W3P87-1775, Forwards Response to Generic Ltr 87-12 Re Operation of Reactor While RCS Partially Filled,Including Identified Improvements.Reactor Continues to Meet Licensing Basis
| ML20235E375 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 09/21/1987 |
| From: | Cook K LOUISIANA POWER & LIGHT CO. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| GL-87-12, W3P87-1775, NUDOCS 8709280088 | |
| Download: ML20235E375 (37) | |
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N USNRC-DS l OUISI ANA P O W E R & L I G H T! WATEFfQpfpEQ PM f$hd}O. KILLONA, LA 70066 ENPusileCI September 21, 1987 W3P87-1775 A4.05 QA U.S. Nuclear Regulatory Commission ATIN: Document Control Desk Washington, D.C. 20555
Subject:
Waterford SES Unit 3 Docket No. 50-382 License No. NPF-38 Response to Generic Letter 87-12 Gentlemen:
As requested in Generic Letter 87-12, LP&L has reviewed the operation of Waterford 3 during partially filled reactor coolan e system conditions. The results of our review, as well as identified improvements, are attached.
The review conducted in responr.e to Generic Letter 87-12 was beneficial in consolidating and clarifying the Waterford 3 approach to partially filled operation. Consequently, we are able to conclude that Waterford 3 continues to meet its licensing basis.
Should you require additional information please feel free to contact me or Mike Meisner at (504) 595-2832.
Very truly yours, K.W. Cook Nuclear Safety &
Regulatory Affairs Manager Attachments: Affidavit Response to GL 87-12 KWC:MJM:pmb
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! cc: E.L. Blake, W.M. Stevenson, J.A. Calvo, J.H. Wilson, R.D. Martin, 1
NRC Resident Inspectors Office, F.J. Miraglia, W. Lyons 8709280088 870921 ' I DR ADOCM 05000382 PDR I[
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l W3P87-1775 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the matter of )
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Louisiana Power & Light Company ) Docket No. 50-382 Waterford 3 Steam Electric Station )
AFFIDAVIT K.W. Cook, being duly sworn, hereby deposes and says that he is Nuclear Safety & Regulatory Affairs Manager of Louisiana Power & Light Company; that he is duly authorized to sign and file with the Nuclear Regulatory Commission the attached response to NRC Generic Letter 87-12; that he is familiar with the content thereof; and that the matters set forth therein are true and correct to the best of his knowledge, information and belief.
K.W'. Cook Nuclear Safety &
Regulatory Affairs Manager STATE OF LOUISIANA)
) ss PARISH OF ORLEANS )
Subscribedandsworntobeforeme,aNotaryPub/erif/fsdr4/icfnandfortheParish and State above named this ._#/ df day of t , 1987.
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hw kL ov NotaryPublic/
My Commission expires (Ti (Ifn d .
l W3P87-1775 WATERF0RD 3 RESPONSE T0 GENERIC LETTER 87-12 l
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SEPTEMBER, 1987
TABLE OF CONTENTS SECTION TITLE PAGE 1.0
SUMMARY
OF RESULTS 1 2.0 APPROACH TO GENERIC LETTER RESPONSE 3 3.0 SHUTDOWN COOLING SYSTEM DESCRIPTION 4 4.0 ENTRY TO PART-LOOP OPERATION 7 4.1 Drain Down Process 7 4.1.1 Prerequisites for RCS drain down 7 4.1.2 Procedure for RCS drain down 8 4.2 Completion of Drain Down 9 4.3 Time to Drain Down 9 5.0 PART LOOP OPERATION DUXING NORMAL CONDITIONS 11 5.1 Operating rrocedures and Administrative Controls 11 5.2 Instrumentation 12 5.2.1 Level 12 HJTC System 12 Refueling Level Indication System (RLIS) 13 Refueling Water Level Indication System (RWLIS) 17 5.2.2 Temperature 17 5.2.3 Pressure 19 i
5.7.4 LPSI Pump Parameters 19 5.3 Interlocks 19 I 5.4 Vortexing 20
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TABLE OF CONTENTS (Continued)
SECTION TITLE PAGE 6.0 PART-LOOP OPERATION DURING LOSS OF SHUTDOWN COOLING 22 6.1 Operating Procedures and Administrative Controls 22 6.2 Instrumentation 23 6.2.1 Level 23 6.2.2 Temperature 24 6.2.3 Pressure 24 6.2.4 LPSI Pump Parameters 24 6.3 containment Isolation 25 6.4 Pumps Available to Control RCS Inventory 25 7.0 GENERIC LETTER 87-12 POSTULATED LOSS OF SHUTDOWN COOLING SCENARIO 27 7.1 Loss of Shutdown Cooling Scenario Progression at Waterford 3 27 7.2 Time to Core Uncover
- 28 7.3 Significance of Postulated Scenario for Waterford 3 19 7.4 Steam Generator Availability During Refueling Outages 29 8.0 TRAINING 31
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SUMMARY
6F RESULTS At the direction of Waterford 3 management, a multi-disciplinary task force l has conducted a review of plant operations during partially drained RCS conditions. Based on this in-depth review and implementation of improvements discussed in the remainder of this submittal, we are confident that Waterford 3 part-loop operation meets the license basis of the plant and poses no undue risk to the public health and safety.
In reaching this conclusion, consideration was given to the adequacy of normal and off-normal operating procedures, design and performance of the shutdown cooling system, instrumentation and controls available to the operator to maintain shutdown cooling and to mitigate a loss of shutdown cooling event, and operator training. While these facts were generally found to be acceptable, improvements have been identified to increase operator capability and flexibility in dealing with a plant upset during part-loop operation. Changes range from a minor study to confirm RCS drain down levels for various maintenance activities to a major revielon of the Loss of Shutdown Cooling procedure.
Of primary concern during our review was the capability of the Waterford 3 design to respond to a limiting loss of shutdown cooling event. In this regard the task force was guided by the event postulated in Generic Letter 87-12 wherein an opening exists in the RCS (e.g. Reactor coolant pump seal replacement) while drained to the hot leg midpoint (see Section 7). '
Calculations were performed (and later confirmed through parallel work by the CE Owners Group) leading to several important conclusions:
- 1. Reactor Coolant Pump (RCP) design is such that RCS pressurization of approximately 60 psig is necessary before significant water loss could occur when an RCP seal is removed.
- 2. For a closed RCS (including RCP seal replacement) steam generator boil dry time (with no makeup)is approximately 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> with a maximum RCS pressure of approximately 35 psig.
- 3. For an open RCS (major disassembly or removal of an RCP) core uncovery will not occur for 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />, which includes the very conservative assumption that no primary system steam is condensed in the steam generator (s).
- 4. Steam generator initial level and makeup capability are the key i parameters in extending the period prior to core uncovery.
- 5. In addition to restoring shutdown cooling, the event may be terminated at any time prior to core damage by high pressure safety injection (HPSI) flow.
As a prerequisite to RCS drain down, Waterford 3 procedures require availability of a HPSI pump. It has also been our practice to maintain level in the steam generator (s) during part-loop operation when steam generator maintenance is not being performed.
4 To provide additional assurance of the capability to mitigate the scenario postulated in Generic Letter 87-12, the following conditions for part-loop operation are being implemented in appropriate Waterford 3 procedures:
- 2. At least one steam generator is operable with level ct 70% wide range or greater when steam generator maintenance is not being performed, and
- 3. A secondary side steaming path is available.
When steam generator maintenance is necessary, the time needed to complete the work vill be minimized to limit steam generator unavailability.
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j 2.0 APPROACH TO GENERIC LETTER RESPONSE In recognition of the importance of adequate design, training, !
procedures and instrumentation during part-loop operation, Waterford 3
[ management directed the creation of a multi-disciplinary task force to l- review Generic Letter 87-12. The task force consisted of one or more L members from the operations, engineering, tachnical support, training, l
licensing and analysis groups and received additional support, as i needed, from other organizations.
The task force adopted as its goal performance of an integrated review of Waterford 3 and industry operations during partially filled RCS
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conditions. As a result, the scope of the task force review i encompassed not.only Generic Letter 87-12, but also previous NRC and )
industry critiques of applicable loss of shutdown cooling events including the July, 1986 loss of shutdown cooling event at Waterford
- 3. Although reviews of such events had been conducted (and improvements implemented) at Waterford 3 on a case-by-case basis in the past, the task force felt that there was benefit in performing a l comprehensive review to-integrate the previous efforts.
The task force has identified a number of improvements, primarily administrative and procedural, which will be implemented prior to or ,
during the cecond refueling outage for Waterford 3. These changes are described in the appropriate sections below.
In addition to the Waterford 3 task force, LP6L has participated in the CE Owners Group (CEOG) generic review of the core damage scenario initiated from partially drained conditions as postulated in Generic Letter 87-12. The results of the CEOG review coupled with Waterford 3-specific conditions and system responses are discussed in Section 7.
3.0 SHUTDOWN COOLING SYSTEM DESCRIPTION A full description of the Shutdown Cooling System (SDCS) is contained in the Waterford 3 FSAR Section 9.3.6. For the purposes of the Generic Letter 87-12 response, a simplified diagram of the SDCS with approximate elevations is shown in Figure 1.
During shutdown cooling operation, the reactor coolant is diverted from the RCS via the shutdown cooling nozzle located in each hot leg. The flow is then cooled by circulating through a shutdown cooling heat exchanger via a low pressure safety injection (LPSI) pump. Two independent and redundant shutdown cooling trains are available.
The cooled flow returns to the RCS through the safety injection lines connected to the cold legs. Plant cooldown rate is controlled by two flow control valves which permit proportioning the amount of shutdown cooling flow passing through the heat exchanger and heat exchanger bypass line.
A vacuum priming pump is connected to the shutdown cooling suction line high point to aid in removing entrapped air in the system.
SDCS components are provided with overpressure protection devices where design pressure and temperature are less than the RCS design limits. Each shutdown cooling suction line is equipped with three isolation valves in series. With this arrangement a redundant, parallel shutdown cooling path is available should a single failure preclude the availability of one of the shutdown cooling trains.
Each valve inside containment is provided with an interlock to prevent opening and to initiate automatic closure whenever the RCS pressure exceeds 392 psia (to prevent opening) and 700 psia (to initiate closure). These interlocks are discussed further in Section 5.3.
A pressure relief valve in each shutdown cooling suction line protects the system from overpressurization during system operation when the suction valves are open and the system is not isolated from the RCS. The valves are sized considering transients due to the simultaneous inadvertent operation of all charging pumps, HPSI pumps, and pressurizer heaters.
Additional pressure relief valves are provided to protect isolated sections of piping from thermally induced overpressure.
Piping and valves are provided in the Chemical and Volume Control System (CVCS) such that during shutdown cooling a portion of the flow can be bypassed from the outlet of the shutdown cooling heat exchangers through the letdown portion of the CVCS and return to the shutdown cooling suction lines. Flow through this bypass stream provides filtration and ion exchange of the reactor coolant, provides monitoring of boron level via CVCS components and provides the primary RCS drain path (to the Holdup Tanks) to initiate part-loop operation. RCS draining may also be accomplished, when necessary, through the LPSI pump minimum flow recirculation line connection to the Refueling Water Storage Pool (RWSP).
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Operation of the SDCS is controlled and monitored through the use of installed instrumentation. The instrumentation provides the capability to determine heat removal, cooldown rate, shutdown cooling flow and the capability to detect degradation in flow or heat removal capacity. The instrumentation provided for the SDCS.in addition to those normally available for RCS monitoring consists of: a) Temperature measurements - shutdown cooling heat exchanger inlet and outlet temperature and the temperature of the shutdown cooling flow to the low pressure safety injection header. All i temperatures are indicated in the control room. The shutdown cooling heat exchanger inlet temperatures and the low pressure safety injection header temperature are recorded to facilitate control of the RCS cooldown rate. b) Pressure measurements - LPSI header pressure and shutdown cooling heat exchanger inlet pressure. These pressures are indicated in the control room and, when used with the low pressure pump performance curves, provide an alternate means of measuring system flow rates. c) Total shutdown cooling flow rate is measured by the shutdown cooling flow indicator in the control room. With the aid of this flow indicator, the operator ensures that sufficient flow for decay heat removal is maintained. 4.0 ENTRY TO PART-LOOP OPERATION 5 RCS draining is initiated to provide a water-free work location for l various maintenance and refueling activities including: Pressurizer heater removal Reactor head removal Safety injection check valve repair Steam generator tube inspection / repair RTD (resistance temperature detector) repair RCP seal replacement The RCS may be drained to different levels depending upon the type of maintenance to be performed. For instance, pressurizer heater removal requires minimal draining to empty the pressurizer (to approximately 23' MSL-Mean Sea Level) whereas steam generator tube inspection requires a drain down to the hot leg centerline (13' 4 " MSL). The discussions which follow are primarily concerned with part-loop operation, i.e. operation at RCS levels below the top of the hot leg (15' 1h" MSL). 4.1 Drain Down Process J The RCS drain doeu process is governed by Operating Procedure OP-1-003, Reactor Coolant System Drain Down. 4.1.1 Prerequisites for RCS drain down Prior to initiating draining, the appropriate portions of the Shutdown Cooling System (SDCS), Chemical and Volume Control System (CVCS), and Baron Management System must be in service. Waterford 3 also requires one train of High Pressure Safety Injection (HPSI) to be in service as a prerequisite for draining the RCS. The HPSI availability, as will be discussed in more detail later, is a major element in the mitigation of the potential adverse effects of a loss of shutdown cooling event. Although not presently proceduralized, it is the practice at Waterford 3 to maintain secondary steam generator levels high in the normal measurement band, or in wet layup, unless the steam generator is required to be drained for maintenance. Recognizing the benefits of maintaining a heat sink for a potential loss of shutdown cooling event during part-loop operation (see Section 7), Waterford 3 will include in OP-1-003 a requirement that at least one steam generator be maintained operable (i.e. minimum level greater than or equal to 70% wide range level) during the drain down and part-loop operation, at times when steam generator maintenance is not being performed. Means will also be provided to steam to the atmosphere, or condenser if available (see Section 6.1). Prior to beginning the drain down, certain initial conditions must be satisfied. In the context of Generic Letter 87-12, the key initial actions include:
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a) Insta11ation'and line up of the'RCS refueling level indicator (see Section 5.2.1 for details of level indication), b) Installation of vent rigs on both LPSI pumps to facilitate venting of pump casings in the event of air intrusion, and c) Placing the vacuum priming system in operation for both shutdown cooling loops. 4.1.2 Procedure for RCS drain down The RCS is normally drained through the CVCS to the Holdup Tanks. Upon obtaining permission from the Operations Superintendent, draining may occur to the Refueling Water Storage Pool, if necessary. A number of cautions and limitations apply during the drain down. Amongst others, these include: a) Draining shall be at a rate which precludes drawing a vacuum on the RCS, b) Communications shall be maintained between personnel monitoring the local refueling level indicator and Control Foom personnel during the draining process, c) When the SDCS is in operation, RCS level shall be maintained greater than the hot leg center line, d) Two SDCS loops shall be operable and at least one SDCS loop shall be in operation with RCS loops not filled (Technical Specification 3.4.1.5), e) Simultaneous draining via two or more paths is not permitted, f) Heated Junction Thermocouple (HJTC) level indication (see Section 5.2.1) shall be monitored continuously while draining, and frequently while the RCS is partially drained, but shall never be used as a sole means of level indication, g) If an operating LPSI pump begins to cavitate, draining shall be secured, and filling begun by any means possible (charging pump, LPSI or HPSI pump), h) Prior to completely draining the pressurizer, the refueling level indicator is to be cross checked with the pressurizer level, and l i) When level is approximately at the top of the hot leg, draining is to be secured temporarily to ensure a
" good" level reading (by allowing the steam generator l tubes to fully drain), and LPSI pump flow is to be l reduced to approximately 3000 gpm.
4.2 Completion of Drain Down l As previously noted, the ultimate RCS level attained in the drain down is a function of the maintenance activities planned for the outage l and in no case is lower than the centerline of the hot leg (13' 4 " MSL). Procedure OP-1-003 currently lists the following levels that should be maintained for each maintenance activity: Maintenance Work Level (MSL) Pressurizer heater removal 23' 2 " - 23' 4 " Reactor head removal 18' 10 " - 19' h" SI check valve repair 15' h" - 15' 4 " Top of hot leg 15' 1 " SG tube inspection / repair 13' 4h" RTD repair 13' 4 " RCP seal replacement 13' 4 " Waterford 3 is conducting an engineering review to ensure that the listed levels are consistent with the maintenance work to be performed and to determine acceptable operating ranges at mid-loop. Should changes be necessary they will be incorporated into OP-1-003 upon completion of the review. 4.3 Time to Drain Down The time necessary to drain the RCS after a reactor shutdown is strongly dependent on the circumstances of the shutdown and the maintenance work planned for the outage. For the purposes of Generic Letter 87-12, a scenario has been assumed to determine the minimum time needed to drain the RCS to the hot leg centerline. In this scenario a reactor trip has occurred from 100% power and plant management has decided to cool down and drain down to replace an RCP seal. In the sequence of events below, the time assumed for each step where a reference is not noted is estimated based on experience from past shutdowns. The time to complete OP-902-002, Uncomplicated Reactor Trip procedure, and to enter OP-10-001, General Plant Operations procedure is 30 minutes. The time to sample and analyze RCS and pressurizer boron concentration is 30 minutes. The time to cool down to 350'F at the maximum cool down rate of 100*F/hr is approximately 2 hours. The most restrictive portion of the RCS cool down is the degassing of the RCS. Assuming that degassing was initiated at the start of cool down and performed in conjunction with the cool down of the pressurizer, the time for completion would be approximately 20 hours. This includes the pressurizer solid and depressurized and the RCS degassed to less than 5 cc/kg hydrogen and cooled down to 200'F. The time to cool down to less than 140*F (a prerequisite for draining) at 30*F/hr would be approximately 2 hours. Also included in this time is a containment entry by Health Physics personnel for survey purposes. I After containment is opened for access, vent rigs would be installed on the pressurizer and the local level indicator would be placed in service in approximately 2 hours. The time to drain the RCS from 100% pressurizer level, assuming 100 ' gpm through letdown, and allowing time to stabilize level at the top of the hot leg (as steam generator tubes drain), would be approximately 10 hours. The minimum estimated time to drain to the hot leg centerline following a reactor' trip is, therefore, 37 hours. The above scenario is only for purposes of estimating a conservatively short drain time. In practice, additional factors will extend the drain time: The cooldown rate is normally 50-60*F/hr maximum to 200*F which would increase the total drain time approximately 3-4 hours. Depending on the initial amount of hydrogen in the RCS, degassing could take significantly longer. The combination of cooldown of the pressurizer and degassing could take up to 12 hours longer than the 20 hours estimated above. For refueling outagea Waterford 3 adds hydrogen peroxide to the RCS for cleanup purposes. As a prerequisite to the hydrogen peroxide addition, the RCS must be fully degassed at a temperature less than 200*F. The cleanup process takes approximately 24-48 hours which could increase the total drain time for extended cleanup. Normal delays in coordination, and administrative controls on entering containment (e.g. if containment atmosphere was not yet habitable for work after shutdown) could significantly extend the drain time. A realistic drain time under normal conditions, therefore, is in the range of 48-60 hours or longer. l l
5.0 PART LOOP OPERATION DURING NORMAL CONDITIONS 5.1 Operating Procedures and Administrative Controls During normal drained conditions the operators' attention is directed c towards monitoring and maintaining the desired RCS level while l maintenance work activities are being completed. Operational instructions are provided by Operating Procedure OP-1-003, Reactor Coolant System Drain Down. RCS local level indication must be monitored at least once per shift == (as discussed in Section 5.2.1, Waterford 3 intends to install additional control room level indication during the second refueling l outage subject to component availability). Proper operation of the 4 shutdown cooling loop vacuum priming system is also monitored at least once per shift. During extended operation in a drained down condition, operators are directed to periodically vent the LPSI pumps to ensure continued smooth operation. In recognition of their importance to normal part-loop operation, level indication monitoring ~ and LPSI venting (when maintaining level at mid-loop) will be fnereased to twice per shift in a revision to OP-1-003. , Should RCS level increase 4" above the levels required for the l specified maintenance activity, the RCS is to be drained down to the l prescribed level. Should RCS level decrease 4" below the prescribed level (or below the hot leg center 2ine), operators are directed to restore level by: 1
- a. Starting the available charging pump (s) and filling the RCS to the desired level, or
- b. If a faster fill rate is required, use the standby HPSI pamp and open its associated flow control valve to fill to the desired level.
Should a loss of shutdown cooling event occur, operators are required to address the event through Off-Normal Operating Procedure OP-901-046, Loss of Shutdown Cooling (see Section 6). Operations and testing that could perturb the NSSS during and following drain down are presently controlled by the onshift Shift Supervisor. Any maintenance activities that could jeopardize operation during partially drained conditions are scheduled by the Planning and Scheduling Department with the concurrence and control of the Operations Department. Maintenance activities on the shutdown cooling system are not scheduled during refueling outages until the reactor cavity floodup is performed (in Mode 6) with greater than 23 feet of borated water above the reactor vessel flange. In addition, maintenance is not scheduled on equipment required operable by Technical Specifications or administrative control (e.g. a HPSI pump l required for RCS drain down). Coordination of such operations is controlled by the Planning and Scheduling Operations Coordinator who is normally an offshift Shift Supervisor. Prior to planned maintenance work commencing, the work package must be reviewed and approved by the onshift Shift Supervisor. As part of its overall refueling procedure,
f j .s i Waterford 3 has also recently added a position of Containment l Controller for coordination of' containment activities during refueling.- This position is presently intended to be filled by a, individual holding an SRO license. 1 To provide more firm administrative control over testing and maintenance activities during part-loop operations, various procedural changes will be implemented:
- a. '0P-1-003, Reactor Coolant System Drain Down, will contain a i
limitation to more closely control maintenance which may I have'the potential to reduce RCS inventory; or , delay or prevent prompt containment isolation.
- b. PE-5-002, Local Leak Rate Test, will contain a precaution to inform the Shift Supervisor / Control Roo'm Supervisor prior to initiation of any testing which could result in RCS inventory loss of affect the capability to isolate i containment.
- c. OP-903-033, Cold' Shutdown ISI Valve Tests, will contain a precaution that performance of ISI testing on valves that could affect containment isolability, the capability for shutdowr.1 cooling, or whose misoperation could lead to RCS inventoly loss, shauld not normally be allowed.
5.2 Instrumentation Generic Letter 87-12 emphasizes the role of instrumentation in assisting operators to maintain adequate shutdown cooling during part-loop operation. The following Subsections review the instrumentation important to maintaining shutdown cooling at Waterford 3 during normal'part-loop operation. The effects on instrumentation' of a loss of shutdown cooling event during part-loop operation'are discussed in Section 6.2. 5.2.1 Level During part-loop operation Waterford 3 operators utilize two independent and diverse RCS level measurement systems - the Heated Junction Thermocouple (HJTC) System and the Refueling Level Indication System (RLIS). HJTC System The HJTC System (also referred to as the Reactor Vessel Level Monitoring System - RVLMS) is described in detail in Appendix 1.9A of the Waterford 3 FSAR, therefore, only a brief summary description is provided here. The principal funct_ ton of the HJTC System is to measure the water inventory in the reactor vessel above the fuel alignment plate. This is done at discrete elevations by monitoring the temperature difference between adjacent heated and unheated thermocouple.
. . q Eight HJTC sensors are placed at specific elevations (see Figure 2) inside a separator tube to make up a probe assembly. Two redundant probe assemblies are installed. The purpose of the separator tube is to create a collapsed water level inside while a two-phase mixture exists outside the tube. When the collapsed water level falls below a heated junction elevation, its temperature and the sensor differential temperature increase above a predetermined setpoint value. The sensor is then identified as being uncovered (i.e. surrounded by steam).
The HJTC sensor information is provided ou redundant control room displays as percent liquid level in the upper head, percent liquid level in the upper plenum and the covered / uncovered status of each sensor. The HJTC sensor elevations in the upper plenum provide useful information to the operators during part-loop conditions. For instance, three sensor locations closely approximate the top, center , and bottom of the hot leg as shown below. Upper Plenum HJTC Sensor Elevation (MSL) Comment 17'3" 15'3" Top of hot leg - 15'l " 13'6" Center of hot leg - 13'4 " 11'9" Bottom of hot leg - 117 " 10'1" As a result, cross cliecks between the HJTC System and other level measurement systems at appropriate elevations can serve to confirm level measurement accuracy within the resolution capability of the HJTC System. The HJTC System can also serve as a backup in the event of loss of other level measurement capability.' Finally, the sensor located at the hot leg centerline provides a yes/no indication of the potential for LPSI pump cavitation. Refueling Level Indication System (RLIS) The RLIS, which provides local RCS level indication during Modes 5 and 6 operation, consists of one inch rubber (ortac) tubing, including a length of hardened Tygon tube used as a sight glass, in a standpipe arrangement (refer to Figure 3). The tubing runs from the bottom of hot leg #1, through the sight glass, and into the top of the pressurizer. Quick connects are provided at the hot leg and pressurizer connections, as well as the upper and lower ends of the sight glass, for ease of installation and maintenance. A drain is included in the low point of the system. The level scale is permanently mounted outside the pressurizer wall, therefore, any movement in the tubing will not cause an error in indicated level. The scale covers a wide range including the bottom of the hot leg to the top of the scale at 30' MSL (approximately 5' above the bottom of the pressurizer). During installation, the centerline of the hot leg (13.375' MSL) was surveyed and marked on the outside of the pressurizer wall. That marking served as a reference for mounting the permanent scale.
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VESSEL OUIC K CONNECT _ it ',. _ v6 ( r WECTION I NE LOW POINT DRAIN r RUBBER HOSE $ a l The scale is marked at every inch and labeled at every foot. Special elevations are written on the scale: 15.125' (Top of hot leg) i 13.375' (Hot leg centerline) In addition, vertical red highlight bars are drawn at the following l elevation ranges to correspond with the allowed range for maintenance l work (see Section 4.2): 23' 2h" to 23' 4 ""(Pressurizer heater removal) l 18' 10 " to 19' (Reactor head removal) l 15' " to 15' 4 " (SI check valve repair) The RLIS scale high point overlaps the cold calibration level transmitter for pressurizer level. As previously noted, prior to draining below the pressurizer, the two level indications must be cross checked. During installation, the tubing length of the RLIS was maintained to a minimum. Care was exercised in the design and installation to preclude problems due to crimped tubing, small diameter tubing, local high spots, etc. The tubing, hoses and quick connects have an inside diameter of one inch. Since the tubing is a large diameter, level measurement time response is almost instantaneous (less than one second). The RLIS nozzle is located on the same hot leg as the pressurizer and shutdown cooling suction for LPSI pump B. The nozzle is approximately 30 inches upstream of the shutdown cooling suction nozzle The LPSI pump, when running at full flow, has a flow rate of 9.0 ft /second past the nozzle to the level indication. The low flow rate results in a negligible error (decrease) in indicated level. Based on the review of Generic Letter 87-12 Waterford 3 vill implement changes to OP-1-003, RCS Drain Down to: a) Visually inspect, prior to each drain down, RLIS hose and tubing for anomalies which could affect the accuracy of level indication (high points, kinking, brittleness, leaks, clarity of Tygon tube, tube plugging, etc.), and b) Not remove the HJTC System from service (when the reactor vessel head is in place) and perform a cross check for level indication with the RLIS.
Refueling Water Level Indication System (Second Refueling Outage) To address the concerns with temporary local level indicating eystems Waterford 3 is expending extensive effort to implement a permanently installed Refueling Water Level Indication System (RWLIS) with control room indication, by the end of the second refueling outage provided-procurement lead-time is adequate to support this schedule. Should component procurement not he possible, the RWLIS will be installed by the end of the third refueling outage. The design and engineering of this system is presently under development. Consequently, the following description should be considered conceptual and subject to change. The RWLIS will be comprised of a narrow and a wide range differential pressure transmitter attached, through stainless steel piping to a primary system high point near the top of the pressurizer and to the hot leg drain on the RCS (see Figure 4). The signals from these transmitters will be provided to a local indicator in the vicinity of the presently installed sight glass inside containment, and to a remote indicator in the control room. The RWLIS will be designed to operate with the reactor vessel head removed or installed. With the exception of the local RWLIS indicator in containment, the syatem is designed for permanent installation. The RWLIS will compensate for slight positive and negative pressure variations in the RCS by utilizing a differential level instrument arrangement to compare actual RCS water level to a reference leg vented to an optimum system high point. With this approach a "true" steady state RCS water level reading will be provided. The low level sensing line will be permanent 1y' attached to the drain line just below hot leg #1 in the vicinity of the shutdown cooling suction line. This is the same location used by the RLIS. The transmitters and the local indicator will be rerouted on an instrument stand outside the biological shield as close to the pressurizer as possible to minimize tubing runs. The inatrument stand will be located nesr the RLIS level indication for ease of level comparison. The location will also allow easy draining of the reference leg to existing RCS drain pipe to maintain the reference leg dry during normal operation. The reference leg will be permanently attached to the existing upper tap of the pressurizer wide range level instrument. The level signals will be conducted to a control room panel and annunciator window. In accordance with design change procedures, a human factors evaluation will be performed to place the indicator in its optimum location. 5.2.2 Temperature RCS temperature indication during normal part-loop operation is provided by the standard T (hot) and T (cold) RTDs and Core Exit Thermocouple - CETs (when the reactor vessel head is in place). Temperature indication from the CETs is discussed in FSAR Section 1.9A.3.3. It should be noted that RTDs are located in thermowells at 45* angles to the vertical. The RTD response time to changes in liquid temperature will vary depending on RCS level. I L __ _ _ __ __
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t l To ensure availability of CET information, Operating Procedure OP-1-003, Reactor Coolant System Drain Down, will be revised to require at least 2 CETs (normally from independent trains) in service when the reactor vessel head is in place, except when preparing for head removal or replacement. 5.2.3 Pressure RCS pressure indication is available from the standard safety-related pressurizer pressure indicators and LPSI pump discharge pressure described below. 5.2.4 LPSI Pump Parameters The following control room indicators are available to monitor LPSI pump performance and shutdown cooling flow. Indicator Range LPSI pump motor current 0-100 amps LPSI flow 0-5500 gpm LPSI discharge pressure 0-650 psig LPSI suction pressure 0-30 psig (local indication only) With the exception of LPSI suction pressure, these parameters are also available through the Plant Computer for trending purposes. The following control room alarms also exist to indicate problems with the LPSI pumps: LPSI Pump Unavailable LPSI Pump Trip / Trouble LPSI Pump Flow Lost LPSI Pump Bearing Water Flow Low LPSI Pump Min. Flow Isol. Valve Upset 5.3 Interlocks Waterford 3 has reviewed any interlocks that could cause a disturbance to the RCS during and following a drain down and has identified the shutdown cooling system (SDCS) autoclosure interlock as having a potential to interrupt shutdown cooling flow. The SDCS interlocks, which are described in detail in FSAR Section 7.6.1.1, consist of an open permissive signal to allow opening the shutdown cooling isolation valves below an RCS pressure of 392 psia, and an autoclosure signal to close the isolation valves when RCS pressure exceeds 700 psia. The interlock pressure setpoints are selected such that the design pressure of the SDCS is not exceeded. Each of four SDCS isolation valves (2 per SDCS train) are interlocked by one of the four independent pressurizer pressure measurement channels. When RCS pressure exceeds 392 psia and an interlocked SDCS isolation valve is not fully closed, an alarm is sounded in the control room. Autoclosure occurs when RCS pressure exceeds 700 psia. Although none have occurred at Waterford 3, industry experience indicates that spurious SDCS isolation valve autoclosures do occur. To minimize the potential for autoclosure during shutdown cooling, Operating Procedure OP-10-002, Refueling, directs that the autoclosure interlock be inhibited following removal of the reactor vessel head. Further, Waterford 3 believes there may be a good technical basis for disabling the SDCS autoclosure interlocks (ACI) during Mode 5 operation, as discussed below. , Previous experience with spurious ACI actuations causing a loss of l shutdown cooling have been documented in the NRC's Office for Analysis and Evaluation of Operating Data case study report AEOD/C503
.(December, 1985), and the Nuclear Safety Analysis Center report NSAC-52 (January, 1983). Both reports recommend removal of the autoclosure interlocks whenever valve motion is not necessary as long as adequate low temperature overpressure (LTOP) protection is available to protect the SDCS. Waterford 3 has two LTOP relief valves (one per SDCS train) each designed to protect the SDCS from overpressure due to the simultaneous combination of all backup pressurizer heaters. energized, three charging pumps actuated, and both HPSI pumps running (a total of over 3000 GPM relief capacity per valve). Hence, Waterford 3 has adequate LTOP relief capacity.
l l In addition to the ACI and LTOP relief capacity, Waterford 3 maintains administrative and procedural control over the SDCS isolation valves to ensure they are closed prior to exceeding the SDCS design pressure. RCS pressurization is annunciated in the Control Room whenever the pressurizer pressure exceeds 392 psia and its associated SDCS isolation' valve is not in the fully closed position. Thus, the overall design of the Waterford 3 SDCS, in combination with administrative and procedural controls, could provide a high degree of confidence that disabling the autoclosure interlocks during Mode 5 operation will result in a net increase in plant safety. Further investigation is necessary prior to proceeding with a change to the ACI. 5.4 Vortexing Prior to plant licensing, and in response to INPO SER 83-60, Waterford 3 conducted testing to determine the RCS level at which vortexing (LPSI pump cavitation) occurred, and confirm the capability of the shutdown cooling system to operate without vortexing at the hot leg mid-plane (13' 4k" MSL).
l l In conducting the tests, operators were directed to lower RCS water level until cavitation was evident from either the LPSI pump ammeter or the Li'SI pump suction gauge. At a LPSI' flow rate of 4000 gpm it was determined that pump cavitation occurred at an RCS level of 13' 1 ".MSL. Level was increased to 13' 4 " MSL (hot leg mid-plane) and shutdown cooling flow maintained for 2 hours with no further evidence of cavitation. (Tests conducted during the same time period also confirmed acceptable operation of the vacuum priming system.) Because part-loop operation is conducted at a shutdown cooling flow rate of 3000 gpm, rather than the 4000 gpm applied during testing, the actual vortexing level during part-loop operation would be less than the 13' 1h" MSL value identified during testing. Although there appears to be adequate margin between the lowest part-loop operating level (hot leg mid-plane) and expected vortexing level, Waterford 3 is pursuing an analysis to determine the minimum acceptable shutdown cooling flow rate based on decay heat removal considerations. Should a significant flow rate reduction be technically justifiable, appropriate steps will be taken to reduce the required flow during part-loop operation and thereby further reduce the LPSI pump vortexing icvel. l l
6.0 PART-LOOP OPERATION DURING LOSS OF SHUTDOWN COOLING 6.1 Operating Procedures and Administrative Controls During loss of shutdown cooling (SDC) events, operator actions are governed by Off-Normal Operating Procedure OP-901-046, Loss of Shutdown Cooling. The Waterford 3 task force has devoted a great deal of attention to reviewing this procedure and the events it is intended to mitigate. It became apparent that OP-901-046 had become fragmented and difficult to follow as procedural changes were' incorporated over the years to address industry loss of SDC event corrective actions. To remedy this situation, OP-901-046 is being rewritten to citrify operator actions during the many different scenarios in which a loss of SDC event could occur. Additionally (as will be discussed later) certain changes will be implemented for part-loop loss of SDC events identified during the Generic Letter 87-12 review. The description of OP-901-046 provided below covers the unrevised procedure. Upon entry to the procedure the operator is directed to one of three sets of actions depending on the cause of the loss of cooling:
- a. System Leakage
- b. Loss of SDC Flow, or
- c. Loss of SDC Heat Removal Capability.
System leakage (e.g. tube rupture in the SDC heat exchanger) and loss of SDC heat removal capability (e.g. loss of component cooling water to the SDC heat exchanger) are not unique to part-loop operation and will not be discussed in any detail. For a loss of SDC flow during partially drained conditions operator efforts are directed at restoring flow and preventing loss of inventory. When loss of flow is due to air binding of the LPSI pump (s), operators are cautioned that venting of the LPSI pump casing may be necessary, and are directed to place the SDC Vacuum Priming System in service. At this point, operators should:
- a. Secure any draining evolutions in progress and,
- b. Restore RCS level by taking LPSI pump suction from the RWSP for the unaffected pump, closing the SDC suction header isolation valve and running the unaffected LPSI pump until level is restored.
Although the system leakage portion of OP-901-046 directs use of the standby HPSI pumps and provides adequate cautions to establish containment integrity prior to exceeding an RCS temperature of 200*F, i the loss of SDC flow instructions are silent in this regard. To l correct these oversights and implement other improvements identified during the Generic Letter 87-12 review, the following instructions applicable to part-loop operation will be included in the revision to OP-901-046: w_____-___________-__________
'
- hl a. A caution will be added to establish containment integrity (see Section 6.3) prior to exceeding an RCS temperature of 200*F as indicated by the CETs or by estimation using a table of RCS heat up rates vs. time after shutdown (see Section 6.2.2).
- b. If flow cannot be restored, immediate actions are to be initiated to establish containment 1
integrity.
- c. RCS level is to be restored and initially maintained using the standby HPSI pump and, when major disassembly or removal of an RCP has I occurred, simultaneous hot and cold leg injection,
- d. To ensure a heat sink, establish feed and bleed on one or both steam generators using a condensate pump, EFW pump or AFW pump, in conjunction with steam generator blowdown, if available. If secondary side water has started to boil, I, establish steaming through the ADVs (or turbine bypass valves if the condenser is available).
- e. In the event of cavitation or air binding of a LPSI pump, a caution will be added to delay starting the second LPSI pump until it is properly vented and pump suction is restored and verified.
- f. Attempt to restore flow on the affected LPSI pump by slowly increasing flow to the desired flow rate.
6.2 Instrumentation Instrumentation design and operation during normal part-loop operation is reviewed in Section 5.2. The following discussions cover aspects of the instrumentation unique to part-loop loss of shutdown cooling conditions. 6.2.1 Level The HJTC System level measurement will be largely unaffected by the RCS thermal-hydraulic conditions existing during a part-loop loss of shutdown cooling event. Therefore, the HJTC System is available as a reliable cross check for other RCS level indications when the reactor vessel head is in place. In the Refueling Level Indication System, the ortac rubber hosing is rated for a working pressure of 300 psig at 180*F. The Tygon tubing sight glass employed at Waterford 3 has a burst pressure of 144 psig at room temperature and a burst pressure of approximately 30-40 psig at 212" F, based on a conversation with the manufacturer. The tubing is rated for a working pressure of 15 psig at 180*F. Testing during installation included exposure to a negative pressure of approximately 5 psig without tubing collapse, and pressurization to approximately 30 psig at ambient temperature with no tubing damage.
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- The Tygon tubing sight glass temperature will not appreciably exceed I ambient containment atmosphere temperature (assumed as 120*F maximum in accordance with Technical Specification 3.6.1.5 for Modes 1-4). As discussed in Section 7.1, the peak pressure experienced by the Tygon tube will not exceed 35 psig for the closed RCS part-loop loss of shutdown cooling event. Therefore, Tygon tube integrity should not normally be a concern for the issues raised in Generic Letter 87-12.
'l As noted in Section 5.2, the RLIS/RWLIS nozzle is located on the hot !
leg close to the shutdown cooling suction nozzle. As a result, air ingestion through the shutdown cooling system should have a negligible impact on level indication, unlike the cold leg level indicator reviewed in NUREG 1269. 6.2.2 Temperature RCS temperature indication during a loss of shutdown cooling event is important as an indication of approach to boiling and the need for containment isolation, as well as meeting administrative requirements for the Mode change from Mode 5 to Mode 4, which occurs at 200*F. As noted in Section 5.2.2, at least 2 CETs will be available when the reactor vessel head is in place, except when preparing for head removal or replacement. As a backup to the CETs, and to cover situations when the reactor vessel head is not in place, Waterford 3 has prepared a table of RCS heat-up rates vs. time after shutdown. As noted-in Section 6.1, this table will be included in Off-Normal Operating Procedure OP-901-046, Loss of Shutdown Cooling, with instructions to the operators to utilize the table in determining the need for establishing containment isolation. (It should be noted that when the reactor vessel head is not in place (or when preparing for head removal or replacement) the RCS water level is maintained several feet above the hot leg centerline. This precludes the possibility of losing SDC flow due to vortexing). 6.2.3 Pressure RCS pressure indication is reliable during a part-loop loss of shutdown cooling event provided the actual pressure levels or changes are of a sufficient magnitude to match the resolution capability of the indicators (0-3000 psia). The presence of a hole in the cold leg due to major RCP maintenance results in a change in pressure (less than 3 psig in the hot leg - see-Section 7.1) too small to be discriminated on the wide pressure ranges of available pressurizer pressure instrumentation. 6.2.4 LPSI Pump Parameters The LPSI pump indicators and alarms covered in Section 5.2.4 will be unaffected by a part-loop loss of shutdown cooling event, although indications will likely be representative of pump cavitation if the pump is running. _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ i
i l , . l' l 6.3- Containment Isolation Waterford 3 maintains containment integrity during Modes 1-4 in accordance with Technical Specification 3.6.1.1 in order to ensure that the release of radioactive material from the containment atmosphere is restricted to the leakage paths and leak rates assumed in the safety analyses. Part-loop operation is conducted in Mode 5 (Cold Shutdown - K less than0.99,averagecoolanttemperaturelessthanorequalto$bb*F)or lower. To ensure that the safety analysis assumptions are preserved, a means should be available to accurately determine or predict an entry into Mode 4 (i.e. average coolant temperature greater than 200*F). Presently, Off-Normal Operating Procedure OP-901-046, Loss of Shutdown Cooling, cautions operators in the system leakage section that containment integrity must be established prior to exceeding an RCS temperature of 200*F. During the revision to OP-901-046 additional instructions will be provided to clarify the situations under which containment integrity should be initiated (particularly during part-loop operation) and provide RCS temperature criteria for making that determination (see Sections 6.1 and 6.2.2). Waterford 3 is presently taking other steps to facilitate timely containment isolation should it be necessary during part-loop operation. The major impediment to rapid containment closure is closing the containment equipment hatch and removing the hatch platform used for movement of equipment into the containment during outages, which currently requires approximately 4-8 hours. The necessary training and equipment will be in place by the second refueling outage to allow equipment hatch closure within approximately two hours under partially drained RCS conditions. 6.4 Pumps Available to Control RCS Inventory The pumps available to centrol RCS inventory during part-loop operation at Waterford 3 include the HPSI, LPSI and charging pumps. Technical Specification 3.1.2.3, Charging Pumps-Shutdown, requires one HPS1 or one charging pump operable in Modes 5 and 6. One operable HPSI pump is required administratively by OP-1-003, Reactor Coolant System Drain Down, and OP-10-001, General Plant Operations, if drain down of the RCS is planned. Both LPSI pumps are required to be operable by Technical Specification 3.4.1.5, Cold Shutdown-Loops Not Filled, during Mode 5 part-loop operation. The LPSI pump (s) may be used for RCS makeup by opening the suction valves to provide water from the RWSP. This method of makeup is addressed in OP-901-046, Loss of Shutdown Cooling (see Section 6.1). As noted in Section 6.1, OP-901-046 is being revised to clarify the preferred order of RCS makeup (HPSI, then LPSI) during a loss of shutdown cooling event.
Given the procedural requirements for HPSI pump availability, charging pump (s) may or may not be available in accordance with Technical Specification 3.1.2.3. In most situations, however, at least one of the three charging pumps should be available to provide water to the RCS. Similarly, Boric Acid Makeup Pumps could be utilized for RCS makeup although Technical Specification 3.1.2 (Boration Systems Flow Paths - Shutdown) does not require the pumps to be operable. Finally, while not required by Technical Specifications, one Containment Spray pump could be made available if needed. 4 I d
7.0 GENERIC LETTER 87-12 POSTULATED LOSS OF SHUTDOWN COOLING SCENARIO 7.1 Loss Of Shutdown Cooling Scenario Progression At Waterford 3 Shutdown cooling flow can be lost with the RCS in a partially drained condition. When this occurs the water in the core begins to heat up since decay heat is not removed. Local natural circulation loops are established in the reactor vessel with hot water rising up the center of the core and cooler water falling down the periphery of the core. Eventually the water temperature will increase to saturation, and boiling will begin. The time to achieve bulk boiling can be calculated from the core decay heat and the heated water mass. As the coolant starts to boil, steam will be condensed on the structural surfaces above the core (upper guide structure, hot leg pipe, etc.). These surfaces will heat up and eventually fail to condense steam. The steam not condensed will push air into the steam l generators, degrading the heat transfer coefficient and causing pressure to increase. If the RCS is closed (no inventory loss paths), pressure will increase enough (compressing air in the steam generator and increasing the heat transfer rate) so that steam is condensed by cool secondary side l water. Based on conservative assumptions, RCS pressure should not exceed 35 psig. The condensate will fall back to the core and , maintain core cooling by reflux boiling. The RCS will stay in this i condition until the steam generator secondary boils dry, cool water is injected by the HPSI pump, or shutdown cooling flow is restored. It would take more than 6 hours to completely boil away the secondary side inventory (initially at 70% wide range level) with no credit for makeup. Replacement of an RCP seal does not result in any appreciable inventory loss from the RCS. This is because the rotating assembly baffle on the RCP shaft is in contact with the surface of the RCP seal cooling heat exchanger. This metal-to-metal seal prevents any significant RCS leakage until RCS pressure increases enough to lift the RCP shaft rotating assembly (approximately 60 psig for Waterford 3). Therefore, since the maximum RCS pressure for a loss of shutdown cooling event (with a steam generator available) is less than the pressure required to lift the rotating assembly, the RCS remains closed during the replacement of an RCP seal. If the RCS is open on the cold leg side due to a major disassembly or removal of an RCP, the pressure increase in the upper plenum (hot side) will depress the water level in the core and in the steam generator outlet pipe. The level will drop to the elevation of the top of the loop seal (in the RCP suction leg), 20 inches above the top of the active core. The maximum upper plenum pressure is equivalent to the hydrostatic head of liquid from the top of the cold leg to the top of the loop seal, or about 3 psig. (A higher pressure, and therefore lower level, would allow air or steam to be vented through the loop seal and out the RCP opening. This would cause the pressure to decrease and the core level therefore to increase. It should also be noted that vent paths are available but not credited to equalize pressure between the hot and cold sides of the RCS - e.g. leakage j paths around the hot leg nozzles and other in-vessel leakage paths.) l
l i The core will remain covered and cooled even with the level depression. Water from the hot leg and upper plenum will be pushed into the cold leg, overfilling the cold leg. Some inventory will be ! lost ao the excess water spills out through the_RCP opening. As steam enters the steam generator, it will be condensed by cool water on the secondary side. The condensate will fall back to the core. The core water level will remain at 20 inches above the top _of the active core since no additional inventory loss will occur _(i.e., all the steam is condensed due to the high primary to secondary temperature difference). As the secondary water heats up, the steam will penetrate farther into the tubes displacing air which is vented through the loop seal and out the RCP. Eventually the secondary side water temperature will almost equal the primary side steam temperature. At this point it is possible that some steam would not be condensed. This would depend on the heat transfer coefficient and actual temperature difference. Steam exiting the steam generator tubes would be vented through the loop seal and condensed by the water on the vertical suction leg between the RCP and loop seal. When this water reaches saturation, inventory loss would begin as the amount of steam not condensed passed out the open RCP. Water level in the cold leg will decrease as this mass is lost. Eventually the core will uncover and heat up unless water is injected to replenish the lost inventory. 7.2 Time to Core Uncovery The time at which the core becomes uncovered (with no safety injection and an open RCS) depends on the decay heat rate (time after reactor shutdown) and the rate of inventory loss. Hand calculations have been performed (and confirmed through work authorized by the CE Owners Group) to determine the approximate time at which the collapsed water level drops to the top of the active core for the case of an open RCS described above. Water in the RCS is initially assumed to be at 120*F. Heat up of the water to saturation temperature and subsequent steam condensation by the upper guide structure and hot leg metal is calculated in a manner similar to that used in Appendix E of NUREG 1269. The steam produced is then condensed in the steam generator tubes by cool water on the secondary side. Only one steam generator is assumed available. When the secondary side water temperature reaches saturation, it is assumed that a portion of the steam is not condensed. This steam escapes from the RCS through the open RCP resulting in a loss of inventory and decreasing water level. The core uncovery time is strongly dependent on the amount of steam assumed to not be condensed after the secondary side water reaches saturation. This in turn depends on the primary to secondary temperature difference and heat transfer coefficient in the steam generator. If it is assumed that 50% of the steam produced in the core is condensed in the steam generator, then the collapsed water level will fall below the top of the active core in 2.2 hours at one day after reactor shutdown. If 0% of the steam is assumed to be condensed, the most limiting case, the collapsed level will fall below
the top of the active core in 1.5 hours. Even with the standard safety analysis assumption of operator inaction for 30 minutes, this is more than adequate time for operators to replenish the RCS inventory with the standby high pressure safety injection pump and ensure that the core remains cooled. It should be noted that the fuel will not begin to heat up until the two-phase mixture level, which is higher than the collapsed water level, falls below the top of the core. Thus, fuel heat up and core damage would be delayed significantly beyond the times given above. In addition, as discussed in Section 4.3, draining to the hot leg mid-plane under realistic conditions would not be complete for 2-2 days following reactor shutdown compared to the one day assumed above. The lower decay heat values under the realistic scenario result in additional time for operator action prior to the collapsed water level falling below the top of the core. 7.3 Significance of Postulated Scenario for Waterford 3 As noted above, the loss of shutdown cooling scenario during part-loop operation with an open RCP could conservatively be expected to uncover the core in 2.2 hours assuming no operator action.- This extended period of time is largely due to maintaining the steam generator (s) as a heat sink by requiring at least one steam generator be operable with a minimum wide range water level of 70% for part-loop operation (see Section 4.1.1). Practically, there is more than sufficient time to credit operator action to restore RCS inventory and level. The Waterford 3 Loss of Shutdown Cooling procedure (OP-901-046) will direct that the standby HPSI pump, required to be operable by the Reactor Coolant System Drain Down procedure (OP-1-003), be utilized to makeup RCS inventory. In the unlikely event that the HPSI pump should fail, the LPSI pump can be aligned to take suction from the RWSP and provide RCS makeup. Consequently, the core damage scenario postulated by Generic Letter 87-12 is within the Waterford 3 procedural and design basis capability to mitigate without fuel damage or the release of radioactive j material. 7.4 Steam Generator Availability During Refueling Outages During refueling outages, one or both steam generators may be unavailable for limited periods of time due to steam generator inspections, sludge lancing, etc. In these situations the steam generator plenum manway(s) are generally open to the containment atmosphere, providing a direct path for steam escape should RCS boiling occur, with neither steam generator available as a heat sink. For this case, steam would be lost from the RCS without condensation in the steam generators. With no safety injection (and at one day after shutdown) the collapsed water level would drop below the top of the active core approximately 1.5 hours after the loss of shutdown cooling flow. Safety injection flow from the HPSI or LPSI pump within this time would be sufficient to exceed the boil off rate and prevent core uncovery.
To minimize time in this configuration Waterford 3 intends to perform work on both steam generators (if necessary) simultaneously. Experience has shown that steam generator nozzle dam installation can equal or exceed the time necessary to perform steam generator maintenance. At the same time, nozzle dam installation requires draining the RCS to approximately 5 inches. lower than the level needed to perform the maintenance. Therefore, the benefits of nozzle dams. are offset by the increased time of steam generator unavailability and the closer approach to the vortexing level. As a result, Waterford 3 will evaluate the use of nozzle dams prior to an outage. involving steam generator maintenance, taking into account the unique aspects of the outage. Finally, steam generator maintenance activities will be scheduled to minimuze steam generator unavailability consistent with good maintenance practices. 1 i l 1 1
8.0 TRAINING l Operations personnel are provided training in areas affecting shutdown cooling and partially drained reactor coolant system operation in a number of programs and environments. Senior Reactor Operators receive formal classroom training involving l off-normal procedures and significant industry events. There are separate lesson plans for each of these areas which include discussion on actions to be taken should a loss of shutdown cooling event occur. The lesson plans include coverage of the Waterford 3 Loss of Shutdown Cooling Event on July 14, 1986 (LER 86-015). In their training, Reactor Operators are required to perfrcm or simulate, and discuss a normal operating procedure, an off-normal o[ rating procedure and a surveillance test, each related to shutdown cooling system operation. All Licensed Operators are required to review the Waterford 3 Loss of Shutdown Cooling LER 86-015 as part of requalification training, as well as l the procedure changes that occurred as a result of the event. Prior to the first refueling outage, all Licensed Operators were required ! to review selected procedures during requalification training that were not applicable to operating at power, which included procedures OP-1-003, Reactor Coolant System Drain Down and OP-901-046, Loss of Shutdown Cooling. Selected procedures related to shutdown cooling will again be required for l review by Licensed Operators prior to the second refueling outage. Auxiliary Operators receive formal classroom training regarding the shutdown cooling system which is supplemented with discussion of off-normal procedures (including loss of shutdown cooling) with Operations supervision personnel. The Waterford 3 simulator has recently been placed in service and Training personnel are becoming familiar with its capabilities. Licensed Operators will be given the opportunity to experience loss of shutdown cooling l scenarios to the maximum extent possible within the design constraints of l the simulator (e.g. reflux boiling is not presently modeled). Should such scenarios be achievable on the simulator, this training action will be l factored into future requalification training programs for Licensed Operators. At this time, there is no formal classroom training conducted for Maintenance personnel in areas affecting operation with a partially drained Reactor Coolant System. However, "On the Job Training" checklists for maintenance technicians require the use of plant procedures, including i those related to mid-loop operation, to accomplish surveillance testing. l These procedures contain specific precautions, which are intended to i prevent undesirable events like loss of shutdown cooling, to be observed l during the conduct of these evolutions. The need for more formal training of Maintenance personnel on the potential effects of maintenance activity on shutdown cooling operability and containment isolability is being evaluated. Should a need for more formal training in this area be identified as a result of our evaluation, it will be accomplished before the start of the second refueling. l Finally, procedure and policy changes instituted as a result of the Waterford 3 review of Generic Letter 87-12 will be incorporated into the i appropriate training programs following implementation of the changes and, in all. cases, prior to the second refueling outage.
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