ML18092B564
| ML18092B564 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 04/02/1987 |
| From: | ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY |
| To: | |
| Shared Package | |
| ML18092B562 | List: |
| References | |
| S-87-05, S-87-5, NUDOCS 8705110301 | |
| Download: ML18092B564 (58) | |
Text
{{#Wiki_filter:"LICENSING REPORT S-87-05" PATED APRIL 2, 1987 LICENSING REPORT FOR NEW NARROW RANGE TEMPERATURE MEASUREMENT SYSTEM (RTD BYPASS ELIMINATION) PUBLIC SERVICE ELECTRIC & GAS SALEM 1 & 2 NON PROPRIETARY ~ -----~-
- 8705110301 870505----------,
PDR ADOCK 05000272 P PDR
I
.1
1.0 INTRODUCTION
INDEX
2.0 BACKGROUND
3.0 OVERVIEW OF THE PROPOSED SYSTEM 3.1 Mechanical Changes 3.2 RTD Design 3.3 Electronic Modification 3.4 System Accuracy 3.5 ALARA Benefits 3.6 RTD System Time Response
4.0 DESCRIPTION
S OF MECHANICAL MODIFICATIONS 4.1 Hot Leg 4.2 Cold Leg 4.3 Crossover Leg 4.4 RVLIS Connection 4.5 Inspection, Welding and Hydrostatic Test Requirements . 4.6 Analysis of RCS Penetrations 4.7 Debris Control During Modification
5.0 DESCRIPTION
OF ELECTRICAL/INSTRUMENTATION 5. 1 T-hot Averaging 5 .1.1 Existing System 5 .1. 2 Proposed System 5.2 T-Cold Monitoring 5.3 Weed RTD I
5.4 In-Situ Testing -18...:. 5.5 Equipment Qualification l
- 5.6 Detection of a Failed RTD 6.0 ALARA 6.1 Description 6.2 Dose Estimation for the Modifications 6.3 Dose Savings 6.4 ALARA Methods 6.5 Radioactive Waste 6.6 Radiological Problems and Dosimetry 7.0 FUNCTIONAL IMPACTS 7.1 System Accuracy 7.2 Response Time Impact 7.3 Relocation of RTD Instruments 7.4 Reactor Coolant System Flow 8.0 UNREVIEWED SAFETY QUESTION. (USO) DETERMINATION
Attachment:
Appendix A Proposed .UFSAR page changes Dm3/2md II
- 1. INTRODUCTION:
The existing RTD bypass piping is scheduled to be removed during the upcoming 7th Refueling Outage on Unit 1 and the 4th Refueling Outage on Unit 2. The new narrow range inline thermowell mounted RCS temperature measurement systems will be installed during the same outages in lieu of the bypass systems. Combustion Engineering (CE) has been selected by PSE&G to-perform the detailed engineering and installation of the new system. This report is submitted in support of continued operation of the Salem Units with the new RTD System installed. 2 *. BACKGROUND: The current RTD Bypass Piping (B/P) System was designed to address temperature streaming in the hot leg and to allow replacement of direct immersion RTDs without drain down of the Reactor Coolant System.(RCS) The resulting system consists of nearly 280 feet of Reactor Coolant Pressure Boundary (RCPB) piping, additional lengths of vent, drain, and packing leak off lines, 68 associated val~es, 8 sets of flanges and 8 RTD manifolds. Plant experience has demonstrated two major drawbacks to this design: 0 Lack of Reliability - Plant shutdowns have been required because of leakage (from valve packing or mechanical joints) or because of flow reductions due to valve problems. 0 High Radiation Dose The RTD B/P System is a significant contributor to man-rem exposure because the numerous valves and socket welded_ pipes ~re drud traps~ Man-rem is expended not only in maintaining the RTD B/P System but in performing any work near the RTQ B/P System such as Bteam Generator and Reactor Coolant Pump maintenance. These problems are not unique to Salem but appear to be common to all plants with a RTD B/P System. The proposed narrow range, RTD System eliminates all the bypass piping, and its associated problems, while maintaining a fast response time, accurate hot leg temperature determination, and the capability to replace RTD's without draining down of the RCS. An overview of the new* system is provided in Section 3.0. Detailed descriptions are provided in Sections 4.b, 5.0, 6.0 and 7.0. Section 8 provides a saf~ty evaluation demonstrating that the proposed modification ~oes not represent an unreviewed safety question (USQ). r
- 3. OVERVIEW OF THE PROPOSED SYSTEM:
3.1 Mechanical Changes: All the bypass piping, associated valves, and RTD manifolds will be removed. The three mixing scoops in each hot leq are-retained. [ _] The top of the hot leg mixing scoops will be modified to allow welding on a thermowell. The thermowell becomes part of _the RCPB. The RTD nipple and head assembly screws into the thermowell. The nozzle on the cold leg will be modified to ailow welding on a thermowell. The cold leg configuration is simpler because the Steam Generator and RCP provide adequate mixing of the fluid in that piping. . The cross over leg connection, through which the RTD B/P System fluid is returned to the main RCS piping, will no longer be required and will be capped. The RVLIS connection will be made directly into the hot leg at the same elevation where it used to join the bypass piping. 3.2 RTD Design: Weed, dual element RTDs will be used in the new design. Each RTD element will be tested inside a thermowell to ensure that the time response of both
.elements is within the r~quired time. Response time of the RTDs will be verified in the field using loop current step response methodology. RTD accuracy will be superior to the accuracy of the
*present RDF RTDs. The spare RTD element will be wired all the way to the RPS cabinet so that switchover to the spare element can be done from the Control Equipment Room, which is located adjacent to the Control Room.
r
- 3.3 Electronic Modifications:
Each of the three T-hot RTDs per loop will be wired up to a low voltage amplifier (MV/I) and the three signals then averaged to produce one T-hot signal which will replace the loop's T-hot signal of the existing system. The added electronics will be identical to the existing 7100 electronic hardware now used. Figure 5.1 shows. the concept and outlines the added modules required. 3.4 System Accuracy: By rete~tion of ihe mixing concept at the hot leg scoops, the sampling performed by the existing system will be preserved. The new system is however more accurate primarily because of higher accuracy RTDs and because the reduction in error which results from averaging the 3 RTD signals is greater than the error component introduced by the T-hot averager . 3.5 ALARA Benefits: A 3000 man-rem dose savings is projected over the remaining life of the two plants as a result of this modification assuming a 40-year Operating License. The estimated dose of approximately 150 man-rem .associated with the demolition and installation work of the RTD proposed modifications is accounted for in the overall dose savings. *This ALARA and cost benefit analysis takes into account the reduced radiation levels, reduced outage t1me, increased accessibility in the loop. compartments, installation/demolition doses, maintenance requirements and the plant's reliability over the life of the plant. *
- . . *. r
- 3.6 RTD System Time Response:
The modif icatipn will affect the following Technical Specification (T.S.) instrumentation response time: T.S. 3.3.l; Table 3.3-2, Item 7: Overtemperature-AT. T.S. 3.3.2; Table 3.3-5, Item 5: Steam flow high with low low T-AVG. Although the new RTD/thermowell response time is slower than the response time of the existing direct immersion RTD's, the loop travel and thermal lag time has been decreased by an equal amount. Total response time will be maintained and the results of the UFSAR design bases analyses are unchanged *
4.0 DESCRIPTION
S OF MECHANICAL MODIFICATIONS 4.1 Hot Leg: The hot leg installation has three nozzles 120° apart around its circumference._ The nozzles extend into the pipe to form scoops to sample the flow. The scoops will be retained in the new design and will collect a flow samp_le in a manner equivalent to the present configuration. A thermowell will be mounted inside each scoop~ The scoops will be modified so that the flow goes past the thermowell (Figure 4..;.1). Since the existing sample scoops are being retained, the method of sampiing the stratified flow in the hot legs_will remain unchanged. L
- ]
- -s-
The portion of the overall response time attributable to the flow through the. scoop is 0.25 seconds or less. This value includes fluid transit and heat capacity effects and was conservatively estimated using: 4*2 Cold Leg: The cold leg has a ~ingle nozzle without a flow sampling scoop. The nozzle is a 2" IPS 6000# socket weld half coupling. The cold leg thermowell will be installed directly into the coupling (Figure 4-2) ** As is the case with the present system, no flow s~mpling ~ill be neces~ary because the Reactor Coolant Pumps will provide mixing of the flow after it exits the steam generator . 4.3 Crossover Leg: The return for the bypass loops is a 3" nozzle in the crossover leg. This connection will no longer be used. A 3" schedule 160 buttweld cap will be installed on this connection to close it off. (Figure 4-3) 4.4 RVLIS Connection: Currently the Reactor Vessel Level Instrumentation System (RVLIS) taps off the bypass loop on two of the four hot legs. Since the bypass loops will be eliminated, the RVLIS connection must be moved to the hot leg pipe. A new penetration will be made in the hot leg at the same elevation as the previous connection to the bypass loop. The hole size of the new penetration will be the same as exists at the current connection. This will provide the same flow restriction in the new system as exists in th~ current system. (Figure 4-4) 4.5 Inspection Welding and Hydrostatic Test
- Requirements:
4.5.1 Hot Leg & Cold Leg Thermowells and RVLIS Penetration. The following requirements are applicable to the 12 hot leg RTD Scoops, 4 cold leg connections and 2 new RVLIS penetrations.
- 1. Liquid penetrant inspect all accessible field machined surfaces in accordance with the 1983 edition of Section III, Class I or XI of the ASME B & PV Code.
- 2. Welding to be in accordance with the 1983 editions Section III, Class I or XI of the ASME B & PV Code. (Root Pass-GTAW, Fill-GTAW, or SMAW)
- 3. Liquid penetrant inspect weld pass in accordance with the 1983 edition of Section III, Class I, or xi of the ASME B & PV Code *
- 4. Liquid penetrant inspect final weld pa_ss in accordarice with the 1983 edition of Section III, Class I, or XI of the ASME B & PV Code.
- 5. Weld material to be supplied in accordance with ASME Section II with additional requirement of ASME Section III NB-2400 (1983 edition).
4.5.2 Crossover Pi~irig The following inspection and welding requirements are applicable to capping of the 3" Crossover piping at four locations *
.1. Liquid penetrant inspect all field machined surfaces in accordance with the 1983 edition of Section III, Class I or XI of the ASME B & PV Code.
- 2. Welding to be in accordance with the 1983 edition of Section III, Class I or XI of the ASME B & PV Code. (Root-GTAW, Fill-GTAW, or SMAW)
- 3. Liquid penetrant inspect the root weld pass in accordance with the 1983 edition of Section III, Class I or XI Qf the ASME B & PV Code.
- 4. Liquid penetrant inspect final weld pass in accordance with the 1983 edition of Section III, Class I or XI of the ASME Code.
- 5.
- Radiographically inspect the completed weld in accordance .with the 1983 edition of Section III, Class I or XI of the ASME B & PV Code.
- 6. An open butt weld configuration will be used with helium as purge gas with approved water soluble* dams for purge gas sealing. *
- 7. Weld.material to be supplied in accordance with ASME Section II &
additional requirement of ASME Section III NB-2400 (1983 edition).
4.5.3 Hydrostatic Test Requirements Hydrostatic testing of all nozzles will be done during inservice in accordance with the 1983 edition of the ASME Section XI IWB 5000. 4.6 Analysis of RCS Penetrations: The thermowells are pressure boundary p*arts which completely enclose the RTD. They will be machined from a solid bar of SB-166, a nickel-chromium-iron alloy and will be shop hydrotested to 1.25 times the RCS design pressure. The external deiign
- pressure and design temperature will be the reactor coolant system design pressure and temperature.
The RTD therefore will not be part of the pressure boundary. For both the hot leg and cold leg, the nozzle, thermowell, and the entire thermowell/nozzle assembly will each be analyzed to the ASME B&PV Code, Section III, .Class 1. The analysis of the entire assembly will consider the weight of the. RTD, the RTD head assembly and an assumed length of cabling. The effect of seismic and flow induced loads will also be considered. Seismic response spectra specific to Salem will be used. Flow induced vibration will also be evaluated. The crossover leg connection will be analyzed to the same requirements as the hot and cold leg connections. Since the connection will be capped and have no piping loads, stress levels will be* lower than what exists in t~e currerit system. The change in the RVLIS piping and the new connection to the hot leg*will be analyzed i~ accordance with ASME Section _III Cla~s 1. 4.7 Debris Control During Modific~tion: Control *of metal chips and fragments will be as follows: Hot Leg RTD Scoop Modifications: (12 locations) Cold Leg RTD Connection: (4 locations) A freeze plug will be installed prior to any cut off operation. A mechanical plug will be installed after the freeze plug is removed and pr i.or to any machining which would develop metal chip or fragments. Arty chip or fragments will be removed by vacuum and other approved method prior to removal of the mechanical plug. The holes in the scoops will be made using the EDM process. Hot Leg RVLIS Penetrations: (2 locations) The penetrations for the RVLIS system will be drilled to within 5/8 inch of the hot leg pipe I.D. This area will be cleaned prior to the next operation which is the removal of the remaining 5/8
- inch using the EDM process.
Electrical Discharge Machining (EDM) is a process that utilizes electrical discharges, .or sparks, to machine metal. The surface being machined is bombarded with high intensity electrical energy pulses that gradually melt away the stock until the desired configuration is obtained. A high energy spark, through vaporization, melting, and an explosive effect, dislodges a minute particle of metal from the workpiece, leaving a small crater. The dislodged particle, is then solidified arid washed away by th~ dielectric fluid. Crossover Piping: (4 16cations) Th~ primary system will be drained prior .to any cutting operation on .the crossover piping. The cutting operation will be performed using an abrasive cut-off wheel which does not develop metal chips. A mechanical plug will be installed prior t6 machining of the weld prep and the nozzle vacuum cleaned prior to removal. A water soluble dam will be installed in the cros~over nozzle for the helium purge while welding *
- HOT LEG THERMOWELL
-II -
FIGURE 4-1 .
/
.- SOCKET WELD TO EXISTING HALF COUPLii~G 1!/l/1;
. COLD LEG THERMOWELL FI GUR.E 4-2
- - \ 2.-
,.r-3" SCH 160 BUTT WELD CAP
.- ORIGINAL NOZZLE
\
* \\ ' :
l l : \ \ \
\
\ \ \
CROSSOVER LEG CONNECTION FIGURE Lf-3
/*- 3/W' SOCKET \~ELD
/ COUPLING
-- 0 375 HOLE I II CNE~i PENETRAT 101~)
RVLIS CONNECTION FlGURE 4-4
- \ 4-
5.0 DESCRIPTION
OF ELECTRICAL/INSTRUMENTATION 5.1 T-Hot Averaging: 5.1.l Existing System. The fluid from .the three scoops for each loop is mixed together before being directed to th~ T-hot RTD manifold. At the manifold, a single RTD is used to measure the temperature. The voltage drop across this RTD is amplified by a low voltage amplifier (MV/I) before being combined with the T-cold signal to generate the loop's ~T and T-AVG signals used by the Reactor Protection System (RPS).Refer to Figure 5-1. 5 .1. 2 Proposed System. The proposed system will locate a dual element RTD in each of the three scoops. Averaging of RTDs at the three locations will be done electronically. Refer to Figure 5-1. The voltage drop across one element at each lo*cation will be amplified by its respective MV/I. The amplified signal from the three MV/Is will be averaged* together to generate a single T-hot signal for that loop, T-HAVE, wh1ch along with the T-cold signal is then used to generate the loop's 6 T and T-AVG signal. The second element at each location is considered an installed spare. It will be wired up to the RPS cabinet, but not normally connected to any electronics. On failure of the first element, the *second element is available. (Refer to Section
- 5. 6. )
5.2 T-Cold Monitoring: The impact on the T-cold portion of the system is limited to: (l} Relocation of the RTD from the manifold into a thermowell directly in the RCS cold leg piping. (2} Changing to one dual element RTD instead of two single element RTD's. As with the hot leg RTD's, both elements will be wired to the RPS cabinet but only one element will input into the electronics. As before, the RCS fluip is mixed by the Reactor Coolant Pump before reaching the cold leg. No sampling scoop is required. The proposed *1ocation of the cold leg RTD/thermowell is the nozzle used as a tap-off point for the existing cold leg RTD bypass line.
- 5.3 Weed RTD:
- The Weed RTD will meet IEEE-323-1974, IEEE-344-1975 and NUREG 0588/10CFR50.49 to.the following levels which envelope Salem Units *1 & 2 levels:
- DBE Containment Peak Temperature 49 5°F Environment:
Long Term Temperature 300°F Pressure 100 PSI Humidity: 100% Radiation: 30 O. M Rads-'. Chemical Spray: 11,000 ppm ,Boron and pH :4-11.0 Seismic: 20 G acceleration from l-33Hz *
- Based on an RTD head temperature of 160°F, the RTD will have a qualified life of 13 years.
The temperature at the. RTD head has been conservatively determined to be less than 160°F. The limiting com~onent is the epoxy used at t~e top of the RTD. If a lower head temperature is demonstrated, the qualified life can be increased. The RTD has a 40 year qualified life with head temperature of 120°F.
- The 4-wire dual RTD will be qualified by ~imilarity to the tested 3-wire RTD. The design and the technique of construction will be similar.
Dual element RTDs have been supplied and are in use at other operating plants including Waterford Steam Electric Station Unit-3. The RTDs are provided with Resistance vs. Temperature (R vs T) calibration curves which are accurate to a specification of + 0.05°F at 554°F. The RTD drift is specified to be within +1°F over a five (5) year period. The Weed RTD has the fastest response time of any available thermowell mounted RTD. It is the best available qualified RTD'for this application in terms of accuracy and drift. The RTD calibration is performed by immersion in ice and oil baths whose temperature is monitored by a standard RTD calibrated to NBS standards. The RTD/thermowell response time is me~sured by plunge method by causing a step change from ambient room temperature to elevated temperature. All RTDs must meet a specified response time requirement and will, therefore, be interchangeable. The dual ~lement design provide~ an installed spare wired up to the Hagan instrument racks for use when the primary element failure is detected. For this reason both elements of each RTD will be tested by Loop Current Step Response (LCSR) for in-situ response time after installation.
- The contact between the RTD and Thermowell is a critical item in maintainina the response time.
IT..
.J The desi~n of the taper and the spring load on the RTD ensures that the metal to metal conta~t is maintained. No soft metal or ncontaot fluidh is used. In addition the contact is not at the very bottom of the tharmowell which would be most susceptible to dirt intrusion.
Th~ Salem procedures will include a requirement for visual examination of the RTD ~nd thermowell taper areas for cleaninese just prior to RTD insertion. The above provides high assurance of consistant rosponse time. 5.4 In-situ Testing: The Weed RTD is capable of being tegted by the in-situ LCSR method. A continuous current of 20mA will not*damage the RTD.
- 5.5 quipmc~t Weed R'l.'D:
ou~11r1~~llun: The Weed RTD will be qualified to IE~E-323~1974, IEfE-344-1975 and NUREG 0588/10CFRSD.49 to levels which envelope Salem units l & 2 levels. Containment Penetration Feedthrough A~~emblies: The Conax feedthrough assemblies each contain twenty (20) #16 AWG copper conductors. These are desfgned for installation at spare penetration feedthrough locations. The feedthrough assemblies are qualified to meet IEEE-323-1974 and . IE!E-344-1975. The pigtail is additionally . qualified to IEEE-383-1974, seetion 2.5, vertical Flame Test. The qualification levels equal or exceed the requirements of Salem Units 1 & 2. RTD Quick Disconnect Assemblies: The Quick Disconnects will be supplied by Conax and will be installed at th& RTD head using qualified thread sealant. The male connector half will be I II connected to and be made an integral part of the *. I' RTD assembly. The female conn~ctor half will be provided with a 15 foot pigtail with il6 AWG wires for splicing to field cable. The assembly -also provides an environmental seal for protection of the RTD terminals in the RTD head from harsh environment during accident conditions. The assembly is qualified to IEEE-323-1974, and IEEE-344-1975 and the pigtails are additionally qualified to IEEE-383-1974, Section 2.5 Vertical Plame Test, The qualif ieation levels meet or exceed Salam units l & 2 levels. Flexible Conduit: Flexible conduit will be installed between the RTD Ouick Disconnect Assembly and the splicing Junctiori Box. The 3/4" s.s. flexible conduit is qualified to IEEE-344-1975. The qualification levels meet or exceed Salem Uni ts *1 & 2 levels. Tile flexible
- conduit will be provided primarily for mechartical protection of pigtail wires and no credit will be t~ken for environmental protection provided b~ its installation.
- The flexible conduit and the ~nd fittings will be supplied by service Air Company.
Field Cables: A new cable per loop (for new T~hot RTDs ) will be provided between the Jtinction Boxes and the Process Instrument Racks. The cable has $ixteen (16) twisted pairs with shield for each pair. The conductors are #16 AWG tinned copper insulated with FR-EPDM and Hypalon jacket. Each pair twisted with a*18 AWG drain wire in contact with an aluminm/mylar individual pair shieid. The cable is qualified to IEEE~323-1974, IEEE-344-1975 and ~EEE-383-197(, Section 2.s, Vertical Flame Test. The qualification levels meet or .exceed Salem Units l & 2 levels
- Splicing Ju.m.:tion Boxes:
one Junction Box shall be provided per loop for the threo hot leg RTDs. The existing Junction Box will be utilized £or the cold leg RTDs. The pigtails from the RTD Quick Disconnect Ass~mblies will be spliced to the field cables. The Junction Box will be provided with knockouts and fitting~ for the floxible conduit. The Junction Boxes will also be provided with cable entry fittings sized to accept the new field cables. The new Junction Boxes will be NEMA 4X, stainless steel enclosures with gasketed* cover. The new Junction Boxes will be seismically qualified (structural integrity) by analysis and similarity to previously tested boxes. The test level will envelope the response spectra at the instalied location. The. Junction Boxes provide an enclosure to contain the splices. The Junction*sox is not credited to p~ovide an environmental seal duri~g accident conditions..
- Splicing:
Burndy Series YSV butt splices will be used. Raychem WCSF Nuclear Grade shrink tubing will be applied over the splice. The splice shrink tubing is qualified to IEEE-323-1974, IEEE-344-1975 and IEEE-383-1974. The butt splices and heat shrink tubing shall also be utilized ror terminating ponetration feedthrough pigtails to the field cables. RPS Hardware: The added MV/I, I/~ and summators are identical to the existing 7100 electronic components. The electronic module used to derive th~ loop's average T-hot signal (*r-HAVE) from the indi11idual T-hot inputs is identical to the module now in ~se to derive. the loop's average temperature (T-AVG) from the T-hot and T-cold inputs.
- The added el~ctronics wi!l be installed in spare locations in the existing RPS cabinets. Divisional separation will be maintained. All additional mounting hardware will be identical to existing mounting hardware. All new electronics and mounting hardware is being procured through the same source used to procure replacements for the present equipment. The RPS cabinets will be shown to meet the seismic qualification. Cabinet wiring will meet Class lE requirements.
The above ensures that the added electronics will be compatible with existing electronics. It also minimizes the *impact on present training and procedures. In addition, all the equipment has been fully qualified and has a demonstrated high reliability, 5.6 Detection of a Failed RTD: A failed RTD would be picked up by the TAVG/ AT deviation alarm which is set at + 2°F. Also, each channel is checked every eight hours. on failure of a RTD, the channel would be tripped and Technical Specification Action Statement would go in effect starting then. Since all logics for controls and protection function require 2 out of 4 logic, the failed channel would have no impact on the safe operation or shutdown of th$ plant. As discussed in Sections s.1.2 (Hot Leg) and 5.2 (Cold Leg), the second element of each RTD is an "installed spare" which is, wired all the way to the RPS. This facilitates switching to the spare element as well as minimizing the time that one channel would have to be tripped.
-21~
RTD BYPASS ELIMINATION ~ALOG RTD AVERAGING
- SALEM 1& 2 COLD LEG TEMPERATURE HOT LEG TEMPERATURES ADDITIONAL ELECTRONICS REO'D. FOR I'!)
f'..!l I THE MODIFICATIONS FIGURE 5.1 LICENSING REPORT S-87-05 S-SUMMATOR VBB AT T AVG.
6.0 ALARA 6.1 *oescription: The project will involve the removal of all the RTD manifold bypass piping which consist of about 280 feet of highly contaminated two and three inch pipe and 68 valves per unit. In addition, about 160 feet of sligh~ly contaminated 3/4 inch drain piping will be removed. Following the removal of the piping, existing penetrations into the RCS piping will be modified to allow the installation of the thermowell and a new penetration installed for the RVLIS. The major steps involved in installation of the hot leg thermowells are:
- 1) Cut the pipe stubs remaining.
- 2) Prep the end of the nozzle for acceptance of the thermowell.
- 3) Using the EDM tool, -bore out the new fl6w holes.
- 4) Install therniowells and weld.
The major steps involved in installation of the cold leg thermowells are:
- 1) Cut the pipe stubs remaining.
- 2) Prep the end of the nozzle.
- 3) Bore out the cold leg cdupling t6 accomodate the new thermowell.
- 4) Install thermowells and welg *
- The major steps involved at the cross over leg are:
- 1) Cut the pipe stubs remaining.
- 2) End prep the nozzle.
- 3) Weld on the new caps.
The isolation valves and the RTD manifolds are the major sources of radiation in the existing system. It is expected that the removal of these components will be the most exposure intensive portion of the demolition phase. Remotely operated pipe cutters will.be used to the greatest extent possible. The radiation exposure rates at the RCS penetration work areas will be reduced by the removal of the isolation valves and RTD manifolds as well as the installation of large quantities of temporary shielding on the RCS piping.
- 6.2 Dose Estimation For The Modifications:
- The refined collective exposure estimates have not been prepared at this time but the preliminary estimates are:
Task -Man Rem Scaffolding/Temporary Power & -Light 4 Insulation Removal 2 Pipe/Valve Demolition 20 Pipe Disposal 1 RCS Penetration Modification 50 TOTAL/Unit 77 The work areas at the RCS penetration are expected to have radiation levels of 100 mR/hr to 200 mR/hr. The maximum exposure rate of 500 mR/hr to 1 R/hr will be seen in the vicinity of the present RTD manifolds and isolation valves which will be removed from the containment as soon as the RTD system is released by the plant Operations De par tmen t *
- 6.3 Dose Savings:
The arrangement of the Salem RTD manifold piping is such that the high radiation fields generated by them increases the collective exposure received during steam generator and reactor coolant pump inspections or maintenance. Although temporary shielding is used to reduce these radiation levels during long outages, it is not used for forced outages as it requires at least 1 1/2 days to erect~ Even with the use of temporary shielding, approximately half of the dose.received during steam generator and reactor coolant pump maintenance is attributed to the RTD oypass piping which is in the same area. The removal of the RTD manifolds is expected to reduce the collective expos~re by about 3,000 man ie~ over the r~maining life of the two units assuming a 40 year operating license. In addition, it is* estimated that at least one outage day per year will be .saved on each unit due to the avoidance of leaks and equipment failures. During the period of 1984 through 1986
- 6.4 the RTD system has required maintenance which resulted in collective exposures of about 95 man-rem.
ALARA Methods: The project is being reviewed and planned in accordance with PSE&G ALARA procedures. These plans include the producing of a vid~o which will show the arrangement of the work areas, potential interferences and access pathways. The ta.pe will be used to develop job plans and to familarize the workers with higher/lower radiation areas. Remotely operated equipment will be used to minimize stay times during the pipe demolition phase and portable. containment devices installed so that respirators will not be required. A mock-up of the small bore piping will be used to train workers in the procedures for glove box installation/removal and operation of the pipe cutting equipment. Extensive use of temporary shielding is planned f6r the modification of the RCS piping penetrations. The use of respirators will be minimized by local decontamination and by the use of a contaminment system_during the machining process *
- Similar ALARA work procedures and awareness has resulted in a relatively low man-rem exposure during 1985 and 1986 of 210 man-rem per year for I
each Salem unit. This is among the *lowest in the USA. A significant portion of the Health Physics (HP) personnel will be dedicated to this job and will receive specific training. To further ensure ALARA adherence, the installation procedures will be reviewed by HP personnel prior to finalization. Complementary HP proc~dures will be pre~ared. Although this p~rticular modification has not been previously installed on an operational plant the work to be performed is not unique. 6.5 Radioactive Waste: The waste generated by this project will consist largely of the removed piping and semi-encapsulated insulation. It is expected that a waste volume of less than 200 ft3 will result. The disposal method will be the usual low level waste burial process *
- 6.6 Radiological Problems and Dosimetry:
Although some components of the RTD manifold piping system do present rather high exposure rates, these can be managed by the use of the ALARA planning
,process. The glove boxes and/or other containment systems will contain almost all of the loose surface contamination and/or airborne radioactivity that might be produced. Although some concern has been expressed that some alpha-particle emitting nuclides could be present in the corrosion product films, these will also be controlled by the containment/ventilation devices. If possible, a sample of the corrosion products will be obtained and tested before starting work.
While the use of temporary shielding is expected to preclude the necessity of using multiple dosimetry,a plan is being formulated to have selected workers *wear dosimeters on their hands during the initial phases of the work. If the exposures are as expected, the use of extremity dosimeters will be discontinued
- 7.0 SYSTEM FUNCTIONAL IMPACT
- The narrow range RTD temperature outputs are used for a number of purposes including reactor trips, Engineered Safety Features (ESF) actuation, and control functions.
7.1 These are discussed in Section a.a. System Accuracy: The accuracy of the proposed system is an improvement over the existing system because of the following: Hot leg scoop mixing has been retained as discussed in Section 4.1. The replac~ment RTD has a significantly improved accuracy over the existing RTD's. The accuracy of the new RTD's is discussed in Section 5.3. Since the new RTD's will not be in contact with primary fluid and will be provided with a quick disconnect at the head, they can readily be removed. Little, if any, decontamination would be required to allow transport.to a, testing facility to
~ecalibrate the RTD's *
- Each hot leg RTD will be wired to a MV/I before averaging the three signals to obtain the loop's T-hot. By having three parallel path T-hot RTD's, MV/I's, and interconnecting wiring, the processing error associated with these components is reduced to 33% compared to the present single RTD, MV/I, and interconnecting wiring. *
- This reduction in processing error more than compensates for the addition of an averager to obtain a single T-hot for the loop.* The error introduced by the av~rager is minimized by maintaining the same electronic desi~n as now exists in the RPS. . .
There is no change to the cold leg's electronics; and therefore, no impact to the accuracy other than the increase obtained from a more accurate RTD.
Salem presently assumes a 3-1/2% error in primary flow determination. For Unit 2 this is controlled through Technical Specification 3.2.3. For Unit 1, it is controlled through PSE&G procedures. This allowance continues to be conservative. PSE&G intends, as a separate effort, to demonstrate that a significantly reduced allowance error is acceptable. 7.2 Response Time Impact: This modification will impact the following Technical Specification (T.S.) instrumentation response times: Table 3.3-2 item #7 - OvertemperatureAT: add l.75 seconds. Table 3.3-5 item #5 - Steam flow high with low-low T-AVG: add 1.75 sec. However, as discussed below although these T.S.'s times are affected, there is no change to any
-design bases since the total system response time remains unchanged.
The above Technical Specification response times do not include a 2.0 second delay, which is analytically established, to account for the existing RTD bypass loop thermal lag and travel time. With the proposed system, this component is reduced by 1.75 seconds to 0.25 seconds. The 0.25 seconds accounts for the transient time and thermal lag of the hot leg mixing scoop.
~28-
With the propsed system, the response time of the RTD's will be determined with the RTD's in the thermowells (using loop current step response methodology); therefore,-the thermal Tag associated
.with the thermowell will be included in the RTD tested response time. The response time of the proposed RTD/thermowell is 1.75 seconds slower than the existing direct immersion RTD's response time.
Tabuiating the overtemperature delta-T response time components: Existing Proposed Direct Immersion RTD 3.0 sec. N/A Combined RTD/Thermowell N/A 4. 7 5 sec. Eleqtronics/Electrical 1.0 sec. 1.0 sec. Delays Subtotal (T.S. Response Time) 4.0 sec. 5.75 sec. Loop or Scoop transient and thermal lag *
- 2.0 sec. 0.25 sec.
Total System Response Time 6.0 sec. 6.0 sec. The affected ESF time responses are similarly tabulated in Tables 7-1 and 1~2. In all cases, the total response time remains unchanged. These times bound best estimate response times. The allocated time for RTD/thermowell includes a 10% error allowance for LCSR testing. Since the total system response time remains unchanged, no other evaluations/analyses are required to demonstrate the acceptability of
*changing the aforementioned T.S. times.
Supplemental information sent with LCR 83-19 (amendments 60 and 31 Units 1 and 2, r~spectively) provides an evaluation for an increase in system response time of 3.0 seconds. This information remains applicaple to the proposed modification and is referenced in this report so as to support the conservatism of the existing times. 7.3 Relocation of RTD Instruments: I The function of the RTD's in the bypass piping manifolds is to measure the RCS hot leg and cold leg temperatures. Accordingly, physical relocation of the RTD's .into thermowells mounted directly in the RCS piping is consistant with the function of the RTD's. At the proposed location, the RTD thermowells will be directly in the RCS flow path and not ~ave to rely on a subsystem~ 7.4 Reactor Coolant System Flow: Elimination of the RTD Bypass Piping System will have a very slight increase of approximately 0.1% in the flow ~hrough the Reactor Vessel and Steam Generator. Although this flow increase thereotically results .in better heat removal at the core and better heat transfer at the Steam Generator, the change is too small to be significant *
** TABLE 7.1 Response Times of ected ESF Functions
*with Existing RTD Configuration Response Time Time Delay in Seconds for Impacted ESF Functions Component Rx. Trip( 2 ) FWI(J) sws< 5 >
Scoop Transport and o.o<~> o.o< 0 > o.o< 0 > Thermal Lag Bypass Line Transport 2 2 2 2 2 2 2 and Thermal Lag Direct Immersion 3 3 3 3 3 3 3 RTD Response Time Combined Thermowell N/A N/A N/A N/A N/A N/A N/A and RTD Response Time Diesel Generator 10 N/A N/A 10 35 N/A N/A I w I-' Startup and Loading (9) I 6 57 Electronic and 11 1 6 16 11 Actuated Component Delay Time Total Response Time 16.0*/26.0** 6.0 11.0 21.0*/31.0** 16. 0*/51. O** 11.0 62 Tech Spec LCO 14.0*/24.0** 4.0 9.0 19.0*/29.0** 14.0*/49.0** 9.0 60 Response Time Notes
~Safety Injection (2) Reactor Trip from Safety Injection (3)
- Feedwater Isolation I.
(4) Containment *solation, Phase A (5) Service Water System (6) Steam Line Isolation (7) Auxiliary Feedwater (8) Included with bypass line transport and thermal lag (9) Required when offsite power is not available
- With offsite power available
** *Without offsite power available
TABLE 7.2 Response Times o ected ESF Funct;ons with Proposed RTD Configuration Response Time Time Delay in Seconds for Impacted ESF Functions Component SI(l). 3 Rx. Trip( 2 ) FWl( ) CI( 4 ) sws< 5 > SLI ( 6 ) AWF(7) Scoop Transport and .25 .25 .25 .25 .25 .25 .25 Thermal Lag Bypass Line Transport N/A N/A N/A N/A N/A N/A N/A and Thermal Lag Direct Immersion N/A N/A N/A N/A N/A N/A N/A RTD Response Time Combined Thermowell 4.75 4.75 4.75 4.75 4.75 4.75 4.75 and RTD Response Time Diesel Generator 10 N/A N/A 10 35 N/A N/A 11 w N Startup and Loading (8) I Electronic and
- 11 1 6 16 11 6 57 Actuated Component Delay Time Total Response Time 16.0*/26.0** 6.0 11.0 21.0*/31.0** 16.0*/51.0** 11.0 62 Tech Spec LCO 15.75*/25.75 5.75 *10.75 .20.75*/30.75** 15.75*/50.75** 10. 75 61. 75 Response Time Notes
~Safety Injection (2) Reactor Trip from Safety Injection (3) Feedwater Isolation (4) Containment Isolation, Phase A (5) Service Water System (6) Steam Line Isolation (7) Auxiliary Feedwater (8) Required when offsite power is not available
- With offsite power available
**
- Without offsite power available
8.0 UNREVIEWED SAFETY QUESTION (USQ) DETERMINATION Title 10 of the Code of Federal Regulations Section 50.59 states that the l{censee may: (i) make changes to the facility as described in the safety analysis report, (ii) make change~ in the procedures *as described in the safety analysis report, (iii) conduct tests or experiments not described in the ~afety analysis report without prior Nuclear Re~ulatory Commission approval unless *the proposed change, test or experiment involves a change to the technical specifications incorporated in the license or an unreviewed safety question. A proposed change involves an unreviewed
. safety question if, a) the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased, or b) the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created, or c) the margin of safety as defined in the basis for any technical specification is reduced.
The plant change proposed herein involves the replacement of the existing Reactor Coolant System narrow range Resistance Temperature Detectors (RTDs). Although the proposed change does involve a change of Technical Specifications 3.3.1 and 3.3.2, it is convenient to employ the definition
*of unreviewed safety question from 10 CFR 50.59 to document that there is no safety concern.
As described in the Updated Final Safety Analjsis Report (UFSAR) Section 5.6.l, Reactor Coolant System (RCS) hot and cold leg temperatures are measured by narrow range direct immersion RTDs located in bypass manifolds. Through the use of a bypass manifold around each steam generator, hot leg temperatures are obtained by mixing the flow from three scoop connections which extend. into the flow stream at locations 120° apart circumferentially. Flow for the cold leg manifold is obtained
- downstream of the pump. Both hot leg and cold leg bypass flows enter a common return line to the cross over leg. UFSAR Figures 5.1-6A and B illustrate the existing configuration.
As discussed in UFSAR section 7.2.3.2, the existing RTD temperature outputs are used for a number of purposes. They are used oy the Reactor Protective System for the Overt~mperature Delta~T (OT~T) and Overpower Delta-T (OP~T) trip functions. - Reactor control _ is based upon Tavg signals derived from protective system channels. An engineered safety feature (ESF) actuation and steam line isolation is actuated on steam flow in two steam lines - high coincident with Tavg-low-low. The output from the narrow range RTDs are used to calculate the Tavg input to this ESF actuation. 'The Tavg signals . are provided to the pressurizer _level control system, the steam dump control system, the reactivity computer, and certain interlocks. The functions that utili~e temperature input from the existing narrow range RTDs will not be affected by their proposed removal and replacement because the signal~ derived from the proposed
- replacements will be equivalent to those provided by the existing RTDs.
Only one of the functions which use signals from the existing RTDs, the OT~T, is a primary trip credited in the accident analysis (UFSAR Chapter 15). The steam flow in two steam lines - high coincident with Tavg-low-low ESF actuation provides backup protection _for the steam line break accident. The response times of these functions will remain within that assumed in the accident analyses. As such, the consequences of the accident analyses will not be affected. The proposed change involves removal of the existing-bypass lines and replacement of the existing RTDs with ther~owell RTDs. Three dual eleme~t RTDs will be used for each hot leg. These will be located in the existing scoops. One dual element thermowell RTD will be located in the existing cold leg penetration. The nozzles in the crossover legs for the return lines will no longer be needed, and they will be capped. The proposed replacement of the existing"RTD and bypass line elimination does not in~61Ve an unreviewed safety question because: The removal and replacement of the existing RTDs will not increase the* probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report (SAR). The consequences of an accident or malfunction of equipment important to safety previously evaluated is considered first. There are four events of interest; (1) Uncontrqlled Boron Dilution During Full Power, (2) Los~ of External Load, (3) Uncontrolled Withdrawal of a Control Rod at Power, and (4) Major Secondary Pipe Rupture. The first three. ~vents are of interest because the OT6T trip is the primary trip credited in the safety analyses. The fourth event is considered because steam flow in two steam lines high coincident with Tavg-low-low is one of the sign~ls that could be credited to initiate an Engineered Safety Features actuation. The OT6T trip will
~ontinue to function in a manner consistent with the existing ~nalysis assumptions for the first three events .. The total actual response time will be within the six seconds currently assumed. Similarly, the ESF .1 I
response times will remain within those assumed in the safety analysis. , I Hence there will be no increase in the consequences of an accident or malfunction of equipment important to safety previously evaluated. There will be no increase in the probability of occurrence of an accident or malfunction of equipment important to safety pre~iously e~aluated in the SAR. The events of interest are those initiated by a failure of those systems that use temperature inputs from the narrow range RTDs or could be initiated by a mechanical failure of components affected by the ~reposed change. There are four such events. These are: (1) Uncontrolled Withdrawal of a Control Rod at Power, (2) Excessive Load
*r~crease, (3) Accidental Depressurization of the Main Steam System, and (4) Small Break Loss of Coolant Accident (SBLOCA). The Uncontrolled Rod Withdrawal event is an ANS Condition II (moderate frequency) event potentially initiated by a failure of the reactor control system. The Excess Load and Accidental Depressurization of the Main Steam System events are also Condition II events. They are ~otentially initiated by a failure of the steam dump control system. The input to the reactor
- control system and steam dump control system from the 1
repl~~ement RTDs will be equivalent to those currently provided by the existing RTDs. The proposed modification will be done in a manner consistent with the plant design bases. As such there will be no degradation in t~e performance of. or increase in the number of challenges to equipment assumed to function during an accident situation. Furthermore, there will be no increase in the probability of failure, or degradation, of the performance of the systems designed to ~educe the number of challenges to equipment assumed to function during an accident situation. Hence the first three events will remain Cdndition Il events.
'The SBLOCA is an ANS Condition III (infrequent) event. It could be initiated by the highly unlikely ejection of a thermowell or the failure of one of the caps that will cover one of the existing cross over leg penetrations. The removal and. replacement of the existing RTDs will result in 6nly two additional RCS penetrations for the reactor vessel level instrumentation system (RVLIS). All ch~nges will pr~serve the qualification of the Reactor Coolant System pressure boundary. The scoops, cross over leg nozzle caps, RVLIS connections, and thermowells will be analyzed to the ASME Boiler and Pressure Vessel Code, Section III, Class 1 and installed in accordance with Section XI of this code . . The diameter of the RVLIS penetrations will be 0.375. inch. Breaks of
- this size are considered leaks rather than SBLOCAs, since the normal charging system can maintain normal RCS inv~ntory so that RCS pressure and pressurizer level do not decrease. The SBLOCA will thus remain a Condition III event. Hence, there will be nc>>increase in the probability of occurrence of an accident or malfunction of equipment important to safety previously evaluated in the SAR. Additionally, approximately 280 feet of small diameter pipe and the.associated valves will be removed
'from the primary system pressure boundary, thereby. elimfnating the possibility of a SBLOCA caused by a failure in this section of piping.
An increase in the measured response times will not increase the probability of occurrence of a previously evaluated accident or malfunction of equipment important to safety because the numerical valu~ of response time is not the i~itiator of such an everit. The total response time of the functions that use signals from the narrow range
- RTDs will be unchanged after the proposed modification.
The probability.and consequences of flooding and jet impingement have been reviewed. The thermowells, caps, and Reactor Vessel Level Instrumentatiori System (RVLIS)' corinections will be in the same or immediate locations of the existing RTD bypass loop connections. Therefore, the consequences of a postulated flooding -0f the proposed RTDs or the impingement of a jet on the proposed RTDs are bounded by the results of existing analyses .. Th~re is no increase in the probability of flooding or jet impingement as the number of components and welded joints will be reduced considerably. The consequences of a missile due to the postulated ejection of a thermowell has been reviewed . . The cold leg thermowells are considered first. These are located just beneath massive Rector Coolant Pump (RCP) support *beams, which would stop the travel of such a postulated missile. No appreciable damage to the support beams would result. These supports are designed to limit RCP motions in the event of a complete double ended break of a cold leg, which is not postulated to occur simultaneously with a thermowell ejection. This new cold leg thermowell will be adjacent to and at the same orientation as the wide range RTD thermowells which have been previously reviewed. A missile created by the postulated ejection of the hot leg thermowells is considered next. If one of a coolant loop's two lower hot leg RTD thermowell were eje~ted, it would stop its travel after striking the fl6or below ~t (at elevation 81 feet). There would be no significant damage, as there are no** vital components between the thermowells and the floor. Similarly, a missile created by the postulated ejection of a coolant loop's upper thermowell would strike the bqttom of .the operations deck above it (at elevation 127 feet). Again, there would no significant damage as no vital components are in the
.direct path of the postulated missile. -As such; the consequences are acceptable. Th~ proposed thermowells to be mounted on the top of the hot leg piping wi 11 be adjacent to and at t~~ same orientation as the Thot wide range RTD thermowell which was previously reviewed .
- The possibility of an accident or ~alfunction of a diffe~ent type than any evaluated previously in the SAR is not created. The proposed changes will be pe~formed in a manner c6nsistent with the applicable standard~,
. preserve the existing design bases, and will not adversely impact the qualification of any plant systems. This will preclude adverse control and protection system interactions. The design installation, and inspection of the new equipment will be done in accordance with ASME Boiler and Pressure Vessel Code. criteria. By adherence to industry standards, the pressure boundary integrity will be preserved. Hence, the possibility of a different type of accident than any evaluated in the safety analysis will not be created.
There will be no reduction in the margin of safety as defined in the bases of any teth~ical specification. The applicable margins of safety are defined in Bases Sections 2.1.1 and 2.1.2. Bases Section 2.1.1 states that the minimum value of the Departure from Nucleate Boiling Ratio (DNBR) during steady state operation, normal operational transients, and anticipated transients is limited to 1.30. This value corresponds to a 95 percent probability at a 95 percent confidence level that Departure from Nucleate Boiling (DNB) will not octur. The restrictions of this fuel cladding integrity safety limit prevent overheating of the fuel and possible cladding perforation which would result in the.release of fission products to the coolant. The minimum DNBR reported in the accident analyses will be unaffected by the*pr6posed change. Bases Section 2.1.2 states that the Safety Limit on maximum RCS pressure is 2735 psig .. This Safety Limit protects the integrity of the RCS from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere. The maximum RCS pressure reported .in the accident analyses is unaffected by the proposed change. The proposed change will not result in a decrease of these margins of safety. As discussed earlier, the response time and setpoint of the ESF actuation function and the OT T trip function will remain within the assumptions used in the safety analyses~ As such, the analysis of the events which credit these functions will remain as presented in the UFSAR. Consequently, the margins of saf~ty between the Fuel Clad~ing Safety Limit (i.e., DNBR) and RCS pressure boundary Safety Limit and the actual failure of these barriers will not be.reduced. In conclusion, the removal and replacement of the existing narrow range RTDs and elimination of the by~ass lines does not involve an unrevi~wed safety question. Neither the probability of occurrence nor the consequence of an accident or malfunction of equipment important to safety previously evaluated in the SAR is created. The possibility for an accident or malfunction of a different type than any evaluated previously in the SAR is not created. There will be no reduction in the margin of safety as defined in the basis of any technical specification . APPENDIX A LICENSING REPORT S-87-05 PROPOSED*FSAR CHANGES:
- 1) Sections 5.6 and 7.2.3.2 - Revise description to reflect elimination of the bypass piping and use of thermowells mounted ditectly in the main piping.
- 2) Figures 5.1 - 6A and 68 - Revise to reflect elimination of the bypass piping and use of
-thermowells mounted directly in the main piping.
Also, drawing changes reflect that RVLIS will be connected directly to the hot leg piping.
~) Table 15.2-1 ~ Revise times in the analyses to reflect reallocatiori of times. specifically, change reflects*
that the electronic processing time is 1.0 second *
.Y
\ .
INSTRUMENTATION APPLICATION 5.6 1.".! "\ Process control instrumentati"on is provided for the purpose of acquiring *
- . . . *~~*:*
data. on the- pressurizer and on a per loop basis for the key process parameters of the RCS (inc 1udi ng the reactor pump motors) as we 11 as for the residual heat removal system. The pick-off points for the reactor coolant system are shown in the ~low diagram (Figures 5.l-6A~ B and C); and \ *
~ .....
for the residual heat removal system, in flow diagram Figures 5.5-2A and B. In general these input signals are used for the following purposes:
- 1. Provide i"nput to the reactor trip system described in Chapter 7.
- 2. Provide input to the engineered safety features actuation system described in Chapter 7.
- 3. Furnish input signals to the non-safety related control systems an*d, -
surveillance circuits. 5.6~1 LOOP.TEMPERATURE
-r ,.,_ .:Jr- "
sistance Temperature Detector Bypass Manifold is provided for e reactor lant loop hot and cold leg. A bypass manifold aro steam
.generator obt
- hot leg temperature by mixing the flow om three scoop
'fonnections, which e . nd into the flow stream at ations 1200 apart in the cross-sectional plane, the reactor co nt leg. The hot leg bypass flow exits the manifold to a co ret the pump
- Flow for the cold leg bypas e mixing action of the pump,* nly one connection is required to o in a representative sample. This conne
- n is located ossible to the pump discharge in order to minimize bypass pipi and to obtain optimum fluid mixing. This conne~tion is in the s
... -~-**- -** .. - ------- .......------~- ___
elative position in each loop *
,_.__:..__ ,. __ .., _____._ **-**- --**-*"'- *-- *- -- - . - ...--.--. ---- - -- _..----.........._ -- -~------* -- - ...
5.6-l Revision 4* SGS/UFSAR July 22, 1982
--- ----**--~-*---;*--*---
__........ ---- -- - ----*-*** ----** - --~-
- INSERT A One hot leg and one cold leg temperature reading is provided from each coolant loop to use for protection. Narrow Range thermowell Resistance Temperature Detectors (RTDs) are provided for each coolant loop: In the hot legs, sampling sco~ps are used because the flow is stratified. That is, the fluid temperature is not uniform over a cross section of the hot leg. One dual element RTD is mounted in each of the three sampling scoops associated with each hot leg.
The scoops extend into the flow stream at locations 120° apart. tn the cross sectional plane. Each scoop has five orifices which sample the hot l*g flo~ along the leading edge of the scoop. Outlet ports are provided in the scoops to direct the sampled fluid past the sensing element of the RTDs. One of each RTDis dual elements is used while the other i~ an installed spare. Three
- readings from each hot leg are averaged to provide a hot leg reading for that loop.
One dual element RTD is mounted iri a thermowell associated with each cold leg: No flo~ sampling is needed because coolant flow is well mixed by the reactor coolant pumps. One RTD element is used while th~ other is an installed spare. The thermowells are pressure boundary parts whi~h completely enclose the RTD. They have been shop hydrotested to 1.25 times the RCS design pressure. The external design pressure and temperature are the RCS design temperature and pressure. The RTD is not part of the pressure boundary. T~e scoop, thermowell, and thermowell/scoop assembly have been analyzed to the ASME Boiler and Pressure Vessel Code, Section III, Class 1. The effects of seismic and flow induced loads were considered in the design. resistance temperature detectors extend directly (witho thenno wells) to reduce the time delay to. a minim1J11. lation val ** in series are provided on each side of the .. ::: to allow for. res The valve -~* :.~:
- the ma_i n coolant pi
- ng is 1ocated above the elevation of the reactor ve In addi- ......
-~ ..
~*
ti on, vents and drains are used, in conjunction with the ..' old bypass manifold join to fonn a canmon
- bined (hot and cold leg) flow passes through a flow i cator i.
discharging to the suction side of the reactor cool ant pump. * '. L .* r**.
.Sc.,-~ TC."T'C s, ~-* .
t\Te1t1'9!F'li&:e ae*H,11'9\ located in the thenno wells in the cold and hot leg piping of each loop, supply signals to wide-range temperature recorders. This i nfonnati on is used by the *operator to control cool ant
- temperature during startup and shutdown *
- 5. 6. 2 PRESSURIZER TEMPEAATU1£ There are two temperature detectors in the pressurizer, one in the steam phase and one in the water.phase._ Both detectors supply signals to temperature i ndi caters and high-temperature a 1anns. The steam-phase de~ector, 1 ocated near the top of the vesse 1, a.1 erts the operator if the steam becomes superheated. In addition, it is usedduring startup to detennine water temperature when the pressurizer is complet_e1y filled with water.* The water phase detector, located at an elevation near the center of the heaters, .is used during cool down to* ensure that the pres-surizer temperature is consistent with the Reactor Coolant System.
Temperatures in tre pre-ssurizer safety and relief valve discharge lines are measured and indicated. An increase in a discharge line temperature. is-an indication of. .leakage through .:the .. associated valve. An alann is *
- .. **-* . - . - -- - ~.
actuated on high temperature.
--***. ~GS-:UFSAR 5.6-2 Re vision 0
*-***** **---**--* - -* **-*-. **-*. --...---*--- -- July .. 22, 1982 ______ _
- 7.2.J.2 Specific Control and Protection Interactions Nuc 1ear F1ux
/,..-_
Four power-ranye nuclear flux channels are provided for overpower pro-tection. Isolated outputs from all four channels are auctioneered for automatic rod control.* Lf any channel fails in such a wa.y as to produce a low output, that cj1anne l is i neap ab le of f)roper overpower protection. In principle, the same failure may cause rod withdrawal and 11ence, over-power. rwo out of four overpower trip logic wi 11 ensure an overpower trip if needed even with an independent failure in another channel. In addi don, tne control system wi 11 respond only to rapid changes in indicated nuclear flux; slow changes or drifts are compensated by the temperature control siynals. Fjnally, an overpower signal from any nuclear channel will block automatic rod withdrawal. fhe set point for this rod stop is below the reactor trip set point
- Coolant Temperature leg and one cold ley temperature measurement is made fore reacto coolant loop to provide protection. In addition, OJ lifiers located in the protection rack, the di ff ere nee mea ements for each loop channel per loop /4 reactor trip system uetectors (RTu'si are inserted into reactor cool a ps - a uypass loop from upstream of the steam generator downstr~arn of t steam generator is used fur the llot ley RTO' s d bypass loop from down earn of the reactor cool-ant µump to u ream of the pump is used for the c rhe ocated in rilanifolds \~ithin the containment a into the reactor ci.>olant bypass loop ermo\.ells are not used in urder to keep the detector thennal SGS-UFSA~ 7.t!-28 ~evi si on IJ July I.I., l98l
-4'%**
INSERT B One hot leg and one cold leg temperature reading is provided from each coolant loop to use for protection. N~rrow Range th~rmow~ll Resistance Temperature Detectors (RTDs) are provided for each coolant loop .. In the hot legs, sampling scoops are used because the flow is stratified. That is, the fluid temperatur~ is-not uniform over 'a cross section of the hot leg. One dual element RTD is mounted in each of the three sampling scoops associated with each hot leg. The scoops extend into the flow stream at locations 120° apart in the cross sectional plane. Each scoop has five orifices which sample the hot leg flow along the leading edg~ of the scoop. Outlet ports are provided in the scoops to direct the sampled fluid past the sensing element of the RTDs. One of each RTDs dual elements is us~d for protection while the other is an installed
- spare. Three protection readings from each hot leg are averaged to provide a hot leg reading for that loop.
One dual element RTD is mounted in a thermowell associated with each cold leg. No flow sampling is needed because coblant flow is well mixed by the reactor coolant pumps. As is the case with the hot leg, one element is used while the other is an installed spare. Certain.control signals are derived from individu~l protective channels through isolation amplifiers. The isolation amplifiers are classified as a part of the protective system. The reactor control system use~ the highest of four isolated Tavg signals. Th~ RTDs are a ~ast response design which conform to the applicable IEEE standards and 10 CFR 50.49 requirements.
s The bypass.arrangement pennits replacement of defective ture lements_ ~ile the p 1ant ; s at hut shutdown wi tnout draining or depres izi."9 the reactor coolant loops. T11ree sam\J 1i probes clre inst"lled in a cross-sectional imate ly 12\l degree intervals.
- Each of th
- hot leg at app i)robes, 'fttlich ext ds se~erCl 1 inches into t11e hot leg c 1ant stream, ifices Ji stributed along its 1 gth. In this way a tu ta l of fifteen luc a
- ons in tne not leg sere cire sampled providing a representative cool ant t perature measureme fhe t\'.O inch di am<<!ter pipe lealii ng to tile manifold anted ni ng the mperature mea~uri ng e 1e-
- ments (RTu 1 sJ provides mixing s to ~ive an accurate temp~r-ature measurement.
the orifices. This i1as been don by designing a aller overall orifice
- flow area than that of th common flow channel
- tni n c.ne iJrobe. rhi s arrangeml.!nt has al so n applied to tile flow tran *ti on from tne three probe flow channels o the pipe leading to the temp~r ure element manifold. The to al flow area of the three probe ci1anne has therefore 11 been JesiyneJ oe less than that of i;he 2 pipe connecti n the probes to the mani f ld.
leg primary coo*lant flow is nell mixed by the reactor cool fhi:refore, the cold le~ sample is tdken dire~tly from inch pipe taµ off the cold leg downstream of the iJU1np. f\1e main requi ~ent fur rei.l.Ctor iJrOtection i S that the ternµerature difference bet~en clle not ley and cold leg vary li nearlj- wi tll µower. All 6T setpoints are in tenns of tl1e full po....er 6r; thus, aosolute llr measurements a.re not required. L. i neari ty of 6 r ~Ii th f-'Ower wi 11 be verified during startup tests .
- . S(iS~llrSAR .]
- l-"t.9 i{evision u July a, 1982
- Reactor protection loyic using reactor coolant loop temperatures f s 2/4 -:
(*.: with one channel per reactor coolant loop. fhi s complies with all app 1icab1 e IEt:t: 27~ c r1 teri a. Reac
- sed upon signals derived em channels after isolation sucn t:1ui; no feedback perturb the protection. channe 1s.
- re.a.."'"9or" -
.)ince~control is based on the hignest average temri~rature from the four loops, the control rods are al~s moved .
based UiJIJn the most.pessimistic
~
temperature measurement with respect to margins c.:> UN8. A spurious low average temperature measunnent from a'1)' loop ~ernperature control channel wi 11 cause no control actioil. A spurious hig.i average temperature measurement will cause* rod insertion (safe Jirection). ividual low flow alanns with individuul status li~hts for e*ach
*coo 1ant loop bypuss flow is provided on the main contra board.
The, alar and status lights provide tneop~rator with il'llRed' e inc.Jica:.. in cne llypass loops associ d with any reactor coo 1ant oop.
,,r Local indicators are p vided to ..1onitor tota low through the IHIJ bypass manifolds for each uop. Che indic are located inside con-tai rment but are accessible ing po F1ow wi 11 be mo ni to red : "-
- 1. Prior to restori n*3 te erat1.1re chann~'ls'*to*,. normal service fol lowi rig reopening of bypa s loop i so 1ati on valves \l.on~never a bypass loop has
- 2. i.ln a .,ier* die basis.
- 3. bypuss loop low flow alarm l see dbove) *
- SG~-UFSAR 7.2-30 Revision 0 July l2, 1982
- Iii aa1H'eh"' ~hannel deviation signals in the control system will give I
***-* an alann . ;f "ny-~. tem~erature cilannel deviC&tes s;gnificantl.t from the auctioneered (hi_ghest). Automatic rod withdrawal blocks will also occur if any one of four nuclear channels indi_cates an overpo~r condition or if. uny two of four tamperature channe 1s- indicate an overtemp~rature or overpo....er condition. rwo-out-1Jf-four (2/4) trip logic is used to ensure th at an overtemiJerature or overpo~r .1 T trip wi 11 ace ur if needed even with ail independent failure in clnotiler channel. rinally, as shown in Sectilrn 15.l, tne co1nuination Jf trips* on nuclear overpv....er, and hi~h 1-1ressurizer ~res~ure also serve to limit an excursion for anJ rate of reactivitJ insertion.
Far up~ration witn d loop vut of service only one safety relateu set-poi nt must be mu11ual1J Nset to a more restrictive value. fhe setpoint involved is the overtem!)erature t.f reactor trip. ihe setpoint cnanye must be made to .one protection channe 1 for each of ~lie operating loops. If the avertemperature t.f setpoints can be reset (lowered) before turn-
. i ng off one pump, tne sett-iai nt s110uld be reset duri n*J operation with al 1 loops in service. If the overtamperature t.r setpoints cannot oe reset without cau*sing a reactor crip before turnin:j off one pump, reactor ponier should be reduced below the setpoi nt of P-8, t11e affected pump turned off, and tile setiloi nts reset. Any time one pilnp is turned off or tri~s off W1en above P-8, an automatic reactor trip will occur.
rhe P-8 de.ts essentiully as a high neutron flux rea..;tor trip .,.,;,en oper-ating witn one loop no~ in service. rne f'-d sctpai nt .,,; 11. nonnal l.t be set in suc11 a waJ c11ut t11e u1~dr{ is aoove 1.30 l for anticiµateJ transients) even wi tllout resetting the over-
. temperature t.f trips to the values ap1-1ropriate for uperation witn a loop al.it of service. Setting P-!j in _tl1i s wqy restricts the operati n~ pow.er level with a loop out of service to i1 *-1 ...ilue consiJarauly lo'tJer thdn chat
** wnich can be.safely allo\'ted afte.r resettiny *i:he overti:mperature t.r set-
- poi nts. After ti1e overtemper<Jture t.f setpoi nts 11ave been reduced to the SGS-uf SAK 7.2-31 Ke vision O July l~, 1982 .
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~--* TABLE 15.2-1 (Sheet 2 of 10) TIME SE~ENCE OF EVENTS FOR CONDITION II EVENTS Event Time (sec.) Accident Uncontrolled RCCA Initiation of uncontrolled Withdrawal at RCCA withdrawal at maximum Power reactivity insertion rate . (7.5 x 10- 4 ~K/sec.) a
- l. Case A
.t(.
.~
Power range high neutron flux high trip point reached 1.5
\
Rods begin to fall into core 2.0 I! Minimum DNBR occurs 2.7 I
- 2. Case B Initiation of uncontrolled RCCA withdrawal at a small reactivity insertion rate (3.0 x 10- 5 ~K/sec. for 3 loop, 3.0 x 10-S ~K/sec.
-~
for41oop) a Overtemperature ~T reactor trip signal initiated Rods begin to fall into core 34.6 Minimum DNBK occurs 34. 7 ( Revision a July 22, 1982 SGS-UFSAR r
, ...* TABLE lS.2-1 (Sheet J of 10) TIME SE()JENCE OF EVENTS FOR CONDITIOH II EVENTS I:.
\
Event Time (sec.) Accident Uncontrolled Boron unution
- 1. Dilution during refueling and Dilution begins 0 startup Operator isolates source of dilution; minimum margin to criticality occurs
-2400 or more t
I
- 2. Dilution During Ful 1
.. Power Opera ti on
- a. Automatic One percent shutdown r:largi n f.
Reactor
-1300 Control 1 ost I* .
- b. Manual
~ Reactor Control Dilution begins 0 II
\
~ii;t;F tFi~ iiitpeiR;; Feasf!eel w 6vertemperature 6.T ~erv
.,.~ ,,~ *,~ ~%i,P I \I Rods begin to fall into core 54 One percent shutdown is lost (if dil~tion c6ntinues) after trip) -goo
- Revision O SGS-UFSAR July 22, 1982
\
I
- ~-** TABLE 15.2-1 (Sheet 5 of 10) i
\
TIME SE~ENCE OF EVENTS FOR CONDITION I I EVENTS Accident Events Time (sec.) \' I Loss of External rJ ~ctri cal Load
- 1. With pressurizer Loss of electrical load 0 control (BOL)
Initiation of steam release from steam generator safety valves 9.0
/D, I Overtemperature ti.T ~OU+~ ~
\
S-1 ~ 1,...;f1 Ja..../ Rods begin to drop 11.1 Minimum DNBR occurs 11.5 Peak pressurizer pressure occurs 12.5
- 2. With pressurizer control ( EOL) Loss of electrical load a Initiation of steam rel~ase from steam generator I
safety valves 9.0 Overtemperature ti.T Reactor /D * $"'" Trip .Pei At ~eaeAeti :~ 1~
**~
I 1~1+1~ ~ l' *. Rods begin to drop
- 11. 5
-* . - -*------**-*- . *Revision* a SGS-UFSAR July 22, 1982}}