IR 05000335/1987018
| ML17221A356 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 08/18/1987 |
| From: | Bibb H, Crlenjak R, Wilson B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML17221A355 | List: |
| References | |
| 50-335-87-18, 50-389-87-17, NUDOCS 8708270328 | |
| Download: ML17221A356 (15) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION REGION
101 MARIETTAST., N.W., SUITE 3100 ATLANTA,GEORGIA 30303 Report Nos.:
50-335/87-18 and 50-389/87-17 Licensee:
Florida Power and Light Company 9250 West Flagler Street Miami, FL 33102 Docket Nos.:
50-335 and 50-389 Facility Name:
St.
Lucie 1 and
License Nos.:
DPR-67 and NPF-16 Inspection Conducted:
July 5 - Aug st 1, 1987 Inspectors V~. Crlenjak, Sensor Resident Inspector
/
E.
ibb, Resident Insp Approved by:
. A. Wilson, Section Chief Division of Reactor Projects ate Signed 5'd'3'e igned C'zs'r Da e
S gned SUMMARY Scope:
This inspection involved onsite activities in the areas of Technical Specification compliance, operator performance, overall plant operations, quality'ssurance practices, station and corporate management practices, corrective and preventive maintenance activities, site security procedures, radiatio'n control activities, surveillance activities, and low temperature overpressure mitigation systems.
Results:
Of the areas inspected, no violations or deviations were identified.
One unresolved item was opened, paragraph 3.
8708270328 870818 PDR ADOCK 05000335
REPORT DETAILS Licensee Employees Contacted
- K. Harris, St.
Lucie Vice President
- G. J. Boissy, Plant Manager R. Sipos, Services Manager J.
H. Barrow, Operations Superintendent T. A. Dillard, Maintenance Superintendent
- J. B. Harper, QA Superintendent
"L. W. Pearce, Operations Supervisor R. J. Frechette, Chemistry Supervisor
"C.
F. Lappla, I&C Supervisor
- C.
AD Pell, Technical Staff Supervisor E. J. Wunderlich, Reactor Engi'neering Supervisor
"H. F.
Buchanan, Health Physics Supervisor G. Longhouser, Security Supervisor
- C. L. Burton, Reliability and Support Supervisor J.
Barrow, Fire Prevention Coordinator
- R. E.
Dawson, Assistant Plant Superintendent Electrical
- C. Wilson, Assistant Plant Superintendent Mechanical N.
G.
Roos, Quality Control Superviso)
Other licensee employees contacted included technicians, operators, mechanics, security force members, and office personnel.
- Attended exit interview 2.
Exit Interview The inspection scope and findings were summarized on August 6, 1987, with those persons indicated in paragraph
above.
The licensee did not identify as proprietary any of the materials provided to or reviewed by the inspectors during this inspection.
3.
Unresolved Item An Unresolved Item is a matter about which more information is required to determine whether it is acceptable or may involve a violation or deviation.
One Unresolved Item was identified -
389/87-17-01, paragraph 4.
4.
Plant Tours (Units 1 and 2)
The inspectors conducted plant tours periodically during the inspection interval to verify that monitoring equipment was recording as required, equipment was properly tagged, operations personnel were aware of plant conditions, and plant housekeeping efforts were adequate.
The inspectors also determined that appropriate radiation controls were properly
established, critical clean areas were being controlled in accordance with procedures, excess equipment or material was stored properly and combustible materials and debris were disposed of expeditiously.
During tours, the inspectors looked for the existence of unusual fluid leaks, piping vibrations, pipe hanger and seismic restraint settings, various valve breaker positions, equipment caution and danger tags, component positions, adequacy of fire fighting equipment, and instrument calibration dates.
Some tours were conducted on backshifts.
The inspectors routinely conducted partial walkdowns of ECCS systems.
Valve, breaker/switch lineups and equipment conditions were randomly verified both locally and in the control room.
During the inspection period the inspectors conducted a
complete walkdown in the accessible areas of the Unit
high and low pressure safety injection systems, Units 1 and 2; chemical volume control systems, diesel generator s, and AC electrical breakers to verify that the lineups were in accordance with licensee requirements for operability and equipment material conditions were satisfactory.
During a routine walkdown of the AC switchgear on Unit 2, the inspector noted that breaker 2-41381 (Reactor Building Maintenance Hatch Hoist Motor) was in the "on" position.
This appeared to be a repeat occurrence of a similar violation cited in report number 50 389/86 19, wherein breaker 2-41378 (Containment Elevator Fan Motor) was found "on" with the reactor at power.
This was a violation of the licensee's procedures which required that certain specified electrical power supplies which penetrate containment and do not meet the requirements of Regulatory Guide 1.63, Electrical Penetration Assemblies in Containment Structures for Light-Mater-Cooled Nuclear Power Plants, have their associated breakers maintained opened during power operations.
As of the end of this report period, it was not certain that breaker 2-41381 fits these requirements.
The licensee will maintain the breaker open while they complete thei r evaluation, this is considered an Unresolved Item (UNR 389/87-17-01).
Plant Operations Review (Units 1 and 2)
The inspectors, periodically during the inspection interval, reviewed shift logs and operations records, including data sheets, instrument traces, and records of equipment malfunctions.
This review included control room logs and auxiliary logs, operating orders, standing orders, jumper logs and equipment tagout records.
The inspector routinely observed operator alertness and demeanor during plant 'tours.
During routine operations, operator performance and response actions were observed and evaluated.
The inspectors conducted random off-hours inspections during the reporting interval to assure that operations and
security remained at an acceptable level.
Shift turnovers were observed to verify that they were conducted in accordance with approved licensee procedures.
The inspectors performed an in-depth review of the following safety-related tagouts (clearances):
Unit
Clearance No.
1"5-122 1-6-106 1-6-107
Waste Gas Compressor
[Preventive Naintenance]
18 Diesel Generator PM V-5205 PZR [Pressurizer]
Steam Space Sample Valve - Repack Unit 2 Clearance No.
Descri tion 2-11-27 2-6-70 2-6-100 2-7-3 2-7-15 2A S/G [Steam Generator]
Blowdown Sample Valve 28 Waste Gas Compressor Post Accident Sampling System - Change V-5763
ICW [Intake Cooling Water]
Pump Repack 2A Charging Pump - Charge Accumulator On July 25, Unit 2 turbine generator was taken off line, with the reactor maintained critical, due to a condenser tube leak.
The leak was caused by an internal condenser impingement plate which broke loose and fell on the tube, causing a break about one inch by one-eighth inch.
The tube was plugged along with two other tubes.
The plant was back on line on July 28.
A 30 percent power limit was imposed for a few days in order to clean up secondary chemistry to meet the Electric Power Research Institute (EPRI) guidelines.
6.
Technical Specification Compliance (Units 1 and 2)
During thi s reporting interval, the inspector verified compliance with limiting conditions for operations ( LCOs)
and results of selected surveillance tests.
These verifications were accomplished by direct observation of monitoring instrumentation, valve positions, switch positions, and review of completed logs and records.
The licensee's compliance with LCO action statements were reviewed on selected occurrences as they happene Maintenance Observation Station maintenance activities of selected safety-related systems and components were observed/reviewed to ascertain that they were conducted in accordance with requirements.
The following items were considered during this review:
limiting conditions for operations were met; activities were accomplished using approved 'rocedures, functional tests, and/or calibrations were performed prior to returning components or systems to service; quality control records were maintained; activities were accomplished by qualified personnel; parts and materials used were properly certified; and radiological controls were implemented as required.
Work requests were reviewed to determine status of outstanding jobs and to assure the priority was assigned to safety-related equipment.
The inspectors observed portions of the following maintenance activities:
Unit
PWO Ho.
0090/61 0227/61
~0 Build Security Barrier in Condensate Storage Tank Trench 1A ICW Pump Replace Expansion Joint Unit 2 2001/62 2C Auxiliary Feedwater Pump -
PM 2003/62 2A Charging Pump Accumulators -
PM 2004/62 2B Charging Pump Accumulators -
PM Physical Protection (Units 1 and 2)
The inspectors verified by observation and interviews during the reporting interval that measures taken to assure the physical protection of the facility met current, requirements.
Areas inspected included the organization of the security force, the establishment and maintenance of gates, doors and isolation zones in the proper conditions, that access control and badging was proper, and procedures were followed.
Surveillance Observations During the inspection period, the inspectors verified plant operations in compliance with selected technical specifications (TS)
requirements.
Typical of these were confirmation of compliance with the TS for reactor coolant chemistry, refueling water tank, containment pressure, control room ventilation and AC and DC electrical sources.
The inspectors verified that testing was performed in accordance with adequate procedures, test instrumentation was calibrated, limiting conditions for operations were met, removal and restoration of the affected components
were accomplished, test results met requirements and were reviewed by personnel other than the individual directing the test, and that any deficiencies identified during the test were properly reviewed and resolved by appropriate management personnel.
The inspectors observed portions of the following survei llances:
Proc No.
Descri tion 2-0010123 Appendix C Valve Switch Deviation Log 2-0030151 Remote Shutdown Monitoring Instrumentation Periodic Channel Check-2-0410050 HPSI/LPSI
[High Pressure Safety Injection/Low Pressure Safety Injection] Periodic Check 1-0110050 Control Element Assembly Periodic Exercise 01220026 Reactivity - Deviation From Design 1-0700050 Auxiliary Feedwater Per iodic Test 10.
Low Temperature Overpressure Mitigation Review (TI 2500/19)
This tempor'ary instruction (TI) was issued as a one time inspection module to verify that the pressurized water reactor (PWR)
licensees have an effective mitigation system for low-temperature overpressure (LTOP)
transient conditions in accordance with their commitments concerning Unresolved Safety. Issue (USI) A-26.
The, items to be,-..verified were divided into several areas:
design; administrative controls and procedures; training and-.
equipment modifications; and surveillance.
The format of this inspection report will be as follows:
the specific
,inspection requirement will be identified, followed by a discussion of the licensee's implementation which meets the requirement.
a.
"
Design (1)
Determination of whether there is documentation.to show that the overpressure protection system is designed to prevent exceeding the applicable technical specification and
CFR 50, Appendix G, limits for the reactor pressure vessel during.plant cooldown or startup.
For Unit 1, the required documentation is provided in the form of a study prepared by Combustion Engineering (CE),
Ines
, dated April 10, 1978, and included as Appendix 5B of the, Unit 1.FSAR.
For Unit 2, the licensee did their own analysis'nd.-this is discussed in paragraph 5.2.6 of the Unit 2 FSA Determination of whether there are drawings or sketches to show that the pressure protection system is designed to protect the vessel, given a single failure, in addition to a failure that initiated the pressure transient.
Sufficient electrical and mechanical drawings are in place for Unit 1 and Unit 2 to show redundant and independent operation of the two power operated relief valves (PORYs).
Determination of whether there are drawings or sketches to show that the system is not vulnerable to an event that causes a
pressure transient and a
failure of equipment needed to terminate the transient.
For Unit 1, the CE study provides analysis to support mitigation of a
pressure transient which assumes a
single failure that defeats one PORV.
Unit 2 FSAR paragraph 6.2.6.1.2 states:
"In the LTOP mode, the PORV overpressure protection system is designed to protect the reactor vessel given a single failure in addition to a failure that initiated the pressure transient.
The event initiating the pressure transient is considered to result from either an operator error or equipment malfunction.
The PORV system has redundancy in actuation channels and functions during a loss of offsite power.
When the PORVs are used to limit safety valve opening, at the high setpoint, redundancy is not required; therefore, only one PORV is aligned to the RCS, since this is an equipment protective rather than a
safety-related function."
Determination of whether there is documentation to show that the setpoints are supported by plant-specific analysis if the licensee uses a
PORY.
The PORV setpoints for Units
and 2 were reviewed (465 psia)
and found to be less than the highest allowable pressure (approximately 550 psia)
at lowest system temperature.
This information is contained in Chapter 5.2 of the FSAR.
(e)
Determination of whether the licensee's
CFR 50.59 evaluation has been prepared.
The Low Temperature Overpressure (LTOP) mitigating system for Unit 2 was a part of initial licensing and was adequately described in the FSAR.
The
CFR 50.59 evaluation for Unit
was submitted to the NRC by letter L-78-129 dated April 13, 1978, and was answered and accepted by the NRC in a letter dated September 6,
1983, issuing Amendment No.
60 to the facility Operating License.
The amendment revised the Technical Specifications to:
(1) incorporate limits and surveillance requirements associated with the overpressure mitigation system by the addition of new specifications that define the low
temperature reactor coolant system overpressure protection range; (2) incorporate a limit on the maximum primary-to-secondary differential temperature that is permitted prior to starting a
reactor coolant pump; (3) incorporate new requirements on the operability of power operated relief valves; (4) add a note to limit the establishment of a high pressure safety injection pump flow path under certain conditions; and (5) review requirements on the positioning of certain safety injection valves'b.
Administrative Controls and Procedures ( 1)
Determination of whether procedures exist for the following:
A (a)
To minimize,the time in a water-solid condition.
(b)
To minimize the temperature differentials between the steam generators and reactor vessel while in a
water-solid condition.
(c)
To restrict the number of high pressure safety injection charging pumps to no more than one when the reactor coolant system is in the low-temperature overpressure condition.
(d)
To alert operators to the automatic operation of the low-temperature overpressure protection system.
For item (a), the FSAR and other plant procedures were reviewed; however, no specific mention was made of minimizing the amount of time spent in water solid conditions.
However, operators are trained to inherently minimize the amount of time spent in any transient condition.
Additionally, precaution notes emphasize the need for extra care while water solid.
For items (b), (c),
and (d),
the following precautions contain the necessary precautions to address the specific concerns:
1-0030121, 2-0030121 Reactor Plant Heatup-Cold to Hot Standby (2)
Determination of whether the plant-installed system is in accordance with the plant license.
The installed equipment for both units meets the Technical Specification requirements, the plant license and the FSAR description.
Training and Equipment Modifications (1)
Determination of whether all operators have received training concerning RCS low-temperature overpressure event causes, the
operations and maintenance of the system that mitigates the events, and the consequences of inadvertent actuation.
The inspector reviewed the licensee's training program for new operators and the requalification program and found them to address these concerns adequately.
Determination of whether permanent modifications or procedural changes have been made that result in a
system that provides mitigation for RCS low-temperature overpressure events.
For example, has the system been modified to disable the pressurizer heaters and unneeded high pressure injection or charging pumps during cold, water-solid condition?
The most limiting transients initiated by a single operator error, or equipment fai lure are:
(a)
An inadvertent safety injection actuation (mass input).
(b)
A reactor coolant, pump start when a
positive steam.
generator to reactor vessel Delta T exists (energy input).
The transients were determined as most limiting by conservative analyses which maximize mass and energy additions to the RCS.
In addition, the RCS is assumed to be in a water-solid condition at the time of the transient; such a condition has been noticed to exist infrequently during plant operation since the operator is instructed to avoid water-solid conditions whenever possible.
Units
and
FSARs, figures S.b.7 and 5.2-24, respectively, show the results of the inadvertent safety injection actuation transient analysis when the RCS is in the LTOP mode.
The mass addition due to the simultaneous operation of the two HPSI and three charging pumps was considered, along with the simultaneous addition of energy from decay heat and the pressurizer heaters.
The LTOP system provided adequate system protection by relieving system pressure at 465 psia increasing, well below the design highest allowable pressure (approximately 550 psia)
at lowest system temperature.
Units
and
FSARs, figures 5.b.8 and 5.2-25, respectively, show that results of the transient analysis of a reactor coolant pump start when a
steam generator to reactor vessel Delta T of 100 F exists.
In addition to considering the energy addition to the RCS from the steam generator secondary side, energy addition from decay heat, reactor coolant pump, and pressurizer heaters was also included.
In the analysis, the steam generators were assumed to be filled to the zero power level, resulting in about
percent of the secondary water inventory being above the U-tubes.
For conservatism, the secondary water, both around and
above the U-tubes, was assumed to be thermally mixed in order to maximize the energy input to the primary side
~
As, a result of the temperature distribution within the steam generator during the transient, the water inventory above the tubes is thermally isolated from the heat transfer region.
Therefore, the heat transfer rate, and thus the primary side pressure, is not sensitive to the secondary side water level as long as the tubes are covered.
On the basis of experience,'he Delta T value of 100'
used in the analysis is larger than any Delta T that might be expected during plant operation.
During RCS cooldown using the shutdown cooling system, coolant circulation with the reactor coolant pumps serves to cool the steam generator to keep'he temperature difference between the reactor vessel and the steam generator minimal; Steam dumps are used to reduce steam generator secondary fluid to below 220'.
If the steam generator were held at 220'
and the reactor vessel, were cooled to the refueling temperature, the steam generator - reactor vessel Delta T would still be less than 100'.
In fact, procedures will direct the operator to maintain the Delta T
below approximately 20~
F.
LTOP transients have not been analyzed for the simultaneous startup of more than one reactor coolant pump (RCP).
Such operation is procedurally precluded since the operator starts only one RCP at a time and a
second RCP is not started until system pressure is stabilized.
Additionally, there is an LTOP transient alarm that indicates a
pressure transient is occurring.
Determination of whether there is documentation to show that modification to a specific piece of equipment will not result in the equipment being out of its design basis.
The addition of the LTOP system did not modify any existing equipment which result in being out of its design basis.
Determination of whether the instrumentation and control system incorporates an alarm with a
setpoint below the maximum allowable pressure for existing temperature conditions to alert the operator of a pressure transient.
The following is excerpted from the Unit 2 FSAR (Unit 1 is similar):
LTOP Systems Alarms a).
During cooldown when the
"Mode Selector Switch" is in the "Normal" position and cold leg temperature reaches a value 280 F,
a
"PORV LTOP Condition" alarm will alert the operator to select
"LTOP" on the
"Mode Selector Switch".
Thus, changing the PORV actuation setpoint from 2400 psia to 460 psia for V1474 and 490 psia for V1475.
This alarm will not reset unless the PORV mode selector switch is in the
"LTOP" position and the motor operated isolation valve of the associated PORV ss
"OPEN".
b).
PORV Normal Condition Alarm c).
During heatup when the
"Mode Selector Switch" is in the
"LTOP" position and cold leg temperature is greater than 320 F
a
"PORV Normal Condition" alarm will alert the operator to select
"Normal" on the
"Mode Selector, Switch".
Thus, changing the PORV actuation setpoint from 460 psia for V1474 (490 psia for V1475) to 2400 psia.
This alarm will not reset unless the PORV Mode Selector Switch is in the
"Normal" position and its associated isolation valve is
"OPEN".
Therefore, one alarm will be lit during normal operation as long as one PORV remains isolated.
When RCS temperature (cold leg) is less than 280 F or the mode selector switch is in "LTOP" and pressurizer pressure is greater than 460 psia, an LTOP transient is occur ring, actuating the
PORV Test Condition Alarm A
",PORV Test Condition" alarm will alert the operator whenever the PORV protective system is in the
"Over ride" or
"Test" position, bypassing all setpoints.
The PORV will remain closed in this condition until the selector switch is placed in the
"OFF" position at which time the alarm will reset.
Surveillance Determination of whether the PORV electronics and setpoints are verified periodically.
Determine the date of the most recent measurement of the PORV stroke times and compare the value with the design basis.
The Technical Specifications require verification of overpressure mitigation system (OMS) setpoints every 18 months.
This is done with IEC Pr'ocedure 1-1200054, 2-1200054
,
Overpressure Mitigation Functional Test every refueling.
The PORVs for Units
and 2 are solenoid operated valves with a
stroke time of 0.3 seconds.
The pump and valve program covers them and measures stroke times during each refueling, under Administrative Procedure 1-0010125A, 2-0010125A Schedule of Periodic Test, Checks and Surveillances, to assure that the stroke time is less than five (5) seconds.
(2)
Determination of whether tests are performed to ensure operability of the system electronics before each cold shutdown.
Operating procedures 1-0030127, 2-0030127 Reactor Plant Cooldown - Hot Standby to Cold Shutdown, contains the following statement under precautions and limits:
"Ensure a functional test has been performed on the PORV actuation channels per IKC Procedure 2-1200054,
"L8w Temperature Overpressure Protection Setpoint Verificatio'n".,
prior to cooling down to a
RCS temperature of 286 F."
(3)
Determination of whether, subsequent to system, valve or electronics maintenance, tests are performed before declaring the system operational.
The following is excerpted from Plant Administrative Procedure 0010432 - Nuclear Plant Work Orders:
"The Operations Department shall:
1)
Be responsible for ensuring that maintenance activities do not jeopardize the safety of the reactor.
They shall consider the possible sa'fety consequence of concur rent or sequential maintenance, testing, or operating activities.
Equipment required to be operable for the mode in which the reactor exists shall be available, and maintenance shall be performed in a
manner such that the license limits are not exceeded.
2)
Be responsible for assuring that post maintenance testing requirements recommended/specified will provide adequate verification that equipment will be capable of performing its intended function.
3)
Be responsible for acceptance of equipment following maintenance activities based on satisfactory post maintenance test completion."
No violations or deviations were identified in this are Functional Test every refueling.
are solenoid operated valves with The pump and valve program covers during each refueling, under 1-0010125A, 2-0010125A - Schedule Survei 1 lances, to assure that the (5) seconds.
The PORVs for Units
and
a stroke time of 0.3 seconds.
them and measures stroke times Administrative Procedure of Periodic Test, Checks and stroke time is less than five (2)
Determination of whether tests are performed to ensure operability of the system electronics before each cold shutdown.
Operating procedures 1-0030127, 2-0030127 Reactor Plant Cooldown - Hot Standby to Cold Shutdown, contains the following statement under precautions and limits:
"Ensure a functional test has been performed on the PORY actuation channels per I&C Procedure 2-1200054,
"Low Temperature Overpressure Protection Setpoint Verification",
prior to cooling down to a
RCS temperature of 286 F."
(3)
Determination of whether, subsequent to system, valve or electronics maintenance, tests are performed before declaring the system operational.
The following is excerpted from Plant Administrative Procedure 0010432 - Nuclear Plant Work Orders:
"The Operations Department shall:
1)
Be responsible for ensuring that maintenance activities do not jeopardize the safety of the reactor.
They shall consider the possible safety consequence of concurrent or sequential maintenance, testing, or operating activities.
Equipment required to be operable for the mode in which the reactor exists shall be available, and maintenance shall be performed in a manner such that the license limits are not exceeded.
2)
Be responsible for assuring that post maintenance testing requirements recommended/specified will provide adequate verification that equipment will be capable of performing its intended function.
3)
Be responsible for acceptance of equipment following maintenance activities based on satisfactory post maintenance test completion."
Ho violations or deviations were identified in this area.