ML17342A380
| ML17342A380 | |
| Person / Time | |
|---|---|
| Site: | Turkey Point  |
| Issue date: | 01/31/1986 |
| From: | Woody C FLORIDA POWER & LIGHT CO. |
| To: | Grace J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| References | |
| L-86-29, NUDOCS 8602190110 | |
| Download: ML17342A380 (76) | |
See also: IR 05000250/1985040
Text
F LORIDA POWER 5 LIGHT COMPANY JAN S 1 198S L-86-29 Dr.3.Nelson Grace Regional Administrator, Region II 101 Marietta Street, N.W.Suite 2900 Atlanta, Georgia 30323 Dear Dr.Grace: Re: Turkey Point Units 3 R 0 Docket Nos.50-25 51 Ins ection Re or 85-00 Florida Power R Light Company (FPL)hereby responds to NRC Inspection
Report Nos.50-250/85-00
and 50-251/85-00.
As requested by NRC Region II's letter dated 3anuary 2, 1986 forwarding
the subject Inspection
Report, the attachment
to this response includes FPL's plans for corrective
action for each unresolved
and inspector followup item identified
in the Report, and describes the actions taken or planned to improve the effectiveness
of FPL's management
control systems for each such item.While we recognize the need for prompt actions to improve the effectiveness
of management
control systems for the items addressed in the subject Inspection
Report, we also request that NRC's deliberations
on the attached response take account of the particular
circumstances
under which the inspection
results were obtained.The subject Inspection
Report describes the results of a follow-up inspection
conducted within less than two months of NRC's intensive Safety Systems Functional
Inspection (SSFI), which was itself unprecedented
in its depth and approach.Without diminishing
the importance
of the matters raised in the subject Report, we believe that the results of any new inspection
program, along with an immediate expansion of that program through regional followup inspection, should be interpreted
with caution.Moreover, the followup inspection
was conducted before FPL had completed and submitted its response to the SSFI.Many of the responses to the SSFI were still in progress and there had been insufficient
time for many of the corrective
actions taken or planned in response to the SSFI to become effective and be reflected in the followup inspection
results.We request in particular
that our December 6, 1985 response to the SSFI and the description
of corrective
actions planned or taken therein, be given consideration
in your deliberations.
In addition, we ask that you give consideration
to the fact that our ongoing corrective
action activities
have been adversely impacted by the demands imposed by the 2-week regional followup inspection.
8602i90iiO
860iSi PDR ADOCK 05000250 8 PDR uE I PEOPLE...SERVING PEOPLE
J!j~
We believe that the corrective
actions described in our response to the SSFI and in the attached response have been timely and will be effective.
We have implemented
appropriate
procedure revisions and demonstrated
the adequacy of systems and components
by analysis and/or test.Moreover, a significant
number of unresolved
items were already the subject of planned updating actions and at the time of the inspection
were still in progress toward satisfactory
resolution.
As we indicated in our December 6, 1985 response to the SSFI, the results of SSFI indicated a need to augment our ongoing, long-term Performance
Enhancement
Program (PEP)in regard to iVlaintenance
and Conf iguration Control.Additionally, the AFW System Availability/Reliability
Study and Safety System Reviews were viewed to be appropriate
and have been commenced.
These actions have been integrated
into our overall management
plan.We acknowledge
that the burden is on FPL to assure effective implementation
of that plan and we fully recognize and accept that responsibility.
At the same time, we ask that you give consideration
to the positive results of our longer term programs that are now beginning to emerge.As stated in our response to the SSFI, FPRL committed (letter L-85-372, dated September 30, 1985)to apply and implement appropriate
technical specification
requirements
to address the availability
and surveillance
testing of the two non-safety grade motor-drive
standby feedwater'pumps at Turkey Point.The Technical Specification
amendment request has been submitted to the NRC (L-86-03 dated 3anuary 30, 1986).Since the inspection
team concluded that there were no administrative
controls or Technical Specification
requirements
in place to assure the availability
of this system on demand, the team found it inappropriate
to give credit for this system during analysis of the inspection
findings.As stated in FPRL's letter L-85-372, however, these standby"pumps have been routinely run in accordance
with plant procedures".
In addition, backup power supply is obtainable
from five non-safety
grade diesel generators
rated at nominal 2500 kw each.These diesels can supply power directly to nuclear side loads via internal (site)cable runs independent
of the switchyard.
Based upon the foregoing factors, FPRL submits that it would be appropriate
for the capabilities
of the standby feedwater system to be taken into consideration
in the NRC's final analysis of its inspection
findings.A formal description
of the detailed scope of work to be completed by the AFW Availability/Reliability
study, discussed in our response to the SSFI, has been issued.A contractor
has been selected;the AFW Availability-Reliability
study is estimated to be completed within twelve (12)weeks.This study which includes reliability
modeling and an evaluation
of component failure contribution
to reliability, will provide a real time assessment
capability
for AFW system readiness.
A summary of this program is being presented to the Region II staff on 3anuary 31, 1986.FPL has undertaken
a two-phase Safety System review to assure that the concerns expressed in the inspection
report do not apply to the operations
and functions of other important safety systems, or that any appropriate
corrective
actions are promptly taken.The Phase I (Iriitial Assessment)
review has been completed by the Safety Engineering
Group.No system problems that might impede the functional
performance
of the systems selected for review were identified.
With the Phase I results as input for prioritization, FPL will now undertake the more formal and detailed, in-depth Phase II (Comprehensive
Assessment)
review of the selected systems.This review will encompass and
r~V
ensure pertinent design bases are clearly specified, providing additional
assurance that the systems will function as designed.Any necessary corrective
actions will be tracked to implementation.
Phase II has been scoped and scheduled and is estimated for completion
within two years of the commencement
af work activities.
Additionally, in order to more efficiently
control and implement design related issues at Turkey Point, a Site Engineering
Manager has been recently appointed to be responsible
for all site design activities.
As an indication
of the level of importance
assigned to this position, he will report directly to the Site Vice President.
In closing, we reemphasize
our commitment
to improving performance
and assuring that the corrective
actions described in the attached response are effective.
We are confident that our.corrective
actions, when viewed in the context of our overall PEP, will continue to achieve improved performance.
Very truly yours, C.O.Wo Group c resident Nuclear nergy COW/dh Attachment
cc: V.Stello, NRC Executive Director for Operations (Acting)H.R.Denton, Director NRR 3.M.Taylor, Director, NRC Office of Inspection
and Enforcement
J S.E.Elrod, Section Chief, Region II H.L.Thompson, 3r., Division Director, PWR Licensing Division A, NRR S.A.Varga, Director, Project Directorate
No.3, PWR Licensing Division A, NRR L.S.Rubenstein, Director, Project Directorate
No.2, PWR Licensing Division A, NRR D.G.McDonald, Senior Project.Manager, NRR H.F.Reis, Esquire L QD 20 No.of 4oP I~4~I 5 02/04/86 ACOSTA DOIRSY CRISLER PLUGGER SRELH HUTCHIHSON
HARSH NEEDH4H PEEBLES SHOPPH4N NILK YOUNG 4NDERSOH DRAIN DANEK CRANCIS HARPER K4RCH HILLER NUTNELL POTERALSKI
SPOONER NILLI4NS JP ARIAS CH4HLY EHSLHEIKR OOTCH HORRKLL KENT NOAD4 P4HZAHI RE IS VAULT CUSTODI4H NOOD4RD DARRON CONDERY PINCHER GOULDY HUENNISER KERN HcDONALD P4RKER RICHARDS VERDUCI YORK
IS
ATTACHMENT
Re: Turkey Point Units 3 R 0 Docket Nos.50-250, 50-251 Ins ection Re ort 85-00 FPL Res onse to Unresolved
Items and Ins ection Followu Items URI 85-00-01: Licensee Administrative
Procedure ADM 701, Section 5.8.1.8 requires that the root cause of equipment failure be identified
by the journeyman
on the completed P WO.The licensee's
failure to implement this procedural
requirement
is considered
an unresolved
item.~Res onse: 3ourneymen, Supervisors
and GEMS personnel have been directed to ensure that the"Analysis of the Cause or Reason" section of P WO's is completed.
As stated in Inspection
Report 85-00 (Page 0)"the inspector reviewed 15 safety related PWO's completed since the SSF inspection
and noted that all 15 had the root cause section completed as required".
Additionally, the Nuclear 3ob Planning System (N3PS), the development
of which has been underway since the inception of PEP, requires an identical section to be completed on the CRT screen.These actions will enhance root cause identification
and appropriate
corrective
action implementation.
N3PS, when fully automated, will automatically
datalog system equipment history.Field engineers are being added to all three maintenance
disciplines
to enhance corrective
actions after root cause identifications.
The inspection
report also credited the Turkey Point Emergency Response Team (ERT)for its capability
to identify root cause of failures which should"subsequently
reduce the repetitive
failures that have occurred at the Turkey Point Plant".Finally, the inspection
report recognizes
that"the automated PWO program which will provide trending information, the automated PM program, and the performance
based maintenance
training program, should also contribute
to a reduction in repetitive
equipment failures on a long-term basis".(IR85-00, Page q)
URI 85-00-02: The post maintenance
testing requirements
of Administrative
Procedure 0190.19 for instrument
and controls and electrical
areas were informally
completed without specific direction or documentation
on the associated
PWO's.This situation appears to be another example of failure to implement or provide adequate procedures
to control safety related activities.
~Ree onse: The Procedure Update Program (PUP)has been writing post-maintenance
testing requirements
into all PEP maintenance
procedures.
AP 0190.28,"Post Maintenance
Test Control" guidance, has been revised to cover IRC and electrical
maintenance
activities, as well as mechanical
maintenance.
PEP is being enhanced to incorporate
formal post-maintenance
testing criteria into the PWO and Maintenance
Procedures.
The SORP Post Maintenance
Guidance Document (Rev.A), issued in September 1985, will be evaluated for incorporation
and consolidation
of formal post-maintenance
testing criteria.This task is identified
as Project 9, Task 6 and is scheduled to be completed by March 31, 1986.
URI 85-00-03: The licensee failed to provide and implement adequate procedures
to ensure that independent
verification
was performed and documented
on the return to service of instrumentation
vital to the operation of two safety related systems, namely, auxiliary feedwater and backup nitrogen as required by NUREG-0737
Item I.C.6, confirmed by a NRC order dated July 10, 1981.This appears to represent another example of a failure to provide and/or implement adequate procedures
to control safety related activities.
~Res'nse: Procedure 0-ADM-031 (Independent
Verification)
dated July 12, 1985, Step 5.3.1 requires independent
verification
of the removal and return to service of components
controlled
by equipment clearance orders.Procedure 0-ADM-107 dated October 25, 1985 (Writer's Guide for Maintenance
Procedures), Step 5.8A.C gives directions
for independent
verification
for preparing maintenance
procedure.
As part of the PEP maintenance
activities, maintenance
and surveillance
procedures
are being revised to incorporate
instrument
alignment independent
verification.
URI 85-00-00: Licensee Procedure 0208.11, Of f Normal Operating Procedure (ONOP)Annuciator
Panel List-Panel I Station Service, contained erroneous operator action in the event of a low pressure alarm on the nitrogen backup system.The errors existed due to a failure to revise the procedure.
following modification
to the system per plant change/modification (PC/M)80-117.~Ree onse: Procedure (ONOP)0208.11 was changed to clarify immediate operator actions in the event of an alarm.It was approved by the Plant Nuclear Safety Committee (PNSC)on September 25, 1985.Further, EOP's have been revised to require operators to shift FCV's to manual from automatic control within 3 minutes of AFW actuation.
Power Plant Engineering
and Nuclear Energy Departments
met in December 1985 to discuss implementation
of the Standard Engineering
Package.As part of these discussions, an agreement was reached with respect to inter-departmental
coordination
of PC/Ms.Prior to initiation
of the design activity, Power Plant Engineering
and Nuclear Energy will schedule an operability
review meeting.This review will ensure that Engineering
is provided with the necessary system operating information.
This will facilitate
Engineering
providing more detailed guidance in the PC/M package concerning
operating and maintenance
procedures.
As part of the total design effort, Engineering
will review the plant procedures
revised by Nuclear Energy with respect to integration
with the design.This inter-department
coordination
will ensure that the pertinent plant procedures
are identified, reviewed and modified to reflect the new system configuration.
Guidance in this area is in the process of being formalized
in Engineering
and Nuclear Energy procedures.
I
URI 85-00-05: The FPRL Q-List (Quality Instruction
3PE-QI-2.3A)
did not designate the nitrogen backup system electrical
and instrumentation
components
as safety-related.Consequently, the requisite controls over maintenance
activities
were not applied to the component.
PC/M 80-117 correctly designated
the activities
performed under the modification
as safety-related;
however, the document verification
checklist contained in the PC/M failed to list the required changes to the Q-List.Although the checklist requires that changes to the Q-List be indicated, the entry under this item on the checklist was"Later." This apparent failure to adequately
revise the Q-List may represent another example of an inadequate
design control.This item is considered
unresolved
pending further NRC evaluation.
~Res ense: In regard to the design verification
process for the Q-List, FPL has recognized
that the current Q-List is a basic systems level document and is not intended to address individual
system components.
FPL previously
discussed this issue with the NRC (refer to Inspection
Report Nos.50-250/80-33
and 50-250/80-30), and has committed to the development
of an updated and more component specific Q-List.This new Q-List is data base effective as of November 15, 1985, and has been identified
for NRC review as Inspection
Followup Items 250/80-33-03
and 251/80-30-03.
The classification
of the nitrogen system components
has been evaluated by FPL and is reflected in the updated Q-List.Power Plant Engineering
has developed draft Quality Instructions
which define the requirements
for modifications
to and updating of the Turkey Point Q-List.These instructions
require that a Q-List impact review be performed.for all Turkey Point Plant Changes/Modifications (PC/Ms).They further provide the engineer with specific guidance on the mechanics of preparing changes to the computerized
Q-List Data Base.These procedures
will be in place by the end of February, 1986.The current Q-List represents
the plant as depicted on design documentation
current as of May, 1985.FPL's contractor
for the Q-List is being retained to update the Q-List to the now-current
documentation.
This effort is currently scheduled for completion
by August 1986.Power Plant Engineering
will then proceed to maintain the Q-List as a"living document" for future PC/Ms generated for Turkey Point.
URI 85-00-06: The nitrogen backup system PAID (Drawing 5610-M-339)
incorrectly
indicated that the system pressure regulators
were set at 55 psig.Although this drawing was listed in PC/M 80-117 as a drawing requiring update, a change at some point in the implementation
of the PC/M which modified the setpoint to 80 psig failed to ensure that the PdclD was again updated.This appears to be another example of inadequate
design control.This item is considered
unresolved
pending further NRC evaluation.
~Res onse: The pressure control valves were originally
set at 55 psig based on the original design of the plant which was not changed by the modifications
made under PCM-80-117.Drawing 5610-M-339, Sheet 1 of 1 reflected this setpoint.Although this setpoint was acceptable
based on vendor confirmation, the setpoint was adjusted to 80 psig after PCM 80-117 was implemented, to coincide with the normal air pressure operating range specified on the flow control valve data sheet.Due to an administrative
oversight, this change was not incorporated
on the referenced
drawing.The pressure setpoint shown on Drawing 5610-M-339
Sheet 1 of 1, Revision 15, is not a safety concern since the valve can operate at pressures significantly
less than 55 psig based on previous discussions
with the vendor and actual tests in the field.Therefore, the oversight did not affect the operability
of the Auxiliary Feedwater System.Changes are currently being proposed by the Drawing Update Group to improve drawing accuracy.These changes are scheduled to be implemented
by the end of 1986.Revision 17 of Drawing 5610-M-339
and Revision 8 of the associated
instrument
index sheet 5610-M-311
Sht 155 have been issued to reflect the current setpoint of 80 psig.
URI 85-00-07: The NRC Region II inspectors
walked down additional
portions of the licensee's
AFW and related systems for Unit 0.The inspectors
observed penetrations
into the AF W headers which were identified
as an abandoned in-place nitrogen blanket system.This system, at one time, provided a means of introducing
nitrogen into the AFW system and subsequently
into the Unit 0 steam generators.
The inspectors
reviewed Turkey Point Procedure 0-OP-075 to determine if the nitrogen system isolation valves (00-0-1610C, 00-0-16108, 00-0-1283, and 00-0-1280)were identified
in the valve alignment attachment.
The aforementioned
valves did not appear in this attachment.
Turkey Point Procedure 0-ADM-031 dated December 10, 1980, Independent
Verification, states that independent
verification
shall be applied to auxiliary feedwater system applicable
procedures.
This condition appears to be another example of an inadequate
procedure.
This item is, considered
unresolved
pending further NRC evaluation.
~Res onse: Independent
verification
of the nitrogen blanket system valve alignment is not considered
to be required by procedure 0-ADM-031.
Valve alignment of the nitrogen blanket system connected to the Train 1 AFW feedwater lines on Unit 0 is addressed in Operating Procedure O-OP-065.3"Nitrogen Gas Supply System".This procedure identifies
these valves to be in the closed position for normal operation.
These valves are tagged and the tags are checked once per month.This system has been reviewed for its necessity for tie-in to the auxiliary feedwater system.It was determined
that this system is no longer required and is scheduled for removal by PC/M 85-181 during the curr'ent Unit 0 refueling outage.
URI 85-00-08: Apparently
no licensee evaluation
had been or was intended to be performed on the scaffolding
around the AFW FCV's.Housekeeping
procedures
AP 0103.11 and ASP-13 appeared to be inadequate.
~Res onse: A system is being established
to provide a means to better control scaffolding.
permit will be required (except in containment
where other close out controls exist and on secondary system areas which do not directly impact safety related systems)prior to erecting a scaffold which will involve a review by operations'personnel.
This system will address the NRC concerns from page 18 and 19 of the Inspection
Report.This scaffolding
Control System will be proceduralized
in Backfit Procedure ASP-26 by February 28, 1986 and in AP 0103.11 by March 31, 1986.
URI 85-00-09: Emergency Operating Procedures
20000, revision dated August 23, 1985, and 20007, revision dated August 26, 1985, did not provide adequate guidance for the control room operators to assure the required 286 gpm of auxiliary feedwater is delivered to each unit within three minutes in the event of a two-unit trip with only one AFW pump available as specified by the Shared Auxiliary Feedwater System, System Description
and Design Basis, Revision I, dated 3anuary 31, 1985.This appears to be another example of failure to provide adequate procedures.
This item is considered
unresolved
pending further NRC evaluation.
'~Res onse: Emergency Operating Procedure 20000 (Loss Of Offsite Power)revision dated December 26, 1985, provides a note following step 5.3.1.This note provides guidance to assure the required AFW flow is provided in the event of a two unit trip.Emergency Operating Procedure 20007 (Loss Of All A.C.Power)revision dated October 30, 1985 provides a note following step 5.2A.This note provides guidance to assure the required AFW flow is provided in the event of a two unit trip>>
i
URI 85-00-10: The licensee apparently
failed to perform an adequate safety evaluation
with regards to PC/M 80-117, in that at the time of PC/M implementation, the auxiliary feedwater system steam vent valves analysis was not performed for the condition of steam vent valve failure at low pressure conditions.
This item is considered
unresolved
pending further NRC evaluation.
~Res onse: The classification
of steam vent valves as non-safety
related components
is consistent
with the original design basis for the plant and was not changed by PCM 80-117.Therefore, ANSI N05.2.11 was not applicable
to either the original design of the vent valve or the subsequent
modifications.
A design analysis was performed prior to the modification
as documented
in Calculation
MO8-162-02, dated November 20, 1981 to justify operation of the AFW system assuming failure of these non-safety
related valves.The calculation
was based on the scenario during the initial operating conditions
of the Auxiliary Feedwater System with maximum steam pressure in the system.This case was considered
bounding for the range of system operation during a transient.
In response to NRC concerns, a confirmatory
analysis has been performed at the lowest steam operating conditions (at the time the RHR System is put into operation)
which has confirmed previous engineering
judgement.
This analysis is documented
in Calculation
MOS-062-02, dated October 11, 1985.The analysis demonstrated
and confirmed adequate steam supply to the Auxiliary Feedwater pump turbines is the event of a complete failure of the steam vent line.The selection of the setpoint for the steam vent valve on the new steam supply header was based on the setpoint established
under the original plant design for the vent valve in the existing header.At the time the new header was added, as an exact duplicate of the existing header, the setpoint of the original steam valve was specified for the new valve.There was no reason to question the validity of, the original valve setpoint since the new valve was functionally
identical.
As stated previously, a design analysis has been performed which confirms the system's operability
with the vent open at low steam pressure.In any event the subject valves will be removed as discussed in the response to URI-85-00-11.
A 0
FP@L is conducting
a design review of the steam vent valve function.This review is expected to be complete by December 30, 1985.The review will address the following questions:
(1)are the vent valves needed for the present auxiliary feedwater system design?(2)What was'the reason for the 150 psi setpoint?(3)If the vent valves are required for system operability, can the 150 psi setpoint be reduced?The result of the design review and the licensee's
actions will be inspected at a future date as an inspector followup item.~Res onse: As part of our AFW system enhancement
task team project (Item 8), we have reviewed the design basis for the original,AFW
steam header leakoff valves to determine their applicability
to our present system.The results of this evaluation
are documented
in Bechtel letter SFB-2107 dated December 19, 1985.The original Turkey Point Auxiliary Feedwater System contained pumps driven by low pressure steam turbines.In order to avoid over speeding and tripping of the turbines upon initial startup, the turbine pressure control valves were maintained
closed.To start the pumps, the steam isolation valves (MOV->>-1003, MOV-~-1000, MOV-+-1005)
were opened and the common steam supply line was pressurized.
A pressure switch, located in the steam supply line then furnished a signal to open the pressure control valves at the turbine inlet to start the pumps.Normally open auxiliary feedwater steam vent valves (CV-"-6008, CV-+-2910)
were provided downstream
of the steam supply isolation valves to prevent pressurization
of the steam supply line, due to isolation valve leakage.Pressurization
of the steam line, as a result of isolation valve leakage, would cause undesirable
cycling of the pressure control valves as well as the pumps.The normally closed pressure control valves at the turbines were removed when the new high pressure turbines were installed.
The new turbines were provided with motorized trip and throttle valves as well as governors with ramp bushings.The motorized trip and throttle valves are normally maintained
open.The governor with ramp bushing will prevent overspeed tripping of the turbine upon initial starting.Therefore, the Auxiliary Steam Vent Valves are no longer required to prevent pressurization
of the steam supply line as a result of isolation valve leakage, and undesirable
cycling of the pressure control valve.As a result of this evaluation, PC/M's85-199 (Unit 3)and 85-200 (Unit 0)are being prepared to remove the vent valves from the system.This modification
is scheduled for implementation
in Unit 0 during the currerit refueling outage.
50
V RI-85-00-12:
Procedure 3-OSP-075.1, dated August 7, 1985, did not adequately
verify that i41OVs 3-1000 and 3-1005 were independently
capable of opening all their associated
AFW flow control valves as designed.The upgrade of Procedures
3-OSP-075.1 and O-OSP-075.1
dated August 1985 required opening both steam supply valves together which fails to ensure that either MOV 3-1000 or MOV 3-1005 could open all associated
flow control valves in trains 1 and 2 thus ensuring a feedwater flowpath.10 CFR 50, Appendix B, Criterion XVI, requires that measures shall be established
to assure that conditions
adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances
are promptly identified
and corrected.
The aforementioned
procedure upgrade performed appears to represent an inadequate
corrective
action to a previously
identified
violation.
This item is considered
unresolved
pending further iXRC evaluation.
~Res onse: Technical Specification 3.8.0 provides LCO's for AFW trains.The procedure cited in the finding provided instructions
which did in fact test the capability
of the trains of AFW.iVo requirement
was known to test the valves independently.
Procedure 3/0-OSP-075.1 now requires that each MOV is independently
verified capable of opening all FCV's.The steam supply MOV's (1003,1000
and 1005),are designed to trigger a switch upon opening to activate solenoids to allow operation of the flow control valves.Opening of each MOV will allow operation of all six flow control valves (three for each train).The violation identified
in Report Nos.50-250/80-35
and 50-251/80-36
addressed the failure to visuaLly verify the flow control valves would operate as designed.This previous violation did not explicitly
require verification
that the MOV's were independently
capable of opening all associated
AFW flow control valves.FPL conducted a procedure upgrade to include visual verification
of valve operation.
Procedures
3-OSP-075.1
and O-OSP-075.1
are followed to insure operability
of each train of auxiliary feedwater.
Operating valves 1000 and 1005 together verified operability
of train 1.However, these procedures
now require that each MOV is independently
verified to be capable of opening all FCV's.This will insure more prompt identification
of individual
component failure.
I
URI 85-00-13: Failure to meet a committment
in a FPL letter dated December 20, 1979 to install adequate communications
and DC lighting to support local AFW operations
during testing and during control room inaccessibility
is unresolved
pending further NRC evaluation.
~Res onse: FPL's letter dated 3uly 22, 1980 indicated that AFW system modificaitons
had negated the need to add DC lighting or a sound powered phone link and thus FPL did not consider the December 20, 1979 committment
to remain in effect.Nevertheless, DC lighting for the AFW pump area, feedwater platforms, and normal steam isolation valves has been completed, and a sound powered phone link between the AFW pump area and the control room will be complete by 3une 1986.
.C I
URI 85-00-10: The control room inaccessibility
procedure 0-ONOP-103
dated August 7, 1985, did not address the local operation of train 2 of the AFW system.This apparent failure to provide an adequate procedure to cover a safety related activity is unresolved
pending further NRC evaluation.
The procedure also did not provide instructions
on how to locally reset and restart a tripped AFW pump.~Res onse: Procedure O-ONOP-103, Control Room Inaccessibility, will be revised as follows: Prior to taking any operator actions at Train 1 of the AFW System, the operator will be directed to check the AFW flow gauges to determine train operability
and report the resulting finding to the Plant Supervisor
-Nuclear.2e A PC/M has been submitted to provide a means (reach rod)to make the instrument
air isolation valves for Train 2 of the AFW System easier to operate.A procedure change will be made as necessary to incorporate
any PC/ill induced information
into the procedure.
3.Instructions
and setpoints will be incorporated
to provide guidance to the operator of how to obtain proper flow to the steam generator utilizing a single AFW pump.Instructions
will be incorporated
to address balancing of the AFW System flow to provide a flow rate of 286 gpm to each unit in the.'vent only one pump is operable concurrent
with a dual unit trip.This will be the same as presently delineated
in EOP 20000 and 20007.5.Instructions
will be incorporated
to provide instructions
to the operator for locally resetting and restarting
a tripped AFW System pump One item of disagreement
with the report should be noted though.Isolation of the nitrogen valves at Train 1 of the AFW System does not isolate nitrogen to the Train 2 valves, therefore, the valve operation isolating nitrogen to Train 2 should not be deleted as stated in the report.No procedure change will be made.
IFI 85-00-15: A review of the training and procedural
improvements
for assurance that the plant can be safely controlled
from outside the control ro'om will be made at a later date by an NRC evaluated walkthrough
of the entire control room inaccessibility
procedure.
~Res onse: Extensive training for'perators and operator trainees on control room inaccessibility
has been conducted.
Each shift crew has conducted a walkthrough
of the procedure.
In addition this subject will be part of the next requalification
cycle training scheduled to start in March-1986.As indicated in the inspection
report"The upgrade in lighting communications, procedural
improvements;
and additional
training and walkthroughs
to be conducted by the license should provide reasonable
assurance that the plant can be safely controlled
from outside the control room."
IFI 85-00-16: Post-accident
radiation zone map 5177-119-SK41-I
dated April 27, 1981 indicates two of the AFW pumps would be in a very high radiation environment
for a Unit 3 LOCA.In addition, although dose rate instruments
are being kept in the control room, it appears that there is no guidance as to their use.~Res onse: The AFW pumps are not required for mitigation
of a large break LOCA on the affected unit.Should the pumps be required for the opposite unit, the personnel that would be required to take action on the two pumps would be staged in the onsite support center, where health physics review of tasks in the high radiation zone would be evaluated prior to dispatching
individuals
into the zones.All personnel required to work in radiation areas are trained to operate radiation detection instruments
and are aware of actions to be taken for high readings on the instruments.
In addition, FPL will conduct a review of the basis for the radiation map in question.
URI 85-00-17: Administrative
Procedure 0103.3, Control and Use of Temporary Systems Alteration, dated 3anuary 31'98k'ection
5.8 states, the Plant Nuclear Safety Committee (PNSC)is responsible
for reviewing applicable
nuclear safety-related
temporary system alterations
within 10 days of the Plant Supervisor-Nuclear
approval date.Apparently, the licensee's'PNSC
failed to review and document TSA 3-80-11-75, TSA 3/0-85-8-75, TSA3/0-80-99-75
and TSA 3/0-80-100-75
within the prescribed
time period.This appears to be another example of failure to implement approved procedures, and it is considered
unresolved
pending further NRC evaluation.
~Res onse: The TSA procedure (0-ADM-503)
has been revised 3anuary 10, 1986, to require prior PNSC review of TSA's for equipment in service or component substitution.
The 10 day review now applies only to TSA's implemented
on equipment or systems out of service or covered by a clearance.
This enhancement
of management
controls should reduce the likelihood
of recurrence.
~%
U RI-85-00-18:
Apparently, the licensee failed to perform an adequate safety evaluation
on TSAs 3-80-11-75, 3/0-80-99-75, 3/0-85-8-75
and 3/0-80-100-75, Removal of AF W Covernor Speed Control Systems, in that the safety evaluations
did not evaluate the mechanical
reliability
of the AFW pumps being operated under constant speed conditions.
This item is considered
unresolved
pending further NRC evaluation.
~Res onse: A safety evaluation
was conducted for these TSAs in accordance
with AP 0103.3.However, the mechanical
reliability
of the pumps operating at a constant speed within the previously
analyzed operating range was not addressed.
To enhance management
control of TSA evaluations, the Procedure (0-ADM-503)
has now been changed so that TSAs for equipment in service will be reviewed by the PNSC prior to installation.
The AFW pumps and turbines are designed to operate up to a normal operating speed of 5900 RPM.The governor on the turbine driver is installed and set to maintain speed at a high speed setpoint of 5900 RPM.It, is capable of receiving an external signal to vary turbine speed between 3200-5900 RPM...Until recently, an external air signal was provided by a differential
pressure controller
which maintained
a constant pressure drop across the flow control valves.Failure of this controller
resulted in the turbine being required to operate at constant speed.Removal of the differential
pressure controller
would also have resulted in the turbine operating at constant speed..This operating mode is within the previously
analyzed operating range for both the turbine and pump.The designed mechanical
reliability
of the equipment is not considered
to be reduced for that reason.
I
URI 85-00-19: It appears that the licensee failed to take prompt and adequate corrective
action to ensure that manual isolation valves 3-20-028 and 0-20-028, the common isolation valves for redundant safety-related
condensate
storage tank level instruments
were properly administratively
controlled.
This item is considered
unresolved
pending further NRC evaluation.
~Res onse: s The design bases for the addition of the redundant condensate
storage tank level indication
system, installed under PCM 80-77, was based on the connection
to the tank being a passive portion of the system which allowed the redundant monitors to be on a common tap.This approach is considered
acceptable
since a single passive failure of this line or the isolation valve is not a design basis for Turkey Point.Prior to the implementation
of PCM 80-77, the condensate
storage tank was provided with a single level transmitter
downstream
of Isolation Valve 028.When the redundant level indication
system was added downstream
of Valve 028, it was assumed that the operation of the valve was adequately
controlled
by administrative
procedure since PCM 80-77 did not modify either the tap off the tank or the valve.The need to control the isolation valve was not addressed in PCM 80-77 since the valve was existing and performing
the same function, and Operating Procedure 7001.1 administratively
controls Valve 028 in the open posi tion.In advertent closure of Valve 028 would create the potential for the operator to have misleading
information
concerning
the condensate
storage tank level.However, this is not considered
to be a significant
safety concern since there are other independent
methods of determining
tank level which would have alerted the operator to recognize that the level indication
system was not functioning
properly.Level Switch LS-3/0-1503, which is safety related, alerts the operator that the minimum Technical Specification
volume of 185,000 gallons is remaining in the tank, would be available since it is not associated
with this level tap.Should the condensate
storage tank reach this level, the operator would have noted a discrepancy
in the level readings and taken corrective
action.Also, control room alarms are available to alert operators with respect to tank level.The condensate
storage tank Technical Specification
also requires, by definition, a minimum volume of water for nineteen hours of Auxiliary Feedwater System'operation.
With the Auxiliary Feedwater System in operation and drawing on the inventory of the condensate
storage tank, the operator would have noticed that the condensate
storage tank level (as indicated by the redundant transmitters
and" checked by log readings)was not decreasing
during this time period and questioned
the validity of the level indication;
appropriate
corrective
action could have been taken.In addition, both condensate
storage tanks are normally aligned to the Auxiliary Feedwater pump suction.Assuming Valve 028 was inadvertently
closed on one tank, the level on the opposite tank would be operable.Since the levels in the
URI 85-00-19 (continued):
two tanks will decrease at approximately
the same rate, a disparity between the levels in the two tanks would have been recognized
by the operators and appropriate
corrective
action could have been taken.It should also be noted that the design modification
process has been substantially
improved since the time this modification
was implemented
in~recognition
of the need to coordinate
changes in the plant with operations
and maintenance
personnel.
A program for the review of proposed plant modifications
has recently been established
to ensure that the effects on operating documents, procedures
and administrative
controls are accommodated
in the design prior to approval of the PCM by the Plant Nuclear Safety Committee (PNSC).Engineering
personnel are also currently on controlled
distribution
for the plant operating procedures, which provides the design.engineer with a better insight into the actual operation of the system and the potential impact of modifications
of the system.Also, utilization
of Standard Engineering
Packages should greatly aid this area.As noted in the NRC report, Valve 028 has been locked open.In addition, the associated
drawings have been revised to show this valve locked open by administrative
control and the valve has been added to the locked valve list.Administrative
Procedure AP0103.5 (Administrative
Control of Valves, Locks and Switches)provides instruction
for placing valves, locks and switches under administrative
control when it is necessary to lock the valve or switch to prevent inadvertent
misoperation
of the valve or switch.These valves were added to the procedure revision approved November 13, 1985.
I
URI-85-00-20:
The apparent failure to ensure that procedure 3/O-OP-018.1, Condensate
Storage Tank, a safety-related
procedure was approved for release by authorized
personnel and appropriately
distributed
prior to the cancellation
of OP-7001.1, is considered
unresolved
pending further NRC evaluation.
~Res onse: This is considered
an isolated case which occurred due to the complexity
of the specific change.In this case several procedures
were being issued to replace one old procedure.
Because one of the replacement
procedures
was undergoing
review in a different section of the plant staff, the old procedure was inadvertently
cancelled when the remaining replacement
procedures
were issued.Administrative
Procedure 0109.7 provides guidance for the PUP group to cancel procedures
as new replacement
procedures
are generated.
IFI 85-00-21: The inspector noted several examples where cancelled procedures
were referenced
in other procedures
still in use.The inspector was informed by the licensee that at present there is no method for cross-referencing
procedures
to ensure that a cancelled or changed procedure does not affect another procedure.
The inspector informed the licensee that the development
of a method to ensure all procedures
are properly updated could be of benefit.This is an inspector followup item.~Res onse: It is currently the procedure writer's responsibility
to ensure that cancelled or changed procedures
do not affect another procedure.
FPL is considering
a proposal for a computer cross referencing
system.It is expected that a decision will be made on this potential enhancement
by March 15, 1986.
URI 85-00-22: The apparent failure to evaluate the impact of design changes on the AFW control system on the nitrogen consumption
rate is considered
an unresolved
item pending further NRC evaluation.
~Res onse:.The modification
to split the nitrogen backup system into two headers was issued for implementation
under the original scope of PCM 80-117 as shown on Drawing 5610-M-339/80-55, Revision 2, dated 3anuary 15, 1982.This design was established
by engineering
judgement (although not fully documented)
based on a technical evaluation
of the original design basis for the nitrogen backup system in consideration
of the following factors: o The total number of flow control valves supplied by the on-line bottles in the split system was half of that in the original design.The original design had one bottle on line serving six flow control valves.The modified system resulted in one bottle on line serving three flow control valves.o The total nitrogen consumption
for the modified system was significantly
less than the original system.The air consumption
rate of original flow control valves was 1.0 scfm per valve, as compared to the air consumption
rate for the replacement
valves of 0.26 scfm, based on the original regulator setpoint of 55 psig.o The pump differential
pressure controllers
were installed and operable, and maintaining
the design pressure drop across the flow control valves.On this basis, valve oscillations
were not an operational
problem.o The design flow rate through each flow control valve was 200 gpm.o The low pressure alarm setpoint for the nitrogen bottle system at the time the PCM was issued was 1005 psig, which allowed 15 minutes for the operator to take the necessary action to valve-in a new bottle.o Air operated hand controll'ers
for the flow control valves supplied by the backup nitrogen were removed from the system as defined in the scope of work under PCM 80-55.The split of the nitrogen system into separate trains under the original scope of PCM 80-117 was considered
acceptable
within the parameters
of the original design basis for this system.However, in response to NRC concerns with the engineering
judgement used as a basis for this modification, a detailed analysis of the nitrogen backup system was performed, based on the original design parameters
described above and worse case results of previous tests performed on the system.The results of this analysis are documented
in Calculation
MO8-062-05, Revision 0, dated November 1, 1985.This analysis demonstrates
that sufficient
nitrogen should have been available from the valved-in bottles in the
I
split system to permit the system to operate for more than 20 minutes without operator action following receipt of a low level alarm at 1005 psi.Based on these results, the split of the nitrogen system was confirmed to be consistent
with the original design basis and operating procedures
for the system.Responses to the specific NRC review team concerns are discussed below: o The NRC review team indicated that a steady-state
consumption
rate of 0.26 scfm was utilized in evaluating
the nitrogen consumption
rate for the new flow control valves, instead of a rate of 0.36 scfm.The 0.26 scfm-consumption
rate was based on the original regulator setpoint of 55 psig.As discussed above, Calculation
MO8-062-05
was prepared to analyze the available nitrogen from the split system.This analysis included a steady-state consumption
rate greater than 0.36 scfm to account for system leakage and the change in the regulator setpoint to 80 psig, and confirmed the acceptability
of the split system.o The NRC review team concluded that the assumption
of instantaneous
steady-state
operations
was not consistent
with the as-designed
valve response.As discussed above, the analysis supporting
the split nitrogen system was based on the original plant design features which limited flow control valve oscillations.
This is consistent
with the vendor's technical literature
which indicates that the valves quickly reach their setpoint with virtually no overshoot.
However, several changes were made to this system after PCM 80-117 was released for implementation, which induced oscillations
in the flow control valves and resulted in a subsequent
increase in the nitrogen demand for the system.These changes were unrelated to the original scope of PCM 80-117 and included the following:
1.The dif ferential pressure controllers
on the Auxiliary Feedwater System were disconnected
due to maintenance
problems with these components
and difficulties
in obtaining spare parts, which resulted in an increased pressure differential
across the flow control valves.This increased pressure differential
resulted in oscillation
in the flow control valves, and in increase in the nitrogen consumption
rate under test conditions
witnessed by the NRC review team.Removal of the pressure controllers
was justified by analysis on the basis that the excessive differential
pressure across since the steam generator pressure would rapidly rise to the safety valve setpoint.Under this condition, the pressure drop across the flow control valve would be essentially
the same regardless
of whether the differential
pressure controller
was installed.
2.Another change which induced oscillations
in the flow control valves was the reduction in the auxiliary feedwater flow rate from 600 gpm to 373 gpm.This reduced flow rate resulted from a Westinghouse
reanalysils
of the feedwater flow requirements
as documented
in
Westinghouse
letter W-PTP-62, dated 3une 3, 1982.As a result of this change, the setpoint for the flow control valves were revised to 125 gpm.Reduction in the setpoint compounded
the oscillation
problems in the flow control valves since the design basis for the system was established
at 200 gpm.Subsequent
to this change, a review of the valve oscillation
problem was made during field testing of this system.These tests confirmed that the control valves performed satisfactorily
under the original design condition of 200 gpm flow at 25 psi differential.
However, oscillating
control valve action was experienced
when the system was tested at the reduced flow rate of 125 gpm and the auxiliary feedwater pumps running at maximum speed.As a result of these tests, modifications
were recommended
to eliminate the valve oscillation
problems.It is anticipated
that new valve trim will be installed by the upcoming refueling outages for each unit.o The NRC review team noted that the reduction in the low pressure alarm setpoint from 1000 psi to 500 psi did not appear to be based on a documented
analysis.This change was made based on a field performance
test conducted on March 1, 1980.The criteria developed for the test was based on steady state operation of the valves on the understanding
that valve oscillations
were not a design basis for the system consistent
with the vendor's technical literature
which indicates that the valves quickly reach their setpoint with virtually no overshoot and the oscillations
identified
during testing would be eliminated
by subsequent
modifications
to the valves.The amount of time required for valving in the next bottle upon actuation of the low level alarm was established
at 10 minutes based on discussions
with plant operations
personnel.
A total of six flow control valves were included in the test to add conservatism
to the setpoint since only three valves are aligned to one open bottle in the new system arrangement.
Also, the valves operated for approximately
15 minutes based on the 500 psig setpoint which was considered
another safety factor margin for the setpoint.On this basis, the test is considered
to be a satisfactory
method of establishing
the low pressure alarm setpoint, in lieu of a documented
analysis.The NRC's concerns with the volume of available nitrogen to the flow control valves are directly related to the valve oscillation
problems witnessed during the inspection.
However, as discussed previously, valve oscillations
are not considered
a safety concern since the high differential
pressure across the flow control valves not expected to exist when the system responds to a design basis accident.FPRL has pursued resolution
of this problem in a systematic
manner through coordination
with its architect-engineer.
NRC's final evaluation
of its inspection
findings should give due recognition
to the fact that this problem was identified
by FPL prior to the inspection
and that design modifications
were in progress.
When the NRC audit identified
increased N2 consumption
due to the valve oscillations
noted during testing, FPL reanalyzed
the nitrogen system and revised the low pressure setpoint to 1350 psig, to allow a minimum of ten minutes for the operator to valve-in bottles based on a conservative
assumption
of valve oscillation
and with no credit taken for placing the valves in the manual mode.The low pressure alarm has been temporarily
reset in the field to this revised setpoint, and the appropriate.design documents have been revised.A surveillance
procedure to dynamically
test the nitrogen back-up system is currently being prepared and is scheduled for issuance by April 30, 1986.When implemented, this test should identify any detrimental
effects on nitrogen consumption
created by modifications
to the AFW system.
l
IF I 85-00-23: The AP 190.15, Document Verification
Checklist, is still not being completed under Step 5 which requires that changes to the"Q" List be listed.The entry on PC/M 83-117 contains"to be determined
by FPRL." Although the PC/M has been turned over and completely
closed out, necessary changes to the"Q" List have not been evaluated.
Plant personnel stated that this was due to the development
of the new, component level"Q"-List which will soon be issued.This check has been neglected on PC/Ms which have been processed during the"Q"-List development.
Once the"Q"-List has been promulgated, engineering
procedures
will be developed for maintenance
of the"Q"-List through evaluations
of system modifications.
This item will be left as an inspector followup item to verify appropriate
procedures
and responsibilities
are developed.
~Res onse: Turkey Point Plant and Power Plant Engineering
procedures
are currently in the process of review and modification
to incorporate
the new component level Q-List.At Turkey Point Plant, the Q-List Task Team has identified
the Administrative
Procedures
which will require modification.
The Procedures
have been marked up with the team's recommended
changes.A meeting will be scheduled for the end of January to discuss these proposed changes with Maintenance, Operations, QC, Procurement, and Procedure Update.At that time, an incorporation
schedule will be developed.
Power Plant Engineering
has developed draft Quality Instructions
for Q-List use and maintenance.
A meeting will be scheduled for the beginning of February to discuss comments on these procedures.
These procedures
are expected to be finalized and implemented
the end of February.Since the new Q-List reflects plant documentation
current as of May, 1985, there exists a"del'ta" between the Q-List and up to date documentation.
The Q-List contractor
is being retained to update the Q-List for FPL and this effort is expected to be complete by August 1986.Power Plant Engineering
will then proceed with future updates of the Q-List.Engineering
and plant personnel are now using the computerized
Q list.Training is underway and for the interim period while the new Q list is being fully implemented, the plant is also checking the PNSC approved prior Q list in addition to the new list.
IF I 85-00-20: The licensee is preparing an engineering
evaluation
addressing
how iong the AFW system must operate without operator action in the automatic flow control mode and subsequently
in the remote-manual
mode.This evaluation
will necessarily
impact on the required volume of stored nitrogen.It was also noted that the licensee is planning to add an additional
five-bottle
nitrogen supply so each train will have five bottles available.
The results of the engineering
evaluation
and the additional
nitrogen supplies will be tracked as an inspector followup item.~Res onse: The original nitrogen backup system for the auxiliary feedwater flow control valves consisted of five nitrogen bottles to supply motive power for six flow control valves.Original calculations
concluded that the five bottle station would provide sufficient
capacity for two hours of valve operation, however this was not a design basis requirement
for the system.The nitrogen station is designed to allow sufficient
time for bottle change out while maintaining
system operation.
A recent review of the present system has resulted in the preparation
of PC/M's to install additional
bottle stations for each unit to extend the presently sufficient
time to change out bottles during AFW system operation with the N2 station in use.The additional
bottle station for Unit 0 is scheduled for installation
during the current refueling outage.An evaluation
to more clearly address the operator action requirements
for the system is presently being prepared for incorporation
into the AFW system design basis document.
URI-85-00-25:
The inspector noted that a portion of the nitrogen system that was outside of the scope of NCR 301-85 was supported at intervals greater than the 36 inches specified by the licensee's"Design Guide for Seismic Class I'Instrument
Tubing Installation" for stainless steel tubing.This document is FPRL's specification
No.5177-3711, Revision 2.The requirement
for a maximum unsupported
span of 36 inches is adopted from the ASME Code Section III.Contrary to this design requirement, the inspector measured two adjacent supports that were 01 inches and 03 inches apart.Not only do these exceed the licensee's
seismic Class I requirements;
but the instrument
tubing was not attached to the central support between these two spans which creates a section of unsupported
tubing approximately
85 inches in length.The inspector also found a section of unsupported
stainless steel instrument
tubing in Unit 0 of 01 inches in length.Pending further review of the circumstances
which resulted in the nitrogen system not being seismically
supported, this item is considered
unresolved.
~Res onse: All of the specific support conditions
identified
during the inspection
have been evaluated and the system has been determined
operable based upon functionality
criteria.To further identify potentially
unsupported
connections
to the auxiliary feedwater system and other safety systems, FPL has initiated a program to walkdown and"as built" all 2 inch and under piping associated
with these systems not evaluated under IE Bulletin 79-10.
~1
URI 85-00-26 Consideration
for NPSH was not documented
in the FPRL calculation, Low Level Alarm on Condensate
Storage Tank, dated November 15, 1979, and the calculation
inadequately
identified
the necessary assumptions
and"design inputs.An informal, undated, untitled, annotated sketch was presented by the licensee as evidence of NPSH consideration, which had not been properly referenced
in the November 15, 1979, calculation.
The sketch lacked proper identification
and detail to permit an understandable
review.It appears that the licensee failed to adequately
document all assumptions
and design inputs for FP2L's calculation
of November 15, 1979, on the Low Level Alarm on Condenste Storage Tank.This item is considered
unresolved
pending further NRC evaluation.
~Res onse: The in'spection
report stated that"...the preparer appeared to have assumed that the minimum NPSH would be below the instrument
tap, because the analysis calculated
the height above the instrument
tap which corresponds
to 20 minutes of water at a usage rate of 600 gpm with a 10'actor for conservation.
The team independently
confirmed that the NPSH is well below the instrument
tap and the design is not deficient".
Engineering
has found evidence on the microfilm records which indicates that NPSH was considered
in the original calculation.
However, this consideration
for NPSH was not documented
in the FPL calculation
dated November 15, 1979.Calculation
MO3-062-01, dated October 1, 1985, was recently performed to confirm that the required NPSH water level for flows anticipated
at the end of the cooldown transient is below the instrument
tap level as assumed in the original calculation.
The new calculation
confirmed the results of the original calculation,.and
therefore the equipment and associated
condensate
tank level setpoints are acceptable.
Enhanced documentation
of calculation
assumptions
are being pursued for both internally
and contractor-developed
calculations.
Quality Instruction
revisions have been or are in the process of being issued to provide enhanced controls for documentation
of calculational
assumptions
and inputs for both internal and contractor-developed
calcula'tions.
I
URI 85-00-27: The inspector located two subsequent
examples of OTSCs which were processed as a change of intent.On November 12, 1985, the PNSC reviewed and the plant manger approved OTSC No.3730 to TOP 206, Reactor Protection
System Periodic Test (Unit 0).The change of intent guidelines
checklist indicated that this change did alter the intent of the original procedure.
This procedure change was made in response to IE Bulletin 85-02.On November 12, 1985, the PNSC also reviewed and the plant manager approved OTSC No.3733 to AP 103.12, Notification
of Significant
Events to the NRC.The change of intent guidelines
checklist indicated that this change did alter the intent of the original procedure.
This procedure change was also made in response to IE Bulletin 85-02.The conduct of these two temporary procedure changes is being further reviewed by the NRC and considered
an unresolved
item.~Res onse: Administrative
Procedure 0109.3 (On the Spot Changes to Procedures)
is being revised to differentiate
between temporary changes made in accordance
with Technical Specification 6.8 and on the spot changes made with prior PNSC approval.The temporary change instructions
will strengthen
controls so that no changes to the intent of procedures
will be made under this portion of the procedures.
Because prior PNSC approved OTSC will be given all required safety reviews, change of intent will be allowed for this type procedure change.
I
IF I 85-00-28: A review of the new maintenance
training and the qualification
tracking system will be an inspector followup item.~Res ense: FPL personnel will be available to discuss the new maintenance
training and the qualification
tracking systems during the implementation
and upon completion
of the project.
URI 35-00-29: There appears to be a difference
in the philosophy
between the manufacturer's
recommended
motor overload heater size and the size chosen by the licensee in order to agree with Regulatory
Guide 1.106.This item will require further review by NRC and is identified
as an unresolved
item.~Ree onse: As stated in our previous response to this item contained in our letter L-S5-039, Turkey Point has made no commitment
to utilize the Limitorque
sizing recommendations
for overload heaters.Our current design meets the general intent of Reg.Guide 1.106.The overload heaters are sized by the heater manufacturer
using his standard sizing criteria and the applicable
plant motor data.We have however, determined
that we will re-evaluate
our philosophy
on overload heater sizing in accordance
with the Limitorque
sizing recommendations
and Regulatory
Guide 1.106, taking whatever action is required.
URI 85-00-30: The safety system functional
inspection
team inspector expressed concern that the MOVs would not function under the conditions
as stated in the manufacturer's
letter and therefore requested a review of the manufacturer's
calculations
or that testing be performed at the low voltage conditions
to verify acceptability.
Additionally, the reduced voltage will result in an increased operating travel time.This added time should be reviewed to determine if any ISI requirements
are affected.This item will be unresolved
until a review of the manufacturer's
calculations
or testing of these MOVs at reduced voltage is accomplished.
~Res onse: As stated in our previous response to this item contained in our letter L-85-039, calculation
5177-062-EOI
was prepared utilizing a very conservative
starting current of 53 amps.Subsequently
this calculation
has been reviewed utilizing the Limitorque
data for starting current and the results indicate that the voltage at all of the subject valves will exceed 90V.This reduced voltage at MOV-0-1003 will result in a slightly increased stroke time which based upon a preliminary
investigation
should not adversely affect the design basis for the Auxiliary Feedwater System.FPL is preceeding
with preparation
of a revision to calculation
5177-062-EOI
utilizing the Limitorque
data for starting current and also with a formal review'o determine what effect, if any, increased MOV stroke time may have on the design basis for the Auxiliary Feedwater System.These items are expected to be completed by February 21, 1986.
'I
URI 85-00-31: The safety system functional
inspection
team expressed concern regarding the operation of a second steam vent valve that had been added between the steam admission valves and the AFW pump turbine.Bechtel Power Corporation
calculations
No.MO8-062-02
approved September 30, 1985 (for low steam pressure conditions)
and No.MO8-162-02, approved November 20, 1981 (for high steam pressure conditions)
indicate that a 17'argin of steam is available with the 3/0" vent valve open which would be the condition for loss of AC.This item will remain unresolved
pending review of the calculations
by the NRC.~Res onse: The non-safety
related classification
for the steam vent valve added under PC/M 80-117 is consistent
with the original design basis for the plant.As discussed previously
in response to item 85-00-10, a detailed analysis was performed prior to equipment installation (Bechtel calculation
MO8-162-02
approved November 20, 1981)which confirmed the capability
of the AFW system to operate if the vent valve failed open.Bechtel calculation
M-08-062-02
approved September 30, 1985 was performed to confirm previous engineering
judgement and insure system operability
with the failed vent for the full operating range of the system.On this basis, there has no reason to power the steam vent from a safety related power source and powering the valve from a non safety AC source is considered
acceptable.
To support the NRC review, the referenced
calculations
are available at the Turkey Point Plant site.
1
URI 35-00-32: It appears that the low nitrogen pressure switches were not reviewed during the Seismic Qualification
of Auxiliary Feedwater System evaluation.
The licensee advises that these switches arg part of the original design and therefore, not designed or considered
safety-related.
However, current plant emergency operating procedures
require action by the operator upon receipt of the nitrogen low pressure alarm.Additionally, consideration
must be given to the fact that the nitrogen system is a backup system and the status of backup system should be known at all times.Since there appears to be confusion as to the safety-related
application
of these switches, this item is unresolved
pending further NRC review.~Res onse: As stated in our previous response to this item contained in our letter L-35-039, the design modifications
for the Auxiliary Feedwater System utilized the pressure switches and annunciation
system installed under the original plant construction, which were neither designed nor maintained
as nuclear safety-related.
As a result, the separation
criteria for these components
was not changed from the original design basis of the plant, and ANSI N05.2.11 does not apply.However, we have determined
that this system will be redesigned
as part of the Auxiliary Feedwater System upgrade which involves the addition and relocation
of the Auxiliary Feedwater Nitrogen Stations.This redesign will consist of the installation
of new qualified pressure switches, indicators
and wiring thereto.A pressure switch at each station will be alarmed in the Control Room with a trouble light and will be of a safety grade design.
~s~URI 85-40-33: The inspector reviewed Tech Spec.0.8.2.b and determined
that it requires the licensee to monthly perform an equalizing
charge on each battery and to take specific gravity and the voltage readings of each cell.It is Plant Procedure 9600.1 which implements
this requirement.
However, it appears that this Technical Specif ication requirement
is more applicable
to lead antimony batteries which were the original type of batteries installed at Turkey Point.The Gould-GNB manual recommends
that lead calcium batteries should be given an equalize charge only when needed.The inspector also observed that Plant Operating Procedure 9600.1 identifies
two different float voltages for the Gould-GNB NCX type batteries.
The inspector questioned
the licensee about the acceptability
of Tech Spec 0.8.2.b and about what effects this monthly equalize charge may have on the lead calcium batteries.
The licensee indicated that they would contact the vendor to determine the acceptability
of the current licensee requirements.
This item is identified
as unresolved
item pending licensee and NRC evaluation.
~Res onse: We have contacted the battery manufacturer, Gould, with regards to the acceptability
of Technical Specification 0.8.2.b and the effect<he monthly equalizing
charge may have on the lead calcium batteries.
Gould has stated that the overcharging
of the batteries at IOOVDC once a month does not have any appreciable
effect on the battery qualified life.In order to clarify the battery charging requirements, a Technical Specification
change to provide for an as-needed equalizing
change will be submitted by March 31, 1986.This date revises that provided in our letter, L-85-039, dated December 6, 1985.This time extension allows for a more thorough review of the requirements
and better coordination
with the Standard Technical Specification
Project.
J y~
URI 85-00-30: Technical Specification 0.8.2.b states that monthly, each battery shall be given an equalizing
charge, and afterwards
specific gravity and voltage readings shall be taken and recorded for each cell.It appears that Plant Operating Procedure 9600.I dated August 7, 1985, was inadequate
in that it did not contain vendor recommendations
for compensating
cell specific gravity readings for electrolyte
temperature
and level.Furthermore, the procedure did not contain acceptance
criteria for the specific gravity readings.This item is considered
unresolved
pending further NRC evaluation.
~Res onse: Plant Operating Procedure 9600.1 dated December 0, 1985 was revised to require specific gravity correction
for electrolyte
temperature
and level.This procedure was also revised to contain acceptance
criteria for the specific gravity readings.This procedure change had been planned prior to the inspection, but had not been completed.
sl.)~~
The original battery specification
and the specification
for the new batteries required a capacity of supplying the specified loads for one hour without the battery terminal voltage falling below 105 VDC.It should be noted that the testing of the batteries did not meet the intent of the specification
design requirements
nor the acceptance
criteria identified
in the FSAR.This appears to be another example of an inadequate
procedure since test procedures
for the initial testing of batteries 3A and 36 did not require load testing that verifies that the commitments
of the FSAR or the design specifications
were met.This item is considered
unresolved
pending further NRC evaluation.
~Res onse: FPL considers that the present one-half hour battery service test is adequate to demonstrate
that the.battery is capable of performing
its intended safety function.This is based on the assurance described in the Bases for the Technical Specification, that considering
any single failure, battery charging current should be supplied in one-half hour or less.However, we are evaluating
the discussion
in the FSAR to determine if the present battery test should be modified.
k g 4
URI 85-00-36: Mechanical
maintenance
personnel were uncertain regarding the type of grease-to be used in MOV gearboxes.
This was considered
a problem for two reasons.First, the mixing of different types of grease in the gearbox could cause hardening or separation
of the lubricant.
The potential for this exists at Turkey Point because its preventive
maintenance
instructions
for Limitorque
gearboxes specify the use of Texaco Marfac, while these same Limitorques
have been supplied with either Exxon Nebula EPO or EPI or Sun 50 EP lubricants.
Secondly, the only Limitorque
lubricant that meets the environmental
qualification
requirements
of 10 CFR 50.09 at Turkey Point is Exxon Nebula EPO or EPI.~Res onse: Exxon Nebula is, for MOV's inside containment, the only Limitorque (MOV'earbox)lubricant that currently meets the environmental
qualification
requirements
of 10 CFR 50.09.Preventive
Maintenance
instructions
for Limitorque
gearboxes, which specified the use of Texaco Marfac, have been pulled and are no longer in use at Turkey Point.The Lubrication
Manual's current revision specifies usage of only environmentally
qualified grease.This manual is now a PNSC reviewed document.MOV gearbox grease has been sampled and changed as appropriate.
Additionally, an Engineering
Evaluation
has been performed and forwarded to the Region II Administrator
on November 27, 1985 (L-85-008)
documenting
the justification
for prior limited operation with various greases in Limitorque
Valve Actuators.
In addition, instructions
were in the process of being revised to address this issue when the inspection
team reviewed this issue.The CEMS planners, who make up the work packages, were not uncertain about the type of grease to be used.Finally the grease guns used for maintenance
are now in a controlled
system..
A C~Ii
URI 85-00-37: During the AFW system walkdown, a Region II NRC inspector observed that the governor oil in the licensee's"A" AFW pump appeared to be different than that in the"B" and"C" turbine governors.
The inspector requested any licensee controls or historical
data along with vendor information
which would identify the control oil in the governors.
The licensee did produce the vendor manual with specific oil requirements;
however, the licensee had to have performed an oil analysis to determine the oil in the governors.
This analysis established
that a Texaco lOW'oil was being used in the"A" AFW governor and Texaco 30W in the"B" and"C" governor.The inspector expressed concern regarding whether the oil met vendor recommendations.
It was later determined
by the licensee that the Texaco 10W did not meet'vendor temperature
recommendations
for the licensee's
AFW operation.
This item is considered
unresolved
pending further NRC evaluation.
~Res onse: To address the concern regarding apparent difference
between the oil used in the"A" AFW turbine governor and that in the"B" and"C" turbine governors, we have sent samples for analysis to determine the specific oil used in each turbine governor.The results of the analyses identified
that the"A" AFW turbine governor contained an SAE 10 weight oil (similar to Texaco Regal 32 used on site)and the"B"'and"C" turbine governors contained an SAE 20 weight oil (similar to Texaco Regal 68 used on site).To determine the correct oil to use in the AFW turbine governors, it is necessary to define the governor oil operating temperature
range.Preliminary
data was taken during turbine operation on November 25, 1985 to determine governor oil operating temperature.
Comparing the data and specific oil weight to'information
provided in Woodward Governor Company Manual 25071C,"Oils for hydraulic controls" it was concluded that both oils are acceptable
for use in the AF W turbine governors.
To determine the optimum oil for use in the AFW turbine governors, we have developed a governor testing program which will provide additional
information
to be used for oil selection.
(~'~o v
IFI 85-00-38: For the AFW flow control valves to meet the requirements
of the present"FAIL-SAFE" test, they must be successfully
exercised full open, then closed from the control room with visual verification
at the valve.The power is not removed, nor is the air/nitrogen
supplies isolated.ASME Code Section XI, Division 1, IWV-3015, Fail-Safe Valves states, when practical, valves with fail-safe actuators shall be tested by observing the operation of the valves upon loss of actuator power.If these valves cannot be tested once every 3 months, they shall be tested during each cold shutdown;in case of frequent cold shutdowns, these valves need not be tested more often than once every 3 months.As of November 22, 1985, a request for engineering
assistance
was being generated by the Turkey Point Nuclear Technical Department
for the review of the present fail-safe testing method for these valves, identification
of inadequacies
in the fail-safe testing procedure, and development
of a method for proper fail-safe testing of these valves.This will be an inspector followup item.~Res onse: The auxiliary feedwater control values are designed to fail in the closed position by spring action on loss of air or power, For these valves to pass a fail-safe test in accordance
with IWV-3015 of Section XI of ASME boiler and pressure vessel code, valves with fail-safe actuators shall be tested by observing"operation
of these valves upon loss of actuator power.Present plant procedures
3/0-OSP-075.1 and 3/O-OSP-075.2
require these valves to be exercised full open, then closed from the control room with visual verification
at the valve to address the code requirement.
Additionally, the valves are verified in the closed position when the testing is terminated
and the steam supply MOV's (1003, 1000 and 1005)for that unit are closed which isolates power to the solenoids.
A review of the present fail-safe testing is being performed to assure compliance
with Section XI and subsequent
recommendation
for any testing modifications
will be provided by March 28, 1986.
K