspection period 98-02. The inspectors
.
i
- reviewed:
A Unit 3 main steam line flow instrument isolation valve found out of its
required position.
Residual heat removal (RHR) valve HV-2-10-65 found out of its required
position.
RHR valve HV-3-10-65 found out of its required position.
- .
- 3D high pressure service water (HPSW) pump packing leakoff valves found
mis-positioned.
Containment atmosphere dilution (CAD) vaporizer isolation valve out-of-
position.
L b.
Observations and Findinas i
l During this inspection period, the inspectors noted that plant status control issues l
co'itinued to be identified by licensee reviews or through events. Both the 3D
!~
HPSW pump leak-off valves and the CAD isolation valve mispositioning issues
,
l~
occurred during the current period.
The inspectors examined the five items listed above for commonalities and potential
- _ generic. implications.mThe. inspectors determined that there was a common element m,
m-of inconsistency between plant procedures, check-off lists, drawings, administrative
- guidance, and/or operations and maintenance practices, among all these items.
l l'
p i'
'
_ _ _. __
]
____ _ _____ ______.;____. _ ___________ _ _. _. _ _ _. _ _ _ _ _ _ _ _ _. _ _. _ _ _ _ _ _ _-
_
___ _
_____
. _. _
_
_
__
_ _ _ _ _ _ _
_ _ _ _ _ _ _ - __ _______
_ - _ _ _ _ - _ _ _
___
.-
,.
Specific findings for each item are described below:
- . Unit 3 Main' Steam Line Flow Instrument Isolation Valve
__
-
-
' NRC Inspection Report 50-277(2781/98-02(Section M4.1) documented that l&C technicians found the low side isolation valve (ISV-3-02-1178L)for the 'B' main steam line' flow instrument shut, when its required position was open. This valve was found shut while Unit 3 was shutdown. l&C personnel concluded that the instrument was operable while Unit 3 was in operation, but they did not determine the specific activity that caused the valve to be left out of position.
l&C personnel identified the following:
The procedure that last controlled the operation of ISV-3-02-117BLdid not
include valve-by-valve restoration instructions. Specifically, ST-l-02B-650-3,
.
" Excess Flow Check Valve Operability," Revision 6, only required technicians to verify the restoration of the instrument. l&C personnel recommended that the procedure be changed to require valve-by-valve instructions.
The inspectors identified:
The ST-l-028-650-3 procedure did not specify independent verification of the
restoration of several safety related instruments. -This was inconsistent with
'
-
t guidance in A-C-33, " Nuclear Generation Group Process for Verification of.
i Quality," Revision 4 and A-C-1, " Procedure Writer's Guide," Revision 0,
~ which indicate that independent verification should be used when returning safety related equipment to normal.
RHR Valve HV-3-10-65 Found Out of its Reauired Position
' NRC Inspection Report 50-277(278)/98-02(Section E2.2) documented that
+
operations personnel found RHR stayfull system valve HV-3-10-65 out of its required position. The valve was open, instead of closed, as specified in plant check-off lists and drawings. While investigating this issue, operations personnel also found the corresponding Unit 2 valve (HV-2-10-65) out of position.
Engineering personnel completed an investigation in May 1998, and noted that:
Shift operations personnel specified an incorrect restoration position in
-
l clearance #97002299for valve HV-3-10-65.' Licensee personnel could not determine what documentation was used as a basis for the specified
- restoration position. Operations has revised the clearance and tagging practices to have operations support personnel, instead of shift personnel, determine most restoration instructions.
!'
,
,
(
_ _. _ _ _ _ _ _ _ _ _ _ _ _ _ _. _ _ _ _ _
J
_ - _ __ -_____-_________--__-_--
_. _ - _
.
.
- CAD Vaoorizer Isolation Valve Out-Of-Position On May.28,1998, during surveillance testing, operators found that.both s.
containment atmosphere dilution (CAD) electric vaporizers were unintentionally taken out of service. They found that a clearance (#98001917)that was applied the previous day left both HV-O-7C-11 A and -118, " liquid nitrogen inlet to A(B)
' CAD vaporizer block valves," closed. The clearance closed HV-O-7C-11 A, but had no instructions regarding the -11B valve. Per system operating procedure SO 7C 1.A-2(3), " CAD System Startup/ Standby Operations," one of these valves should be open and the other closed so that one CAD vaporizer is lined up for l-operation. Operators also noted that the CAD system drawing showed both valves normally open, which did not agree with the system operating procedure.
Operations support personnel referred only to this drawing when preparing the clearance.
L This issue caused both CAD trains to be inoperable for a one day period, since the H
surveillance test that identified the mispositioning was performed the day after the
. clearance was applied. However, no technical specification non-compliances L
_ resulted, because a 30-day action statement applies to one or both trains being.
-
L Operations management discussed this event with all operations personnel and.
. initiated actions to review clearance and tagging policies. A revision to the CAD'
f~
system drawing was also initiated.
The inspectors noted that the CAD system drawing did not reflect the true
' operating configuration of the plant, as described in plant procedures and check-off lists. The inspectors determined that operations support personnel missed an opportunity to identify this problem by not referring to more than one source of plant configuration documentation. Operations shift personnel also missed an Hb'
+ opportunity to identify the clearance error by not using the system operating
- <* - -
.. procedure during the clearance application. The inspectors noted that no technical specification non-compliances resulted from this event.
3D HPSW Pumo Packina Leakoff Valves Found Mispositioned On May 26,1998, during post-maintenance testing, maintenance technicians observed an excessively high packing temperature on the 3D HPSW pump.
Subsequently, the technicians found the pump packing leakoff valves shut during a verification of the system valve lineup. The valves are listed in the system check-off list as being full open. Maintenance personnel noted that'the valves may have been shut for several months, and the pump had operated with excessive packing
'
leakage during this peri ( d. Operations and engineering personnel evaluated this condition and concluded.that the 3D HPSW pump remained operable with the
,,
.
leakoff valves shut.
Maintenance personnel determined that there were inconsistencies between the maintenance procedures, maintenance practices, and the system check-off list.
]
. _ - _ _ _ _ _ _ _ - _ _ _ - _ _ _ _ _ _ _ -
!
..
!
Although the check-off list indicated that the valves should be full open, maintenance technicians, based on vendor recommendations, throttled the valves to
. achieve an optimum packing leakoff. Maintenance personnel also noted that the
!
maintenance procedures provided insufficient instructions in this area.
l The inspectors noted that both maintenance and engineering personnel were continuing to investigate and identify corrective actions for this matter. The l
inspectors determined that, similar to other plant status control issues, this event l
revealed inconsistencies between procedures, practices, and check-off lists.
'
RHR Valve HV-2-10-65 Found Out of its Reauired Position i
in September 1997, Qual 8 y Assurance personnel identified that several procedures t
that re-positioned the HV-2-10-65 valve were in error and did not restore the valve
,
'
to the required position. One of these procedures, RT-O-010-610-2,"2A RHR Heat Exchanger Leak Test," Revision 5, was used by operators in March 1998, but the procedure had not yet been revised to reflect the correct position of valve HV-2-10-65. Thus, the valve was incorrectly restored to the open position. This condition
,
...was found following identification by the operations manager that the corresponding i
Unit 3 valve was out of position.
Engineering personnel completed an investigation in May 1998, and identified a
!
cause for the mispositioning:
l Engineering personnel performed an incomplete search of procedures
)
affected by Non-Conformance Report (NCR) 96-03167, which changed the
~
required position for the HV-2-10-65 valve from open to shut. The corrective action for this item was to train the engineering staff members who perform this function.
The NRC review of this event focused on the timeliness.and effectiveness of the
-
procedure changes associated with the HV-2-10-65 valve. The inspectors identified J
the following issues:
(
The procedure revisions for the QA-identified deficiencies were not
)
performed in a timely manner. QA noted the procedure problems in
'
September and initiated a Performance Enhancement Program (PEP) report, but the required revisions were not completed until late March, after one of the incorrect procedures was used.
,
The potential impact on system operability of the incorrect procedures was
not recognized until after the valves were found out of position. Engineering
!
personnel performed a detailed operability evaluation to determine that the (
RHR system remained operable with the valves in.the open position.
l i
!
l
__ _ _
_ _ - - _ _ _ _ _ - - _ _ _ _ _ _ - - - - - -
,.
- .
1O The corrective action documentation for the procedure revisions did not
+
clearly reflect the status of the changes. For example, one document indicated that the seven procedure changes had been completed in early March, yet some procedure revisions were not actually effective until March 29.
The inspectors determined that these findings represented a violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action. Following QA identification of procedure errors affecting the RHR system, engineering personnel failed to take prompt and effective corrective actions to resolve these procedure deficiencies. As a result, Unit-2 valve HV 2-10-65 was left out of its required position, and an engineering evaluation was necessary to determine that system operability was not affected. (VIO 50-277/98-06-02)
c.
Conclusions NRC review of a number of plant status controlissues related to mis-positioned valves in safety related systems identified a common element of inconsistency
..between plant procedures, check-off lists, drrwings,. administrative guidance, and/or operations and maintenance practices. In some ' instances, plant personnel missed opportunities to identify the discrepancies.
. In one instance, engineering personnel failed to take prompt and effective corrective
-
actions following QA's identification of errors affecting several procedures for the s*
residual heat removal system. Operators used one of the incorrect procedures approximately five months after QA's findings, and as a result, one valve was left out of its required position. Although system operability was not affected, an engineering evaluation was necessary to reach that conclusion. This was-considered a violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action.
a,i 03.2 > (Ocen) Unresolved item 50-277(2781/97-07-03 Unit 2 Cooldown Monitorino
.,
.
,
Followino the November 9 Reactor Scram a.-
Insoection Scoce (71707 & 37551)
The inspectors reviewed the licensee's evaluation of the reactor vessel / vessel drain pipe temperatures used to show that the required vessel temperature limits were maintained following the reactor scram of Unit 2 on November 9,1997. The inspectors also reviewed the actual temperatures indicated at various locations during this scram and the heatup and cooldown rates, including temperature differentials during recirculation pump starts for several heat-ups and cooldowns.
,
b.
Observations and Findinos:
i
- In NRC Inspection Report 50-277(278)/97-07,the inspectors documented that Unit 2 had stayed well within the maximum heat up or cooldown rate during and following the transient and that Surveillance Test (ST)-O-080-500-2, Revision 4,
" Recording and Monitoring Reactor Vessel Temperatures and Pressure" was
'
!
!~
l
_
_
_.__
_
_ _ _ _ _
_ _ _ _ _
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _
.
.
appropriately exited upon meeting the required temperature criteria. This was consistent with the inspectors' observations of the cooldown for Unit 3 on October 4-5,1997..However,.the data, taken during some periods of the plant cooldown after the scram, indicated that vessel metal temperature changes were greater in magnitude or in different directions (heatup vs cooldown) than the worst case recirculation / steam dome temperature changes.
The inspectors noted that the bottom head drain had a recorded heatup rate of 134* Fin 15 minutes and 136* Fin an hour right after the start of the first recirculation pump following the November 9,1997 scram. During this same time period, the steam dome temperature dropped 2*F in 15 minutes and increased 2 F in an hour. Technical Specification 3.4.9.1 required the reactor coolant system heatup and cooldown to be maintained s;100 *F in any one hour.
The inspectors also determined that the temperature difference between the bottom head drain coolant and the reactor pressure vessel during this recirculation pump start was 145 F. This was the maximum temperature difference allowed by Technical Specification 3.4.9.3. The inspectors determined this through independent review of computer time data sheets and the surveillance data of
--
recorded time and temperatures for the November 9 scram.
Based on the above infortnation noted above and other data that was reviewed, the inspectors were concerned that temperature stratification existed in the reactor vessel for some period of time prior to the recirculation pump start. Because of the
information noted above, the inspectors questioned PECO engineering personnel I
about the possibility and to what extent that stratification had existed in the reactor vessel, how these conditions affected thermal stress over the life of the vessel, and how and if these conditions would be needed to be reconciled in the thermal cycle tracking as described in the UFSAR. The inspectors were also interested in the i
relationship between the bottom head drain temperature change and the reactor
!
coolant temperature changes.
l I
PECO engineering initiated an Action P* quest (AR) to evaluate the November 9, 1997 temperature data and other sr' _ auent NRC questions regarding cooldown and heatup after this scram. This AR included an evaluation of temperature changes at the bottom head after the scram and after the recirculation pump start.
The AR also included an evaluation of the possibility of stratification in the reactor l
vessel after the reactor scram. This URI remains open pending completion of this action request by PECO engineering and review by the inspectors of the responses to this AR.
c.
Conclusion i
L
...The issue regarding whether the appropriate and/or most conservative vessel / vessel drain pipe temperatures were used for determining that the technical specification pressure / temperature limits were not violated is still pending. Also, the significance
of temperature changes on the bottom head drain after the scram and first recirculation pump start and the significance of reactor vessel temperature i
!
_ _ _
________________J
- - _ __ __ -_ _ _-__________-_ _ __ ______ ____-___ _ ___ -_ ______________ ________
. _ _ _
_ _. _ _ _ _ _ _
.
.
stratification remain unresolved pending additional response by PECO engineering and review by the inspectors.
Operator Knowledge and Performance 04.1 Unit 3 Recirculation Looo Mismatch Followina the Runback of the 3A Recirculation Pumo a.
Inspection Scoce (71707 & 37551)
On June 7,1998 at 4:02 pm, the 2A and 28 and 3A recirculation pumps ran back to 30% speed due to the unexpected loss of a 500 kV line (Line 5007 line to Three Mile Island) during an electrical storm and the slow opening of 500 kV breaker (No. 25). At the time of the event, Unit 2 was at approximately 70% power for performance of control rod drive hydraulic control unit (HCU) maintenance and Unit 3 was at 100% power. After the run backs, Unit 2 power was at approximately 42% and Unit 3 power dropped to approximately 68%. The 3B recirculation pump remained at full speed during this event. Due to the differences in pump speeds of the Unit 3 pumps, the flows in the recirculation loops were
~
sirrificantly mismatched. The recirculation loop flows remained mismatched oweide of Technical Specification Surveillance Requirement (SR) 3.4.1.1 for over 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
The inspectors reviewed the documentation and control room data including the strip charts related to the runback of the Unit 2 and 3A recirculation pumps. The inspectors also discussed the Unit 3 mismatched recirculation flow with operations and engineering personnel, b.
Observations and Findinos
-
Following tha runback of both of the Unit 2' recirculation pumps, the Table 1 rods were inserted since the unit approached Region 2 (i.e. Immediate Exit region) on the power to flow map. Reactor power decreased to approximately 37%. The 2A and
.
2B reactor feedwater pumps were placed into standby after the runback. Unit 2 remained in a stable condition following the actions noted above.
At 4:34 pm, the first eight rods of Table 1 on Unit 3 were fully inserted to give additional margin from the 115% rod line on the power to flow map. At 6:34 pm, 3B recirculation pump speed was reduced in order to bring the 'B' loop jet pump flow on scale. At 01:13 am on June 8, Unit 3 pow tr was increased by raising 3A recirculation pump speed. The TSA per Technical: specification 3.4.1 was exited at 04:30 am when the flow mismatch between the 'A' and 'B' recirculation loops was less than 10,250,000lbM/hr, Recirculation loop flows were matched at 04:45 am.
The operations crew had recognized the flow mismatch immediately after the 3A runback and entered the Technical Specification Action (TSA) per Technical l
Specification 3.4.1. Based on interviews with the shift managers and control room supervisors for the day and night crews, they all believed that Limiting Condition for i
- - _-_ - _-_ - _-____-_ _ ___-
_ _ _
. - _ _
-.
l l
- Operation (LCO) 3.4.1.D allowed them to not match recirculation loop flows for up I
to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> from the time the TSA was entered.
immediately following the runback, the recirculation system manager, reactor
,
L engineering manager, operations manager and plant manger were notified about the I
condition of Unit 3 and the recirculation ioops flow mismatch. None of these individuals expressed significant c,oncern with the mismatched flows. Also, a reactor engineer was present in the control room, prior to the event, providing
- technical oversight for Unit 2. This engineer provided technical assistance on Unit 3 after the runback and he did not express any significant concerns with the mismatched flows.
On June 11, the licensee discovered that the flow mismatch probably caused the loop 'A' riser brace to exceed the design endurance limit, above which vibrations measurably increase design fatigue usage. This was based on a qualitative assessment by General Electric Nuclear Engineering. However, this assessment concluded that jet pump operability would be maintained through the remainder of the Unit 3 operating cycle.
j The inspectors reviewed the qualitative analysis performed by PECO engineering and General Electric personnel and had no significant concerns with this initial analysis..The inspectors also reviewed Unit 3 plant data following the 3A runback
{
and other recirculation system documentation and had no concerns with fuel
)
thermal limits or other design basis concerns.
Based on interviews with operations and engineering personnel, the inspectors learned that none of the personnel involved with this event knew that the mismatch in recirculation flows could cause excessive vibration of the jet pumps and the jet
)
pump riser braces. Although the day shift, Unit 3 operator told the inspectors that he was uncomfortable with the mismatched flows and that he had discussed his
-
~ concern with his supervisors, he also was unaware of the potential for excessive vibration due to recirculation flow mismatch.
During review of this event, the inspectors determined, based on Technical Specification 3.4.1 Bases information, that Unit 3 was in single recirculation loop operation from 4:02 pm on June 7,1998 until 04:30 am on June 8,1998. Neither the day shift or night shift operations crew entered the abnormal operating procedure for single recirculation loop operation. Based on interviews with operations and engineering personnel, none of the operators, engineers, or managers involved with this event realized that Unit 3 was in single loop operation.
Concerns with recirculation flow mismatch causing excessive vibration of the jet pumps and the jet pump riser braces were described in the Peach Bottom Design
!
- Basis Document (DBD) for the recirculation system. The inspectors. reviewed the
!
procedure used by operations personnel during this event, General Procedure
!
(GP)-5, Revision 38, " Power Operations." This procedure did not contain any information or actions regarding the concern that mismatched recirculation loop i
-___-_
- _ - _ _ _ _ - _ - - _ _ - _ _ _ - _ -
_ - _ _ _ __
l c.
j, L.-
. flows would cause excessive vibrations of the jet pumps and the jet pump riser braces.
The inspectors also reviewed the Abnormal Operating (AO) procedure, AO 2A.1-3,
!
Revision 17, " Recirculation System Single Loop Operation." The inspectors noted that this procedure did not contain any information regarding excessive vibration due to mismatched loop flows. This procedure also did not contain any of the information contained in the Bases information for Technical Specification 3.4.1 about being in single loop operation when the mismatch between the two loops is greater than required limits.
Technical Specification 3.4.1 requires that with one recirculation loop in operation, the Reactor Protection System (RPS) Instrumentation, " Average Power Range-Monitors Flow Biased High Scram," allowable value shall be reset for single recirculation loop operation. The reset of the " Average Power Range Monitors Flow Biased High Scram" allowable value for single loop operation may be delayed for up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after transition from two recirculation loop operation to single loop
,
operation.
!
. Contrary to the above, the inspectors identified that on June 8,1998, plant personnel did not reset the Unit 3 RPS Instrwentation, " Average Power Range Monitors Flow Biased High Scram" within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after transition from two.
h recirculation loop to single loop operation. Specifically, this action was not
.
a*
A performed because operations and engineenng personnel did not recognize that
-
+>
Unit 3 was in single loop operation following the 3A recirculation pump runback on June 7,1998. Per the Bases information for Technical Specification 3.4.1, Unit 3 was in single loop operation since the mismatch in flows between the two
- recirculation foops was greater than the allowable limits in Surveillance Requirement 3.4.1~.1 following the 3A pump runback. (VIO 50 278/98-06-03)
c.
Conclusions J
On June 7,1998, the 3A recirculation pump ran back to 30% speed due to the unexpected loss of a 500 kV line during an electrical storm and the slow opening of 500 kV breaker. The 3B recirculation pump remained at full speed during this event. Due to the differences in pump speeds of the Unit 3 pumps, the flows in the recirculation loops were significantly mismatched. The recirculation loop flows remained mismatched outside of Technical Specification Surveillance Requirement
' (SR) 3.4.1.1 for over 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. This mismatched flow resulted in a violation of Technical Specification 3.4.1.
Operations and engineering personnel did not understand the effects of the recirculation flows mismatch on the jet pumps and jet pump riser braces due to tho me excessive vibration stresses. They also failed to recognize that Unit 3 was in single
,
loop operation following the runback of the 3A recirculation pump.
,
__-_
- - _
l
!-
In addition, no information was documented in the general procedure used during j
this event that would alert the operators to the need to balance recirculation flows l
l
..quickly to prevent high vibration stresses in the jet pump loops. Also, the abnormal procedure for single loop operation did not contain any information regarding the vibration concerns or technical specification bases information about being in single loop operation when the mismatch between the two recirculation loops was greater than required limits, i
Miscel!aneous Operations issuas 08.1 (Cosed) LER 98-001 Failure to Perform Surveillance Recuired for First Recirculation Pumo Start On March 23,1998, during the review of the URI discussed in Section O3.2, the l
licensee discovered that operations personnel had not been verifying the
!
temperature difference between the reactor coolant in the recirculation loop and the I
reactor pressure vessel (RPV) coolant during the first recirculation pump start. This verification was required by Techmcal Specification Surveillance Requirement (SR) 3.4.9.4. The difference between the reactor coolant temperature in the
_
recirculation loop to be started and the RPV coolant temperature was required to be
within 50 F during the recirculation pump start. This requirement had not been incorporated into the surveillance test (ST) or system operation (SO) for the first recirculation pump start following the change to improved technical specifications on January 18,1996.
Operations personnel had started a recirculation pump six times on both units with both loops idle since improved technical specifications had been issued. On October 29,1997, the Unit 3 temperature differential was 84 F (RPV coolant temperature was 163 F with the 'B' recirculation loop temperature at 79*F). An analysis performed by the licensee indicated that the threshold for defining a thermal cycle would typically require a difference of 100 F. Therefore, this event was not counted as a thermal cycle even though it exceeds the 50 F critena.
The licensee found that the ST-O-02B-510-2(3)," Reactor Coolant Temperatures" did not contain the Technical Specification SR 3.4.9.4 requirement for the start of the first recirculation pump. Therefore, the verification of the difference between the recirculation loop and the RPV coolant temperature was not performed within 15 minutes prior to each startup of a recirculation pump with both loops idle.
Plant engineering wrote an NCR to evaluate and document these events.
Engineering personnel concluded that the thermal fatigue effects of the 84*F differential temperature had negligible impact on the affected components and
,
would not have resulted in any fuel performance or reactor thermal limit concerns.
The inspectors performed an on-site review of this LER and the documentation related to the corrective actions for the missed surveillance required by technical specifications. The inspectors reviewed the procedural changes to ST-O-02B-510-2(3), " Reactor Coolant Temperatures" and SO 2A.1.A-2(3), " Starting The First i
I
. _ _ --____ __-__ -___-_- - __
.
i
Recirculation Pump" arid verified that these procedures had been updated to include the Technical Specification SR 3.4.9.4 requirements. During this review, the
. inspectors determined that the licensee missed the above requirement during the conversion from custom technical specifications to improved standard technical specifications. The inspectors were concerned that other technical specification requirements may not have been incorporated into the licensee's administrative procedures during this conversion. The inspectors determined, based on review of the documentation associated with this issue, that no thermal or fuel performance issues existed.
Technical Specification SR 3.4.9.4 required that the difference between the reactor coolant temperature in the recirculation loop to be started and the RPV coolant temperature be verified as less than or equal to 50 degrees F during a recirculation pump start. This SR, which was applicable to reactor operational modes 1,2,3, and 4, was to be accomplished within 15 minutes of the pump start.
Contrary to the above, the licensee did not verify that this requirement was met during six recirculation pump starts between January 18,1996 and March 23,
-
1998. In aadition, on October 29,1997, the Unit 3 temperature differential was 84*F within 15 minutes of the 'B' recirculation pump start. (VIO 50-277(2781/98,
06-04)
c.
Conclusion On March 23,1998, the licensee identified that they failed to properly implement the improved Technical Specification Surveillance Requirement 3.4.9.4 for the start of the first recirculation pump. Between January 18,1996, and March 23,1998, operations personnel were not verifying that the temperature differential between the reactor coolant in the recirculation loop being started and the reactor pressure vessel coolant was within 50 F. On October 29,1997, the 'B' recirculation pump was started with a differential temperature of 84 F. Although this did not exceed design limits nor impact fuel performance, it was a violation of Technical Specification Surveillance Requirement 3.4.9.4.
II. Maintenance M1 Conduct of Maintenance M1.1 General Observations NRC Inspection Procedures 62707 and 61726 were used in the inspection of plant maintenance and surveillance activities. The inspectors observed and reviewed
!
, selected portions of the following maintenance and test activities:
._
-_
_ _ _ - _ - - _ _ -
__
_ _ _ _ _ _ _ - _ _ - _ _ _ - _ _ _ _ _ - - _ _ - _ - _ _ _ _ - _ - _ _
..:
,.
Malistenance Observations:
Observed On:
L,
.E2 Emergency Diesel Generator Lube Oil Standby Heater.
. May_15,1998 -
Temperature Switch, R0713456 Security Fence Work - Rejlace E-Field, C0179243 June 5,1998 Control Room Panel 20C0112, C0181136
_ May 7,1998 Emergency Lighting Battery Preventive Maintenance, R0731366 June 4,1998 Surveillance' Observat.i.q_n_g:
Observed On:
RT-O-052-201 -2 E1 Emergency Diesel Generator Load Run June 9,1998 RT-O-052-202-2 E2 Emergency Diesel Generator Load Run June 18,1998 ST-O-40D-320-2
'A' Control Room Emergency Ventilation June 11,1998.
Train Operational _ est T
ST-O-023-301 -2 High Pressure Coolant injection (HPCI)
June 11,1998 Pump, Valve and Flow Test.
L" ST-O-007-420-2 Primary Containment isolation System June 11,1998 Normally Closed Valves Operability Test a ST-l-07D-485D Primary Containment isolation System June 16,1998 a L
Valves The work and testing performed during these activities was professional and i
thorough. Technicians were experienced and knowledgeable of their assigned
' tasks. The work and testing procedures were present at the jobsite and actively v
+
~
j used by the technicians and operators for activities observed. Good pre-job briefs were observed prior to the performance of the surveillance observed. System managers were usually present during surveillance tests and provided good technical oversight.
During ST-O-023-301-2,a through-wall leak was identified in a one inch drain line i.
-
Toff of the Unit 2 HPCI pump suction line from the torus. The leak was unisolable and was located at a socket weld connection where the drain line was connected to a half coupling attached to the 16 inch pump suction piping. The leak rate through r
'
the flaw was approximately 19 drops per minute.
PECO requested a non-code repair for the leak on this ASME Section XI, Code Class
. 2 piping on June 17,1998. The NRC approved this welded non-code repair on June 19 and the licensee completed the repair and returned the HPCI system to operable status on June 20,1998.
L The inspectors noted that the identification of HPCI leak during ST-O-023 301-2 was an example of good equipment operator performance and showed that operators were attentive to their duties during this surveillance.
I L
_.--- _ _. _ --- _ -- _
_
.
.
M1.2 Scram Solenoid Pilot Valve Replacement Activities (Unit 21
. a.
Insoection Scope (62707)
The inspectors observed portions of on-line scram solenoid pilot valve (SSPV)
replacement activities conducted from June 5 thiough June 10,1998. The inspectors observed maintenance activities in the reactor building and coordination and testing activities in the control room.
b.
Observations and Findinos The inspectors observed that the SSh/ maintenance work was well-controlled by Nuclear Maintenance Division personnel. Maintenance technicians displayed good use of procedures and were knowledgeable of the work and radiological control requirements. Maintenance supervision monitored the work and interfaced well with operations personnel.
Similarly, the inspectors observed that operations department personnel provided
., good coordination and maintained good cognizance of the work status. Operatorsu
'in the field utilized appropriate self-checking techniques while applying clearances.-
Field supervisors oversaw the operators' efforts and communicated well with control room personnel.
- Status-keeping and testing in the control room were also very good. The inspectors observed strong operations supervisory overr/ght sad good coordination with reactor engineers during scram time testing, c.
Conclusions Unit 2 on-line scram solenoid pilot valve replacement activities were particularly
-
- well-executed. Nuclear Maintenance Division technicians and operations personnel displayed good procedure usage and sound work practices. Supervisory personnel provided very good coordination and oversight.
111. Enaineerina E1 Conduct of Engineering E1.1 General Observations (37551)
The inspectors observed generally good engineering support for maintenance and surveillance activities throughout this inspection period. Administrative procedure, AC-CG-50, Revision 0, " Equipment investigation and Troubleshooting Guidelire" was issued during this inspection period. This procedure wes developed because of the challenges that have been noted during the past year with the approach to troubleshooting by several system managers, most. notably with the degraded 2A reactor feedwater pump turbine high level trip function. The inspectors noted that
_ _ _ - _.
___ _ __.
._.
__
_______-____ -
- .
l l.
.
L this procedure provides guidance for a more systematic approach to future l
troubleshooting activities and equipment problem identification and corrective -
~
action.
The inspectors were concerned during this period with the failure of engineering personnel and management to recognize the potential for high vibration stresses on the 'A' jet pump loops due to the large recirculation flow mismatch following'the 3A recirculation pump runback on June 7,1998. The potential for recirculation flow l;
mismatch to cause excessive vibration of the jet pumps and the jet pump riser braces was described in the Peach Bottom Design Basis Document (DBD) for the
!
h recirculation system. This lack of understanding of the effects of this mismatch l
contributed to the failure of engineering personnel to provide the necessary.
L technical information to operations personnel. This resulted in the failure to recognize the need to expeditiously resolve the mismatch between recirculation loop
~
flows.' Since operations personnel were not provided these technical insights, the Unit 3 recirculation loops were operated with mismatched flows greater than the technical specification required limits for over 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. This condition resulted in
,
'
the vibration stresses in the loop 'A' riser braces exceeding the design endurance
< limit, above which vibrations measurably increase design fatigue usage.
.a Also, Unit 3 experienced a runback of the 3A pump in December 1993, due to the r-loss of power to the same relay that dropped out during this event. Part of the
- corrc'etive action for this event was to install a modification which would provide a -
s
- non-interruptible power supply to the recirculation pump runback relays. :This
!-
w
-
L corrective action, which could have prevented the 3A runback on June 7, was
'
L never performed.
t
!-
E2-Engineering Support of Facilities and Equipment
- E2.1 Emeroency Diesel Generator Exhaust Flames l
a.
- Inspection Scope (37551)
The inspectors reviewed two events in which operators observed flames at
<
. emergency diesel generator (EDG) exhaust manifolds caused by lubricating (lube) oil leakage, b.
Observations and Findinas i
- On May 5,1998, during testing, operators observed candle-sized flames on the E2 EDG exhaust manifold. On June 9,1998, plant personnel and the inspectors observed smoking and small flames on the E1 EDG exhaust manifold during routine testing. In both events, lube oil was leaking at the exhaust manifold flanges, and
. the oil occasionally flashed.and self-extinguished as the. temperature of the exhaust -
manifold increased during EDG loading. The smoking and leakage essentially stopped several minutes after the EDGs were fully loaded.
l
-_
_--
_ _ _ - - - - - _ _ _ _ _ _ _ _ _ _ _ _ _ _
'
j
...
i
%
.
'
'
' The exhaust manifold lube oil leakage and occasional self-extinguishing flames were documented in a 1994 Fairbanks Morse Owners' Group document as a known -
mindustry problem with Fairbanks Morse opposed piston EDGs.Jhe document stated
-.
that the lube oil drains past piston rings in the upper inverted pictons and accumulates in the exhaust piping during engine idle periods between surveillance
- tests. When the EDG starts, exhaust pressure forces the excess oil out of the flanged joints. As the exhaust manifold heats up, flames sometime occur until the leaked oil is consumed.
The Owners' Group document discussed a number of practices that some licensees used to limit the exhaust manifold lube oilleakage. Some of these oil reduction strategies were underway at Peach Bottom. However, the inspectors determined that a few of the strategies were not well-implemented or well-communicated to l
operations personnel.
For example, the docuinent recommended that licensees minimize the time that an EDG is run unloaded. This practice lessens the amount of oil that enters the
. exhaust system and reduces the likelihood that flames could develop. - Although the
~ system operating procedure contained a note to limit the time the EDG is run
.
unloaded, the operators interviewed by the inspectors were unaware of what time '
frame was meant by this note. In the case of the E2 EDG event, the diesel was run
- for about one hour unloaded, and the system manager noted that this unloaded run
.
m time was too long and was a significant contributor to the leakage and flames.
I
-
y.~
+ - e e * Following the E2 event the' system operating procedure was changed to specify a maximum unloaded run time of 15 minutes. The procedure was changed again l
following the E1 event to limit this time frame to 10 minutes.
.
NRC concerns about excessive oil leakage on EDGs were documented in NRC
' Inspection Report 50-277(278)/97-04, dated July 24,1997. An oil reduction task force was formed in the fall of 1997 and discussed strategies to address the oil W
leaks. 'One of the initiatives was to change the EDG testing time from backshifts to dayshifts to allow maintenance and engineering personnel to observe the oil leakage
'f during normal working hours. However, this practice was not implemented until May 1998.
Some other oil reduction practices were implemented during a maintenance period L
on E2 that began during the week of May 11,1998. Maintenance technicians
'
replaced the gaskets betwean the ring catcher and exhaust manifold and performed checks for flange flatness. The inspectors observed testing on E2 'during the week
,
?'
of June 15,1998, and noted less leakage and smoking than that observed on E1.
Engineering personnel assessed the impact of the flames on E1 and E2.and concluded that since they were contained within a small area and were self-extinguishing, there-was no significant safety concern and EDG operability was not
<
-
.
affected. However, plant management stated that this condition was not acceptable and plant personnel would continue to pursue oil leak reduction strategies to minimize the likelihood of recurrence.
I f
_ _ _
.-
_ _ - _ _ _, - _ _ _ _ _ _ _
_ _ _ - _ - _ _ - _ _ _ _ _ - _ - _ _
.
.
6.
Conclusions
~Some emergency diesel generator (EDG) oil leak reduction strategies.were not well-
,,
implemented or well-communicated to operations personnel. These factors contributed to oilleaks and flames observed on the E2 and El EDG exhaust manifolds in May and June,1998, respectively. Engineering personnel concluded that EDG operability was not affected by the flames, but they are pursuing additional leak reduction initiatives.
E2.2 Scram Solenoid Pilot Valve Degradation a.
Inspection Scope (37551)
The inspectors reviewed engineering personnel actions in response to two instances of degraded conditions affecting scram solenoid pilot valves (SSPVs) on Unit 2.
b.
Observations and Findinas
On April 28,1998, operators found the SSPV for hydraulic control unit (HCU) 42..
39 continuously venting following half-scram testing. On May 7,1998, operators found the SSPV for HCU 30-23 venting from its exhaust port while the valve was -
normally energized. In both instances, the SSPVs were replaced expeditiously and the soft parts were sent to the corporate laboratory for analysis. Preliminary results L
' + from the laboratory indicated hardening and shrinkage of some of the soft parts.
I Based on these two instances and the laboratory findings, reactor engineering j
personnel recommended that the station promptly plan an or) 'ine HCU maintenance j
campaign to replace all 44 older style (pre-1991) SSPVs on Unit 2. The on-line maintenance effort was planned to begin on June 5,1998, allowing for thorough j
planning of the work.
j On June 1,1998, a third SSPV failure occurred. During half scram testing, control rod 50-23 scrammed full-in. Operators inspected the SSPV and found a significant amount of air venting off. This scram solenoid valve was immediately replaced.
I Nuclear Maintenance Division personnel replaced all older style SSPVs during the
{
period June 5-10,1998. This maintenance activity is discussed in Section M1.2.
l Additional analyses of SSPV Buna-N soft parts were performed by a vendor in May and early June. The analyses indicated that the Buna-N parts had sustained significant aging as evidenced by their hardened state, which led to the observed exhaust port leakage.
j I
L mThe inspectors noted that reactor engineering personnel provided sound engineering j
support by ensuring timely resolution of the SSPV problems. Based on the j
additional analyses of the Buna-N parts and the failure on June 1,1998, reactor
)
engineers made a prudent recommendation to promptly replace all affected SSPVs.
i l
_ - _ _ _ _ - _ - _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
_ _ _ _
- _ _ _ _ _ _ _ _ _ _ _ _ _ _
.
..
c.
Conclusions
-
- Reactor engineering personnel provided good engineering support of plant operations by ensuring timely analysis and effective resolution of degraded j
conditions on Unit 2 scram solenoid pilot valves.
E2.3 (Closed) Valve Operation Test and Evaluation System 10 CFR21 Notification.
" Potential Thrust Limit Exceedance" a.
jnsoection Scoce (92903)
The inspectors reviewed the actions taken to address the 10CFR21 Notification,
" Potential Thrust Limit Exceedance" issued on October 2,1992, by Liberty Technologies, b.
Observations and Findinos:
Liberty Technologies issued a 10CFR21 Notification for two potential problems that can affect certain thrust values previously obtained with the Valve Operation Test _.
and Evaluation System (VOTES). These problems involved the possible use of improper stem material constants and the failure to account for a torque effect when the calibrator is placed on the threaded portion of a small diameter high-lead threaded valve stem. Both problems caused the indicated thrust to be less than the
true thrust during VOTES testingc These problems could cause allowable thrust
- '
- -* *'
limits to be exceeded. Liberty Technologies developed quantitative corrections that
. could be applied to data to determine the true thrust valves.
The inspectors reviewed the actions taken by PECO to incorporate the information provided by this 10CFR21 Notification into their Generic Letter 89-10, Motor Operated Valve (MOV) program. No concerns were identified during these reviews which were documented in inspection reports 50-277(278)/94-12and 50-277(278)/97-07.
c.
Conclusions The licensee appropriately incorporated the information from the 10CFR21 Notification, " Pot..mtial Thrust Limit Exceedance" into the Generic Letter 89-10, Motor Operated Valve program, i
u_
.. _ _ _ _... _. _ _.
_ _ _ _ _ _ _ _. -. _
.-. _ _ _ _ _ _ _ _ _ _ _ _ _. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ______,
.
)I~
l JV. Plant Support i
l R1 Radiological Protection ana Chemistry (RP&C) Controls
i R 1.1 Locked Hioh Radiation Door and Postino inspections Durina Plant Tours a.
Insoection Scoce (71750)
J The inspectors toured the Unit 2 and 3 turbine and reactor buildiregs during the i
inspy on period and inspected high radiation doors to ensure that they were properly posted and locked, if required. The inspectors also reviewed two surveillance reports that documented inspections by Quality Assurance personnel of controls for radiological instrumentation and locked high radiation doors, radiological postings, and plant housekeeping, b.
Observations and Findinos Quality Assurance personnel verified that all plant administrative requirements were
. met for the high radiation doors, nuclear instrumentation, postings and housekeeping in areas observed. Radiation Protection instrumentation was
-
observed to be within the calibration due dates and appropriately source checked.
,
The housekeeping in both units was generally good with floors clean and free of debris, clothing and trash receptacles in designated areas and appropriately
.
identified, and equipment properly stored.
q The ins.nectors tested approximately 20 high radiation doors that were required to be locked. The inspectors also observed numerous radiological postings throughout the Unit 2 and 3 turbine and reactor buildings. All high radiation doors required to be locked were found locked. No deficiencies were noted with the radiological I
postings. All locked high radiation doors tested and postings observed met the requirements of technical specification 5.7. No concerns were identified by the
-
inspectors.
c.
Conclusions
,
Locked high radiation doors and postings in the Unit 2 and 3 turbine and reactor buildings, observed during this inspection period, were adequately maintained per technical specification and plant administrative requirements.
P1 Conduct of Emergency Preparedness Activities P1.1 Observation of Emeroency Preparedness Drill a.
Inspection Scone (71750)
The inspectors observed portions of a Peach Bottom emergency preparedness mini-
drill conducted on June 15,1998. The inspectors observed the activities in the Technical Support Center (TSC), monitored communications between the TSC and
'
i
. -.
..
.
l.
' the Operations Support Center (OSC) and the Emergency Operations Facility (EOF),
and attended the post-drill critique.
l b.
Observations and Findinas The inspectors observed that the TSC facility was maintained with communications l
and plant monitoring equipment, as well as procedures, documentation, and drawings necessary to perform the functions described in PECO's Nuclear j-Emergency Plan. The TSC was activated promptly with adequate staffing. The
'
emergency director and team leaders demonstrated good oversight of plant -
i conditions and appropriately classified the emergency consistent with Emergency
' --
Response Procedure ERP-101, " Classification of Emergencies," Revision 20.
l Supporting members of the TSC adequately maintained status boards and monitoring equipment and investigated technicalissues as directed by the team leaders. Communications both within the TSC and between the TSC and other j
- emergency response facilities were generally good.
- ' Licensee personnel conducted a post-drill critique and concluded that the objectives
-
were met. The drill evaluators identified. minor areas for improvement in
- communications and coordination between the TSC, OSC, ar:d EOF.
The inspectors identified no significant issues during the drill or critique. The L,
.t-- ;,, inspectors noted.that the minbdrill provided a useful training opportunity for some
.~
-
j personnel who were relatively in0xperienced in their emergency response positions.
-
c.
Conclusions
!
Licensee personnel demonstrated good performance during an emergency preparedness mini-drill conducted on June 15,1998. The drill was adequately controlled, and the post-drill critique was good.
S1 Conduct of Security and Safeguards Activities S1.1 Sgcurity Force Personnel Compensatory Actions Durina Motion Detection System Roosirs
- a.
Insoection Scope (71750)
The inspectors conducted routine observations of security force activities while
.
repairs were being made to portions of the protected area (PA) boundary motion detection system.
'
b.
Observations and Findinas During tours of the site PA barriers, the inspectors observed ongoing maintenance g
on the motion detection system along parts of _the security fence. Security was L
present and observing the activity, ensuring access control and maintaining j
NAA/ SOP-7. logs to the PA. Security officers present were following site
_-_ _________ _ _- - _ _ __
e
.
pMedures for this type activity and maintaining access control. No issues or cos.eerns were identified by the inspectors.
c.
Conclusions Security personnel adequately maintained control of the protected area boundary dunng repairs to the boundary motion detection system.
S2 Status of Security Facilities and Equipment S2.'1 General Intearity of Protected Area (PA) Barriers a.
Inspection Scoce (71750)
The inspectors toured the site perimeter to observe the condition of the PA barriers and the isolation zones around the PA barriers.
b.
Observations and Findinas The inspectors observed that tha PA barrier had no openings and was not damaged or degraded. The barrier did not show any signs of erosion at the base and there was no substantial rock debris accumulation against any of the fences. The isolation zones were free of objects and permitted observation by the Central Alarm
.
- Station and Secondary Alarm Station operators and security force members of any
-
unauthorized activities.
c.
Conclusions The licensee was appropriately maintaining the barriers and isolation zones around
'
the protected area.
V. Manaaement Maetinas I
X1 Exit Meeting Summary The inspectors presented the inspection results at the conclusion of the inspection at an exit meeting on June 24,1998. The licensee acknowledged the findings presented. No proprietary information was identified by the licensee.
X2 Predecisional Enforcement Conference -
On May 21,1998, the NRC conducted a predecisional enforcement conference with
.PECO representatives in the NRC Region 1 Office. The conference.was held to discuss apparent violations that were documented in NRC Inspection Reports 50-l 277/98-03 and 50-278/98-05. These apparent violations were related to concerns with the corrective actions for the degraded 2A reactor feedwater pump turbine
-
- _ _ _
_ _ _ - _ - _ _ _ _ _ _ - _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ - _ - _ _ _ _ _ _ _ _ - _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ - - _ _ _ _ - _ - _ _ _ _ _ _ _ _ _ _ _ _.
.
.
high level trip function and inoperability of the 3A core spray pump due to foreign material. Copies of the licensee's presentation materials are Attachment 2.
X3 Review of Updated Final Safety Analysis Report (UFSAR) Commitments A discovery of a licensee operating their facility in a manner contrary to the Updated Final Safety Analysis Report (UFSAR) description highlighted the need for a special focused review that compares plant practices, procedures and/or parameters to the UFSAR descriptions. While performing the inspections discussed in this report, the inspectors reviewed the applicable portions of the UFSAR that related to the areas inspected. The inspectors verified that the UFSAR wording was consistent with the observed plant practices, procedures and/or parameters.
>
)
<
I I.
l l
i
-
-
-__ _
_ - _.
- - _ - _ -
_ _ _ - - _ - _
- _ _ - _ _ _ - - _ _ _ _ _ _ _ - _ _ _ - _ _ _ _ _ _ - _ _ _ _ _
.
i i
.
ATTACHMENT 1 LIST OF ACRONYMS USED L
.AO
. abnormal operating AR.
action request CAD containment atmosphere dilution DBD design basis document ECCS emergency core cooling sistem EDG emergency diesel generator eel escalated enforcement item EOF Emergency Operations Facility FME foreign material exclusion -
GP general procedure HCU hydraulic control unit HPCI high pressure coolant injection HPSW high pressure service water IFl inspector followup item LCO limiting condition for operation LER licensee event report MCREV main control room emergency ventilation MOV motor operated valve NCR non-conformance report NOTICE '
notice of violation OSC-'
Operations Support Facility PA protected area PECO Peco Energy PEP performance enhancement program PDR public document room reacto' feed pump turbine RFPT r
RHR residual heat removal RPS reactor protection system RPV reactor pressure vessel
- SO system operating SSPV scram solenoid pilot valve SRV safety relief valve SR-surveillance requirement ST surveillance test TS technical specification TSA technical specification action TSC Technical Support Center URI unresolved item UFSAR updated final safety analysis report VOTES valve operation test and evaluation system l
_. _ _ _ _ _ _ _ _. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
-_
_ _ - _ _ _ - _ - - _ - _ _ - _ - _ _ - _ _ _ _ _ - _ _ - _ _ - _ _ _ _ _ _ _ _ - - - _ _ _ _ _. _ _ - _
_ _ - _ _ _ _ _ _ - _ _ _ - _ _ _ _ _ - _ _ _ _ - _ - -
.
-
Attachment 1
INSPECTION PROCEDURES USED IP 37551:
Onsite Engineering Observations IP 61726:
Surveillance Observations IP 62707:
Maintenance Observations
!
IP 71707:
Plant Operations l
lP 71750:
Plant Support Observations IP 92903:
Followup - Engineering ITEMS OPENED, CLOSED, AND DISCUSSED
,
'
Opened 50-277/98-06-01 VIO Failure to Energize Trip Relay for Main Control Room Emergency 50-278/98-06-01 VIO Failure to Energize Trip Relay for Main Control Room Emergency 50-277/98-06-02 VIO - Plant Status Control Corrective Action 50-278/98-06-03 VIO Unit 3 Recirculation Loop Mismatch Following the Runback of the 3'A' Recirculation Pump 50-277/98-06-04 VIO Failure to Perform Surveillance Required for First Recirculation Pump Start 50-278/98-06-04 VIO Failure to Perform Surveillance Required for First Recirculation Pump Start Closed 50-277/98-02-01 URI Plant Status Control issues 50-278/98-02-01 URI Plant Status Control issuca 50-277/2-98-01 LER Failure to Perform Surveillance Required for First Recirculation Pump Start 50-278/2-98-01 LER Failure to Perform Surveillance Required for First Recirculation Pump Start 50-277/2-98-03 LER Failure to Meet the TS for Main Control Room Emergency Ventilation 50-278/2-98-03 LER Failure to Meet the TS for Main Control Room Emergency Ventilation 10 CFR21 P21 Potential Thrust Limit Exceedance Discussed 50-277/97-07-03 URI
. Unit 2 Cooldown Monitoring Following the November 9 Reactor Scram
-
_ _ - _ _ - _ _ _ _ -
- _ _ _ _ _ _ _ _ _ _
,
pmn e
uo c
i n
P s
e r
u e
yr fn at o
.
r n
C pI tn8 Sl e9 a
m9
_
ei e1 r
r c
,
r o
e o2
t f
Ca n
y E
a A M aM l
no
n is g
c i
Si e
P e
D
-
r e
Ao rP BF P
,
,,
!
=
_
,r t
e n
h e
g m
a e
s ll g
n a
a o
G ni w
t a
a e
s
_
P Ml n
.
i a
o v
l r
o s
-
_ d e
e i
n
_
n a
gV e
t i
v a
o h
a t
i nO ct e
c t
c
_
g u
i i
e pain l
_ A MO r
n e
f a
o s me
-
o pit uI v
- r p
a a
cit e
r c
c oAiCi
_
t r
n t
f e
e e
c r
Cit n
r d
o
_
s e
e r
R o
o e
ot i
rP DN MRGCO
.
.
.
.
.
,e
.
.
a
.
.
e
r n
s u
in n
s eo o
s l
i A
bt i
a t
a c a
ol r
i l
t a
ef i o
sr pc i
i ee oe V
r t
np u
aI S
dM pl tn ec n mac e
og ui r
ri ePn a
Pr h
p o
yc E
F ae p
r Mf pT A
F oSf o
f n
e C
o o r
n R
es oo i
i uCt r
X ul a
c'
l laxAo i
t FE 3V i
a
'
r
.
.
,;
_
ng n
i ee e
i r
f h
r
,
l b
e o
ToF a
vsn y
r ne ao f
e e
ewWi t h p
i s
t eo c
nt o
n BHn u
ef n
o ad d
o I
s o
n s
i i
.l s
a ePr o o a
t t
W u
Hv i
n t
iEI d
a t
mcMe et p
lc afeFh s n se m
n rf t
u o
gEtct Mm P
i o
n r
e e e
C PBov elp y
e j
r a
O EoPrl ems r
e r
Mt PiaWi o
p l
d e
r er S
C F
eh oe sh t
t t
t et n
e E
r t
,t ai o
r ua t
o P
Ot enMinC r
i s se n
C ain uml a n
eiA c
t l
laoCf g
r i
r
r f
ol e
i emsu e
pb t
e r
po ah i
veh s o r
ODTInF OPMT
_
e e
e
y
,
d d
.
e e
u mn l
r d
g i
n e
ol f
a n
h r
i o
r i
c e
r t
t S
P e
u n i
De t
a a
c nM d
m a
g i
ee f
g i
n o
n cc t
t r
oig ua i
i l
d p
n u
h e
o e e
D s
r r
e o
t R
d
F n
l I
b r
I uf r
e
/
o on m
o at
i r
e n e ca e
.
/
cT or et r l
b n etiAtoS d
a v
a o
el r
r c
sf ePn l
i r
t i
i o
l P
ce nil t e
git ev e
n p
nc t
r t
eu x
e miu l
DSE d
SS I
I me oe e
.
.
_ _,. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
e
.
.
,, !. '.,
- (7
'
,t e
.-
,,
-..., '
-
'
y
' '
.
,
@
,
?
ys.
h (*
y f. '
~-
'
S
.
,s
.
v
..
.
t..,
,
?,s,,
,
'
. " '.
rL
'
i x _~
.
t
,
.
W,
'
b.
"
(:n
.
e-
'
'
i$
wW.-
t
?lc.,
+
k*-
-
.", D'
'.'p.
g.- -
-
@.
-
- @... g L.
he
'
r si., s'
.~
';
[ >%s.
' ' -
'
tf
..
>.
.
. q.
'
h.
~
, ). Ih -g < -
J;.
s,. : ~,.
S
'
4[$..
3:q. N;fQ. ~... L'_:y; y '.
'
..
y
>.;
y.e.
c t,
..g
.. <.
-
-
sr
-
O
.
- .
e, a.e
.
.
>
' '(q '.
f.g.
. a '.
w
Q
';', '
Gw&y
..
.
f}f,.
. :;
%- Q L,. \\ : _
..
'. r.
p
%~
fd,[f-. ; I ~
h
_
.
.,
(
-
'
C
[.
[d j-
' f'
M'L.s'
. i l., ;
'
7[
...
t
. -
j r
. ;< ;
h
.. -
'
&
Y
.f:
' ,W..
l l;U,
, ?ik O
...
.
.
.-
~
.
....,
-
.
.
.
.. - ~ -
%,y h
!. k g. -.
- sw
, ;;
\\
T[gh.'.si
..
-
.a q
~m..
.
]%s.}f,
.
'. -
..
h-
' ' c.
'
I l,' s,.u.
Ah. *.
.j h.
',
'
.a..
-
...g. g p.,. r.... -
- e;.
t
,..
'
,
..
,
)
i _
- g.'
d' k,. f2
- ' ' h
,
,
,,,
.. - -
i -
.
,
N W.M. 3..4. ? ' ',,.--
'
> !J ? i
'
?
7 7e7
-
g*:
?..
n-
.
'
'.. %.j. _ W Q V ~;* '
';,[.l
'
'
.
.
I.
'
eM
.
j..
..
t
,$. p,
p 3 : ;--.i*
-
.e
.
- q y' s.C..
d, s', V
.D
.,*
, r.w ;..
-
f.
Yb
- .
..
-
4l J.'& ' ;. ~.
. y, s' T'
-
.
-
t.&.
4ll!!
-
-
. rp l
.
y U:( '-
M,.,'
.
gun re.
nr
>e
. af.
.
u
.
.
s
.
.
..
.
.
9,
"
[,
-
k.n.g
-
,s
-
.
$
.l,
& ;+L.c/,,
- .e..
s
.$.',
,
- ~;
w:
-
,.
' k.
.
().
,
. g:-
,y,5
.
,
m h?.
.
.
...
.
.../,' [ TR ' *
..
-
/
'
- n
... -
N
.L
- . :l' '$.
.
t
'
(.
kN M,
l.
p.-
~. r
.
i
- -
p4'.
' ;,-
fE 2.
- -
,.:u, y...
\\'
-
,
.
. -
<
.
,
-..,..
(.... 'F' ' ('.k *k yh. '"/, *h,.' -
.
.
'
.
L
.
,f
...,.
,
,
,
- - _ _
L
,
.
la
.
ir
.
.
e
.
e
.
t
.
.
-
a
-
s
.
s
'
M.
n
.
g
.
ie
.
ro F
dn
.e
.
u
.
o
.
.
F
.
'~
.
-
s.
- -
.
.*. *
A
.
,.
?-
f
.
.
.
n
. %.
,
'
.
l!
.
w n
,
3,,
.
,
w.
wt
.
L '
-
_
s.
.
_
.
y,
_
.
.
.
l
_!i i D; :,. ! j. l.s; i
iiiil. f.l,i i i. j.
.
.
_
en
.
-
e s
-
._
a.
.,
i e
,
_
_
g
-
.
- $
, -
v m
,._ '
e!
,
,
s
,
,
j g,
i
_
\\
l
N
~ ~,
f;
-
r e
-
\\
_
.
-
_
.
. _,
.
/
"
-
_
-
,
.
e
.
N
-
c
_.
/
8,
.
'
'
.
vn
' 1 p
.
4l
,
N 2/, x.
.
-
g gM
'..'.
i
.
N'
'
.
-
,
.
a g'
.T*
.
s
.
'
Or
,
s
.
1'
_
\\ M r
~
._
'
-
i'
'
-
'
=
A
[
-
_
N f.
'
_
4_
.
-, D
/,.
.
{.
-
s
. p k
n
.
,
f
-
r 4. -
.a
/ i-
_.
'g\\
,
,
'
\\
-
/
oe
,
- i.
i:j.
ii
.i
.
i n
A. _
i ii
1 f
.
'
,
u
/
'
. \\
s- '__
-
.'
/, /
Az
!
.
=
'
.
d'
t i
,
..
%
~
-
N
.
a a (
s
.
'
,
.
s
._
.
<,
7,l g
,.
.
"
_
r
- yk s
-
-
ct
- yu3
.
.
g
'.
-
-
._
-
-
.
.,
9(
iS
.
c
!.
-
..s '
.
,
.
'
_
-
/n
.
/
a f
,
.
W.
f s
,.,,'
,
4,
.
.
y
.
/
.
i
'
_
M g.f'
s <
.
u N:
' 4
-
.
_
dr
_
N
.
,
c g *
.
' D-.
s
./,
oo
,
n.
'
.
-
-
T r-M
,
t
!I
!!l,ii-!l.tlj
,il
-
.
.
.
r_
e, =,
.
. =
.
~
n.
.
t
.
,
w
.
.
.
4n
,
.
,
.
I13\\
lll
_
.
I g,
.
f
.l
-[
- l
. p
.
v-
-
-
,-
-
.
R
"
..
E
'
'
O N
I n
n =.
A a
Mn w
e R
'%'
-
~
T
E
.
m
S L
=
U ey RD il HO vb RM rm ee vss OA E
r P
IP ne o.m
"
2 ar p
it t as E
M c1 O
C i
a N
f c
M w
ii O
dp W
O y
T oT A
C M
RT E8 N Q tAp
'
- * -
" >.
s GO tNR pG e
.
.
.
.
_=.
_
eh.*
w
"
.
i
\\
v
~
W
. 4
}
,
r f.
'.
.
.,
e
~
h
.
e
.
-
v
.
,
k
.
- y
-
l O
'
i l
f
.
s
~+
M
.
<
n b
.
>..
&
.
,
.
o
.L (
y..
i g
.
t
.
.
.
-
.
a
-
cn
-
.
ii f
a
.
ir
.
t dS
.
m
.
o
.
b M
-
.
i
,
'
wn
[g
~
-
o nO ei Q
A
. h
.
.,
t i
a vr r
u
&..
'
e
'
i vf nr
.
%.
g, h i
n
,4
_
Oo
-
,
C
,
, &.
.
oe g,
'
,~g;
',
n i
i t
a y
ar
-.
- t cS
.
i
'
[
.
'
i
-.
-
f l
.
M,
,
-
'
ia
.
dn
_
r oe
'
t MI
,
n
-
S
.
3'
.
-
-
'
l
-
~;
?,
d n
~
,
Q. e
' a---
-
'
,,.
. Q
E f-
-
.
'
., '.
_w w
-
-
e
.
~
-
-
i
.
'
.
-
' -
v
,
klge
-
.
e
}'
.-.
.
.
.
.
..,
.
,
r
,
, q
.
(^
'
eg
_.
,
_
p
_
n I
.
,
n
_
'
_
vi
-
,
.
.
n
_
.
,
'
Oe
_
-
_
,
=
.
C.
r_
-
'
_
,
.-
.
a,
'
.
nO
.
g*
_
-
a
.N
'
r i
e
"
%__
.
.
e
~
.
,-.
-,'
,
w 3 '
~
.
_
,
or
.
t W n4 fS
.
w. ~,
%)
-
'
.%
..
.
_
.
.
t n
.
b _
._
_
-
-
.
- ,
-
.
ai
.
.
.
a
.
cr
.
~
-
-
_
i t
-,
,
.
.
.,'
-
..
i
.
~
_~
.
i d
_
-
_
~
.
o
.
.
M
'-
.
-
-
.
-
%
..
=
-
.
_
,
7.
,
,
,
r
.
,.
$ m sT f; i
.J t
'
.
s
- .
gr
'
$
.
- .
'
$ e.
_ *
V.
.
p N* h e.
.
,
-
e
' pW,
-
.
\\
yp g 5%N2Eh 3 4i.
h
-
t'
gd
- me
,o j,p@
.
?'
.,
m<
i
-
wg{g'ag5 qe@
g n
sM
ymj s
w;y
.; 7 a
- -
l cg(gw;
{' ygW p
o
-
g, ywgl
-
.
y
y r
r ws. }ug}
9 @S E
t r
4 rS n
-
,
(h wIq wC r
-
y
y+g];eg o
,
}
.
d v M e g ng f
i
,u. g j
r l
e y
el
-
r k
r a
X gM g r
e c
o iFh g d t n
i a
D Cg a
k o
i M
la L
ve
-
r w
g*
j
.
a
.
s4mj$h t
C i
rt
-
y a
s
.
r ly a
d u
ebbmN eM e
r e
l r
a g oT w
A
-
ir o8e" re e vn g
L os m
t DsA" n
aMb ig u
-
nk g
Va c a
S Oi t
e e S
Ql h
-
-
'
_
e
S r
U
o
+
lE R
_
-
O I*
n F
B T
/
d R
s e E
of r
-
E v
w
-
r,, an e
D v o w
" q u\\
e h
o om I
Ql F
S p {'.g
-
W s
C Ce i
i R
N a
-
?
t n
-
I
%
p [, a e
'
, k>g+
'
o
?
k ai
.
p m
lu u
'
.
..
t
_.
e nnl
.
-
t u
d t
=
i c
a
.
e ol tnwh o
"
/
e i
n
.
siMMB
.
v rjp o
t
.
eI ai mE
.
@h p@y sM M
h-
.
'
eM
.
4 p wc f
e
.
RFC yWw gj'
..
1M
s j
.
g i
e iga n
n M
_
s
.-
I dp F k I
r E s e c
r c a e g:q p D +O.h.
l r e Me a
t r
eh vl S
3*
l E
F P
A v
e iC
.
o e
oR
.
e C
iS
.
t a
,.
g;-
M
,
ts
_
, e. gg,,'
y S nF Mi
?
I
.[
,t mc Y
r
l m, u. m e le
"
a r,.W.M
.
y
1
_
r
-
u t
gV wd y
y a
i it oo t
.j2 LM
- e n
e n
n
-
_
v o-p t
r
_" G
~
.i mO
-
.
o/
,
h d u u
O O
.
d it t
l n
r t
s e
ai E r l
ci e o o
,&
S el ir e
w j,t p
p Mv g
a c
R
_
p p
F ot a
e s
b F Cn L
a
/
'
I F
@E
^
_
~
_
l
gE p
T XXXXX y
SYS
w S
e U
i RXXXXXXXXX v
O T
re s
vl o Ot r n
no s
lo oC r
k t
s n
c g t n
n O
o e
i h
t E
k C
i a
R c
C nf d
aM a
n S
e E
n e
i l
r e
h s
a yC M
c P
B r
t F
e e
b t
F Eb A
l h
nf v
C i
s g
a o
f Mo o
t g
ul k
C e
e J
n F
-
i l
s i
e o
o g
oC e
c d k z
P P
L o
r c
e e
c z
c r
d h
k nh e
o i
L l
f Tl a Cla h N i
l l
c d o
o p
a o
C l
t t
t o
mwiE e
n r
r l
n s
p a
n n
r i
n n
o t
e a
v MI M
S g
o o D wC aii t r
F y
e a
t n
CCf S
u r
r l
c in o
s S
r oC t
e iE E
E u
e e
t j
MMe r
v c V n
MoAi o
a s
r r
a e
P TF F
UF TQDF QP
,.
w
.
.
.
.
.
.
.
.
.
.
.
3
,
w n
y
.
ol
.
s e
l e
odtea h
r t
t t
su n n uq s e o ce i
mCod t
.
neleE FA yt vpMlio e
emf r N s
a
.
u sI t
s id cmdl h
a Tneino r
r C
ajoPat n
f r
s opPml o
t o
o o
el earC o
se t
ve nM uvi R
aesToF CDntcC e
m ejMe oh tot eoFt o er r
s Rr pPsy um uS el er i
h aoh on
,.
TFCTTo g
.
,
es U
t la o
ngi h
ene Nc r
ya ed i
ta t
E BiM l
i l
u r
i s
e b
e r
vBn at t
f o
ar g
t Hoi nA t
e u
c cr od a
d ao c e l
F ueF cm ora Ar la C
o es f
l s
sh a e ar n
r e
i u
ot l
r r ao eP a
t oeZ ay t
C nf r
l od An Mla o
c C e gi gi tu n s
nf r
i et iu c
i gl g
i t
e s a c e
t t
g t
x ip
,
BInSE RS y
.
.
,
e x vs e
al o pHr l
t s
d n
n ml o
o ouC Co it rCE a
o sM c
i seF l
ni tid p
ov e
m i
i t
t s
ac u
I cAc if c
e o i
i d
cF r
o ny e
Mal n
nw r
e eeo G
t r
h nr t
ia OaN M
m
2w ecn
.
o L
an l
e lA s
t n
n r
e i
s o
n a
o r
nit f
g M
e c
i o
s s
W iA n
e r
x o
t e
u Dj e
d c
v R
alp e
Ai s
r c
ep eMm t
n r
e n
er cmi1 o
aS vo u
a1 C
e r
n R
LG r
eP i
t t
C d
S
n s
d L
cd if S
nt n
f a
a ee oC o
o rt t
e s
sd rl C Ci t
n n n se op rE U
eio et Cm u e L
n o
or d
mt e
a C
Hu e
sc Sm sf sf s Pe i
i eik i
r s
d d
sAlp u e o
s o s
or aBm mFP MAMT PI 3y
.
.
.
.
>
-
e,d a
s n
n n
a g
gn i
o s
i s
i t
e h
c d
A o sDs E
c e
n M
a Mif s
oo g
e t
e F
r nv i
n r
v i
s a n
y Bu t
oi oi i
t t
i c
jc wci r
no aA ef a
s eT t
t e
i cr M
r i
i u
h r
e vd T
Ao r
c eo d
W C
o NRM aI s
n n
ec f
a vi s nEfoP e
ot l
l l
t l
i a
m oeM d
rl t
nE o
n t
r t
c t
nF em ni oMM o
r a
oat i
t F
n a k
Ce CM el d
r r
og r
l a slaI o
exd a
e Cr ce nt c mmn s
.
nlp n
nh a o P
e t
e a
pI a
cgit r
i u
h me r
d h
n od nod n
n er d
tu n aE BP AS FwECI
,
.
.
.
.
.
=:wT t
N s
r u
s E
l n
e J
s a
o M
n s
a r
i l
i i
i t
t e
t E
a c
n nd t
A Ct Ue e
a r
AS nl M
l e
L eo o
v n s cpt g
r P
ol i
c E
iolamn n
t t
o i
r r
R ct Po g
e C
g r
un nC i
o RSoif e
R C
E yC e
oB l
r l
-
N E
Bnl A
t l
-
sMitoW i
I
A l
f u
l i
o
RdF Wla P
T nt s
ys l
n sa a
iU t
I d
S r
t esl k
e i
es o
r ebh r
N amv nan ac M
O me oI gotn o
r Th cCt nZ uE e
I a
u cl r
igE or C
npE o
ce tu F
U eeMi aM cf r
t
,SBRFP SF AA
.
.
.
.
f
x s
t y
hg le i
r ve s
oic n
f t
I s cn efea A
s uf m
Qe g
qEr Ni v o
i n
nef t
r r
ec h
i
oe se n
g c
r n
eMP Up s i
T a
r e
o o e d
re t
l dt a
uP t
L n
ns u
o i
r r
o a
hi l
ea e
M sgv Se r
h l
h e
o aE mT t
c o n O
nTad a n Mns e e aeT a
e d
i mcmfyat a
l r
n c o eimer o
a t
f t
n ojB r
h s
o e
nye n
r o
ESIdAPt
=P.
.
s=
r E
o t
M a
F g
r i
u t
O i
n s
M o
n o
io t
t i
n at s
c o
iA e
e g
t c
g it s
e n
n a
e v
a a
l vi ch t
o n
cfiC i
I ei n V
e r
e r
g v
v oi d
i i
p e
sC S
s n
n m
i y
f e
et e
i t
t eh u
nh af em P
e ei a
r d
r a
d p
eS pr I
S mmw mg f
o C
e o
l or mo SCIL CP A g.
e eeeeo 3n
$
-
e ena la cn en n
nei e.
b o
r emT m
r u
p isf ic n
p u
e o
mP u
y
DC p
a
r
-
d p
,
t e
S
1 e
n e
y r
a e
F er M
P m
r o
o C
S t
e.
c P
c a A r
e
A o
,-
R Bf A
n=2 P
E x=
g
==
e mr
_
_
e t
i
_
S re
_
,
r g
o a
_
tc n
_m e
a M
r iD tn n
,
e, t
n a
l o
P o
s
,
n r
i h g e
_
s o n n
r i
J r s
a e
W n e u
~
o n di k
c a
on a
r g r
GE M
sd i
n e
e De T
g PA tro F
,
p n
e o
d R
s e
R
.
is e
e s
n s
c n
r r
n o
u u
a n
A
_.
t t
c o
a i
e c
a s
t c
i L
c e
e i
i
,
p F
D f
e A
,
s in p
y v
e g
e n
y t
'e v
i i
i r
r I
l WiS m_
a T
b i
t la c
y mT a
r t
e t
ice mP e
a r
e r
f
=
p u
F p
h o
a m,
S S
R O
W C
S
_
e e
e e
e e
_
_
r n
o f
o to i
et n
a lb c
e a
af r
i r
e i
e c nm pe w
o p
o s
i, nS n
t s
d o
i nu-saei l
t el acn c
i i r
c.
wna a
a n
ht n e
e ci v
po ru eai pC t
Tmc ta A/
edt e
r n
f no r
Co.
pan oc i r
,s i
Rt t
s a a v t
e T h Xla-whi t
P n
t s
o.
F o e n
cd Rmn e e i.
h Vg Alal u
ad e
g i
r
r p
cp l
g e
e vmnm x
h eo o o
=
TscCc
.r e
e
-
-
-
e g
e l
v ba i
g s
r n
e s
g, e
p h
o n
,_
e s
o.
p a
r is~
m w
u o
T c
P lc F
to n
n R
o s
A C+
a
w e
O g
h t
n C..
o c
i
,
t t
E o
e h
p P
=.
s s
e o
.
r
.
l g
b t
u e
r o
g, r
n
,_
T I
e o
m_
te
.
d s
r e
e r
hc o
=
t t
a n
l m_
n o
ui s
tc s
e s
a r
e e
u
-.
n a
l t-
,
m r
t u
r s
a a
as e
e p
ek y
c v
l e
r o
l s
Fo b
s m
W m
s i
lat e
pt e
o r
n n
u i
i I
s g
n q
r s
i Tw A
s c
r s
a e
o et e
H p
r c
e e
iu T
n v
v r
u c
i t
r ol a
l T
a ai P
.
ct v
v f
e e
of n
s p
p F
m wtn a
u u
e mm i
R l
r y
e a
eid d
l g
l t
s wp p
iv r
e r
o o
-
i i
r r
e pN TT m
vOIma aa g
oe
-
d d v ir e n ip d a w
wl p
d a e
e n r
t t
p pl T
ini ni e
ee ad T
iad p
px Pat e Pat e st st iR F h ul F h ul i
r r
u Ris Ris u t
t t
t h
s s
w ae e w ae
_
T Tg e
.
-
P Pi hd, r
-
h hd
,r
-
-
T F
F(
t el e t
a el e
-
v a
1.
C Rl Rl ev n ev n v
,
E a
a a
a en en cogh c o gh inmi nmi
-
P h is g
h g s
s i
s d
s d
_
S t
t is e yl e yl nMR l
led led lel ao lel ao O
vr u
-
-
-
-
vr u
-
e at at e
e T
b b n b n b vnw b vnw E
a ai ai a
el a
el o2 r
r a r a r p a
r p a
E R
e et et em m n em m n p
ps ps D-N u
u pu og pu og i
D I
O Os Os Od mi Od mi s
s s
D d'
. pd p
e s
=
.
eie r
/ef
e l
t e
t lel v o i
l r
r t
a
'
b b
c
at b v hn sp g t
u k'
ne a e f
aip a
ai n r n e
a r
=
rb r
e et s e
wt it o
e ot m r
o h n p
i pt pou m
dt e odde c
o oh s nd owv in ne v ie ins p
it r
si e
e ut o
- o cm t
r n
at m
iper g
- i
- n s
ym ye li h
s laeit e
w r pn f
r e
o t e i
y ph r
r t f gh r s t
r t ss er eiot nc eiui i
o u
i erl l
v e vmcd e
ut sa oo s vs g
t or o e c
n ge D
I di e E
cd cee e n
nial n
cpeuv a
c s
ist vy is s
nl a mt e dd esa M
i p di al v
f t l
f
- v e l
- d vm wn r
e T
f
-
d b
elao o
ip e
o b
o mlsd m mie a
mt d
d a E
lb eet e eo v r7 l
epr mb ml d r
cvf e
e l
ae
. e e
u y
H a
myp ml n a
t r
e u
t lae it oleh p/ b pmg i oe sp T
e i
b c
o1 os t
T p
e c t r o o n/r uun t rt t
t A
O ADa APst I
nei 4Psdf APirt r
i l
l i
a y
y y
y y
u r
r r
r r
i n
a la a
a a
a a
t n
t t
t t
n n
n n
n g
b m
e lai e
e e
e m
us m
m m
m (
t d
o c
l o
o o
o e
a ts ia al a al a al a en m
av m
m m
m e
nl r
d a
x a
a t
h nh h nh s
h nhR iwiw iwiw
\\
ul t
t t
t t
t gi gi gi e
/e iws ip sa iwiwh s
g s
s n
h ih nip ip p
p p
pM ig n pi ir nip
t r
r r
t ot i t ot i r
or t
t t d t
r t
r
ot ot 5,
p )n otud e ot otl t
t t ut ut l
M p
io b en b
a b
a OM r
W ir t t d
ya O
i d
d u d
du ehe n h
T la eb leh at lehe n t
t
?
du iau e u aua aua lisl sl s i si l
a ia m
P y
U.,
M m
ah F pD s F pF m F pF m r
s
-
Y o
le u S
N Dp
.
.
.
.
.
.
.
.
.
.
.
_
1
,
/5
2
7
/
T
/
N
,
9
9
/
9
9
/
E
/
/
/
V
6
1
/
/
/
E
2
1
.
p i r et
-
b a
l
-
a n y-r i
e r
s a
pt ol m-u ss m
aer wl u
e pa S
b l
n i r at g
y i
t t
p ps i
e md li
.
c ue b
c pn a
a i
d a
r-p e
s t
e i r eu p
s t
f d
r O-e oa yt ce
_
a ac l
_
e en DRi s
eo
m r
a e
r i
g l ra o
r e
p
-
-
g g
d
,
d n
n
.
.
i e
n t
t
_
i s
o a
n.
e o
.
t s
h r
d d
s a
.
o r
e et e
e ht n
ab l
L" t
n y
e o
r u
e e
c mt ol o n s
s et e
r e
eis n
c r t c
s v n o
ui v
t oi a
i t
os a c ad n
'e gn g
e o
n i
e c t
i i
mh W-t s
s s
r t
e e a e
o gr v
-
nwvf nu n
r t
gi e
es i
r r
n a
u n r
pt u
se i
h ot o
e o
v oo W
_.
eod at t
zh eh d
s i
l n sd ei l
e d
ek
.
_
gl nl ob a
u n s
-
,
cu p
o ed eo xh
-
Rr E SWn
-
a t
-
-
e eeo L
ll
_
y f
s s
n i
tn u
a e
o i
r c
d e
o i
c n
i g
g l
o in h
l n
t k a c
i r gi s
s t
f e
m o
o r
o e
nr T
o s
i s
d h
m sf e e nl sgs r
ee c
r e
o zp n
a u
na s
,v l
f o.b n
r e s s
i l
u e
at i
n ri r
n, to p
ce eg e
i i
r pn n
e t
t pm xi nd s
c,d v
e sp el a
n b
o u r o i
Aa l
y h q us Pi e
oe
,t s
r w
r sf t
d s
r g
mo
.
e o
u yn ot eet vh d
ni t
r h e
n as f
n c
i m
po s a g
i i
t n
mg e o n
t
,
e n
o a qp s
c ur Ei v
ia ci ei g
a n
t d
p s
o i
a a
r t
r g
a r
h e
y g
h o cck a
n
-
ri gsi f t
t eit alein
-
s n e wt ov t a e
r e
n i
-
k c ge emb R
a r
id
-
Ci i
r t
n t
ac nm ae r
-
yn sACt t
r s
mai t p o
_
e yF C
r l
e eh p uos p
d
-
gh sHCi ru c
sle e r
gt n s n v u o
v t
d o
-
g gdne ee o e s c n
-
-
-
r s
n gt cBb Cdi ia P
er o a
a e
a r
Sp e
.
_
n
_
o i
tc g
u d
g e
e r
n g.
gn s
i eg i
t p
c~
m i
i r t
n i
c-l d
e a
m e
c l
l y
n a
if r
r a
a c
el c
i e
f p
s v
i od f
i r r
i i
n t
o T
n l
e c
eP g
t l
i s
g v
a s
F s
e e
eR y
i r
t l
v S
h s
e e
f g
e s
h a
t i
t h
s s
y i
u md l
t e
p a
e a
i l
i h
r c
t t
pin m
e,f f
p i
m n
i r
..
e o
r i
t a
n r
i t
T e
m si p
e P
b t
S~
e i
r r
n e ed r
F dl a
u t
u i
.gt T b RR b
h t
r s
a n
a geiP uP n
st s
i f
aF C
g y pt t
ggoMRiM e
n n
i l
s a c e
w a
e n
c v
auMiDAaE n
T e
e e
e e
e
V I
V
-
F O
L
-
T A
-
N
-
E V
-
P l
M O
U O
D
_
-
P A
Y P
- R I
f i
T I
_
R
-
-
T N
L O R
A M
E U
M N
O L V
A C
M
T S
P N
E I
A l
VE R
S V
T L
A A
V
P I
l R
T
>
~
~
T
N IO
~
-
Y E
L P
P
V PU r
L S
l A
IN L
R V
O I
R C
l D
D I
,
,
O A
N E
L O
-
_
-
S P
I A
,
R T
'
'