ML20010E608
| ML20010E608 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek, Callaway  |
| Issue date: | 09/01/1981 |
| From: | Petrick N STANDARDIZED NUCLEAR UNIT POWER PLANT SYSTEM |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| SLNRC-81-82, NUDOCS 8109080036 | |
| Download: ML20010E608 (48) | |
Text
.
e SNUPPS Standardized Nuclear Unit Power Nnt System 5 Choke Cherry Road Nicholas A. Petrick Rockville, Maryland 20850 Executive Director (301)869 8010 September 1,1981 SLNRC 81- 82 FILE: 0541 SUBJ: ICSB Review c1 NU 'W vfir. Harold R. Denton, Director Office of Nuclear Reactor Regulation D \ '
U.S. Nuclear Regulatory Commission U g .\$g g\"
Washington, D.C. 20555 %
Docket Nos.: STN 50-482, STN 50-483, and STN 50-486 gj'
Reference:
- 1. SLNRC 81-68, dated August 14, 1981, Same subjeb h ' G U
- 2. SLNRC 81-67, dated August 14, 1981, ICSB Positions -
Dear Mr. Denton:
The referenced letters provided information concerning, and the status of, the agenda items and positions involved in the Instrumentation and Control Systems Branch review of the SNUPPS FSAR. This letter provides the information to close all review matters which have not been closed previously by meetings, correspondence, and FSAR changes.
- 1. ICSB position #6 concerned reactor coolant temperature indication at the auxiliary shutdown panel. Reference 2 provided a response to the position. Enclosure A to this letter is a revised response to position #6.
- 2. ICSB position #8 concerned balance-of-plant instrumentation and controls for safety functions. As a clarit ication to the response to position #8 (see reference 2), the following information is pro-vided. The instrument loops discussed in Table 8-1 can be tested without interfering with normal plant operations and without lif ting instrument leads or using jury rigs. For Table 8-1, item D, an alarm will be provided to indicate that the setpoint of the safety function has been reached.
- 3. Agenda item #45 concerned postulated boron dilution events. The voo3 v
SNUPPS design will incorporate the same design changes that were I reviewed by the NRC on the Comanche Peak docket for mitigation of inadvertent boron dilution. These changes include seismic and en-vironmental qualification of the source range and intermediate range /[
nuclear instrumentation.
- 4. Agenda item #50 concerned setpoint methodology. Reference 1 pro-vided a response to the item f or the balance-of-plant. The following clarification is provided for item 50, part "e". Safety related BOP 8109080036 810901 PDR ADOCK 05000482 A PDR
9 SLNRC 81-82 Page Two instrumentation that could be subject to hostile environments will undergo equipment qualification for that environment. The design goal and the ebjective of qualification testing is that the instru-mentation will remain operable for one year following a design basis event.
- 5. Agenda item #12 concerned sensors and circuits in non-seismically designed structures. An FSAR change was provided in Reference 1 and Revision 6 to the SNUPPS FSAR. A further clarification con-cerning cable routing and sensor qualification is provided with the FSAR changes included herein as Enclosure B. Enclosure B will be incorporated in Revision 7 to the SNUPPS FSAR.
- 6. Agenda items #33 and #35 concerned display instrumentation. In-cluded in the Enclosure B FSAR changes are text and table additions that close these agenda items.
- 7. Agenda item #39 concerned low temperature overpressure protection.
Included in the Enciosure B FSAR changes are revised descriptions of the SNUPPS design. The NRC questioned the failure modes in this system. The pressurizer PORV block valves are normally open and are designed to close upon loss of opening signal. The pressurizer PORVs are normally closed and are designed to utilize energize-to-open logic. The PORVs are designed to close upon loss of energy or loss of opening signal. Therefore, loss of opening signal will result in both PORV and PORV block valve closure, and accidental RCS depressur-ization is precluded. Since each PORV and its associated block valve are on a separate power supply from the other PORV and its block valve, the pressurizer pressure relief system is designed such that no single failure can prevent opening of at least one flow path to the precsurizer relief tank and except for postulated second random failures, all flow paths to the pressurizer relief tank can be closed.
- 8. Agenda item #51 concerned steam generator water level reference leg heatup and the resulting errors in indicated level. Enclosure C is a report describing this matter for the SNUPPS design.
- 9. Agenda item #52 concerned the interface criteria given in WCAP-8584.
Included in the Enclosure B FSAR changes is a change that addresses this matter.
- 10. Agenda item #60 concerned the potential effects of a steamline rup-ture on the automatic rod controi system. More specifically, the postulated scenario is as follows. Following an intermediate steam-line rupture outside containment, the automatic rod control system exhibits a consequential f ailure due to an adverse environment which causes the control rods to begin stepping out prior to receipt of a reactor trip signal on overpower bJr. This scenario results in a lower DNB ratio than presently presented in the SNUPPS FSAR. However, as discussed below, a typical bounding analysis has been performed which calculated that no fuel damage would occur.
SLNRC 81-82 Page Three This scenario is considered to be very unlikely because, among other reasons, the break must occur at certain power levels, the break must be in a certain intermediate size range, the adverse environ-ment must affect the turbine impulse transmitter, and the transmitter must cause a spurious low power signal without causing a reactor trip.
A typical bounding analysis of the intermediate steamline rupture was performed to calculate the extent of fuel damage due to rod con-trol system withdrawal prior to reactor trip. Based upon the reduc-tion in radial peaking f actor with burn-up and conservative end-of-life physics parsmeters, no fuel damage was calculated to occur fol-lowing the intermediate steamline rupture with a consequential rod control system failure.
A plant-specific analysis of this scenario is being performed for SNUPPS. The results of this analysis (scheduled to be available by the end of 1981) are expected to confirm the results of the typical bounding analysis discussed above.
- 11. Agenda item #69 concerned FSAR terminology used in describing extended hot shutdown from outside the control room. The local actions that are referred to do not involve lifting instrument leads. Appropriate wording changes are included in the Enclosure B FSAR changes.
- 12. Agenda item #13 concerned the analysis for detern.ining if the pres-surizer PORV would lift following a turbine trip from below 50% power.
Included with the Enclosure B FSAR changes are additions to the anal-ysis results that consider the low vacuum turbine trip case.
Ver truly yours, b kM\C R Nicholas A. Petrick RLS/dck/3a9 )
Enclosures A. Revised response to Position #6 B. FSAR Changes C. Report on Steam Generator water level errors.
cc: J. K. Bryan UE G. L. Koester KGE D. T. McPhee KCPL W. A. Hansen NRC/ Cal i T. E. Vandel NRC/WC
Enclosure A
- 6. Concern: Information provided by the applicants indicates that the reactor coolant wide range temperature indicators to be provided on the auxiliary shutdown panel will not meet all criteria applicable to safety related displays (such as being provided power from separate Class IE busses).
Position: The staff position is that reactor coolant system tem-perature is required parameter for maintaining the plant in a safe condition. Indicators meeting criteria applicable to safety related displays should be provided for reactor coolant temperature on the auxiliary shutdown panel.
Response: As indicated in Section 7.4.3 of the SNUPPS FSAR, the reactor coolant wide range temperature indicators are not essential for maintaining safe het shutdown (hot standby). Safe hot shutdown can be maintained from the auxiliary shutdown panel through the use of the essential short-term monitoring indicators and controls listed in FSAR Section 7.4.3.1.1. These indicators and controls meet the criteria applicable to safety grade equipment (see FSAR Section 7.4.3.1.4).
The reactor coolant wide range temperature indicators (one per RCS cold leg) located on the auxiliary shutdown panel provide a highly reliable indication of reactor coolant temperature. These instrument loops are powered from protection sets I and II (loops 1 and 2 from protection set II, loops 3 and 4 from protection set I), isolated at 3 the protection set cabinet, and routed to the auxiliary shutdown ;
panel via separation groups five and six. The indicators are the i same model number as the PAMS indicators provided for the same i function on the main control board. I r
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ENCLOSURE B TO SL f;RC 81-92.
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FSAR CHANGES s
. E 5 6 '2 12 SNUPPS reactor trip. Faults on the first stage turbine pressure circuits would result in upscale, conservative output for open circuits and a sustained current, limited by circuit resistance, for short circuits. Mult_iple failures imposed on these redundant circuitr"could potentially disable the P-13 interlock. In t.his event, the nuclear instrumentation power range signals would
~
provide the P-? safety interlock. Refer to functional i diagram, Sheet 4 of Figure 7.2-1. The sensors for the
. P-13 interlock are seismically qualified.
' Evaluations provided in Section 7.6.1 for the trip fluid pressure transmitter loops indicate that credible electrical faults would not degrade the functional performance of the safety-related BOP instrumentation.
In addition, the following measures will be taken to ensure the integrity of the cabling to the reactor protection system (RPS):
- 1. Inputs from the turbine steam stop valves will originate from four separate limit switches (one per valve), cach of which is dedicated to providing an input to one channe) of the RPS. Cables carrying these signals will be routed in individual conduits. The four circuits will be separated I from one another, from non-Class IE circuits, and identified according to the criteria imposed on Class IE circuits from their source up to their terminations with the RPS cabinets.
- 2. Inputs from the emergency trip oil pressure and P-13 interlock instrumentation will be l routed in a similar manner as are the turbine stop valve inputs.
The logic for this trip is shown on Figure 7.2-1 (Sheet 16).
7.2-9a Rev. 7 9/81
action by manual or automatic means, the standard does not specifically preclude the sharing of initiated circuitry logic between automatic and manual functions. It is true that the manual safety injection initiation functions associated with one actuation train (e.g., train A) share
,. portions of the automatic initiation circuitry
. logic of the same logic t.ain; however, a single failure in shared functions does not defeat the protective action of the redundant actuation train
'. (e.g., train B). A single failure in shared functions does not defeat the protective action of the safety function. It is further noted that the sharing of the logic by manual and automatic initiation is consistent with the system level action requirements of the IEEE Standard 279-1971, Section 4.17 and consistent with the minimization of complexity,
- c. Conformance to regulatory guides and associated IEEE standards Conformance to regulatory guides and associated IEEE standards is provided in Sections 7.1.2.5 and 7.1.2.6.
- d. Failure mode and effects analyses Failure mode and effects analyses have been performed on the engineered safety feature systems' equipment, and the results are provided in Reference 3. The inter-fac criteria provided in Appendices B and C of Refe.'nce 3 have been met in the SNUPPS design.
In addition to the consideraticr. given in this reference a loss of instrument air or loss of component cooling water to vital equipment has been considered. Neither the loss of instrument air nor the loss of cooling water (assuming no other accident conditions) can cause safety limits, as given in Chapter 16.0, to be exceeded.
Likewise, loss of either of the two will not adversely affect the core or the reactor coolant system nor will it prevent an orderly shutdown if this is necessary.
t Furthermore, all pneumatically operated valves and controls will assume a preferred operating position upon loss of instrument air. It is also noted that, for conservatism during the accident analysis (Chapter 15.0), credit is not taken for the instrument air systems
_ nor for any control system benefit. .
The design does not provide any circuitry wh_ich will directly trip the reactor coolant pumps on a_ loss of component cooling water. Normally, indication in the control room is provided whenever component cooling 7.3-51 Rev. 7 9/81
- 2=--- - _-
' SMUPPS M N&
- 6. Suction pressure for each auxiliary feedwater Pump (5)
- 7. Auxiliary feedwater pump turbine speed (rpm)
- 8. Discharge pressure for each auxiliary feedwater
.- pump
- 9. Auxiliary feedwater flow to each steam generator
- 10. Condensate storage tank level
- 11. Reactor coolant (cold leg) wide range tempera-ture
- 12. Source range nuclear power indicators
- 13. Intermediate range nuclear power indicators
- 14. Indicating lights (on-off/open-closed) for all power-operated equipment listed in a. above.
An equipment list for the auxiliary shutdown panel is contained in Table 7.4-1.
7.4.3.1.2 Controls at Switchgear Motor Control Centers, and Other Locations In addition to the controls and monitoring indicators listed above, the following essential short-term controls are.provided outside of the control room with a communication network between these control locations and the auxiliary shutdown control panel:
- 1. Reactor trip capability at the reactor trip switchgear.
- 2. START /STOP controls for both centrifugal charging pumps. Location: Charging pump switchgear.
- 3. START /STOP controls for the component cooling water pumps. Location: Component cooling water pumps switchgear.
$ 4. START /STOP controls for the containment fan cooler units. Location: Cooler fan motor control centers. -
- 5. START /STOP controls for the control room air-condi-tioning units. Location: At the equipment.
7.4-16 Rev. 7 9/81
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. SNuPPs
- 6. START /STOP controls for the diesel generators.
Location: Each diesel generator local control ,
panel.
- 7. START /STOP controls forEssential the essential service service pump water pumps. Location:
switchgear. i 7.4.3.1.3 Controls for Extended Hot Standby In crder to maintain an extended hot standby (greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />),. additional negative reactivity must be added to the RCS.
This can be accomplished by manual control of the normal charging and letdown systems via controls at the auxiliary shutdown ,
panel, motor control centers, switchgears, and control of individual equipment at the device location.
In addition to the normal charging and letdown systems, the systems discussed in Appendix 5.4A may be used to maintain an extended hot standby by local actions outside the control room. l 7.4.3.1.4 Design Bases Information In accordance with NRC Genecal Design Criterion 19, the capabil-ity of establishing a hot standby condition and maintaining the station in a safe status in that mode is considered an essential function. To ensure the availability of the auxiliary shutdown control panel and essential short-term control and indications after control room evacuation, the following design features have been utilized:
- a. The auxiliary shutdown control panel, including all essential short-term instrumentation mounted on it, is designed to withstand earthquakes with no loss of essential functions. The essential short-term local control stations are also designed to withstand earth-quakes with no loss of essential functions.
- b. The essential short-term local stations and the auxiliary shutdown control panel, including essential short-term controls and indicators, are desig:.ad to comply with applicable portions of IEEE Standard 279-1971.
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7.4-17
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TABLE 7.4-1 AUXILIARY SHUTDOWN PANEL EQUIPMENT LIST Instrument Unit Sep.
~ No. No. Service (Group BB PI-455B All Pressurizer Pressure NV BB LI-450B All Pressurizer Level 1 BB LI-460B All Pressurizer Level 4 BB PI-483Z All RCS Pressure (wide range) 4 BB PI-485Z All RCS Pressure (wide range) 1 BB 71S-51B All Pzr Htrs Backup GP A NV BB Hl; ~28 . All Pzr Htrs Backup Gp B NV AB PI-516B All SG A Pressure 4 AB PI-524B All SG B Pressure 1 3B PI-535B All SG C Pressure 4 AB PI-544B All SG D Pressure 1 AE-LI-581A All SG A Level (wide range) 1 AE-LI-502A All SG B Level (wide range) 4 AE-LI-583A All SG C Level (wide range) 1 AE-Li-504A All SG D Level (wide range) 4
l
. SNUPPS MMO'437*
l TABLE 7.4-1 (Sheet 2) Y Instrument Unit *Sep.
No. No. Service ,
Group B PIC-1B All SG A Stm Dump to Atmos Ctrl 1 AB PIC-2B All SG B Stm Dump to Atmos Ctrl 2 AB PIC-3B All SG C Stm Dump to Atmos Ctrl 3 AB PIC-4B All SG D Stm Dump to Atmos Ctrl 4 AB HS-1 All SG A Stm Dump Ctrl Xfr Sw 1 AB HS-2 All SG B Stm Dump Ctrl XFR SW 2 AB HS-3 All SG C Stm Dump Ctrl Xfr Sw 3 AB HS-4 All SG D Stm Dump Ctrl Xfr Sw 4 AB ZL-1B A13 SG A Stm Dump to Atmos Vlv 1 Posn AB ZL-2B All SG B Stm Damp to Atmos Vlv 2 Posn AB ZL-3B All SG C Stm Dump to Atmos Vlv 3 Posn AB ZL-4B All SG D Stm Dump to Atmos Vlv 4 Posn BG HIS- All Letdown Orifice A Isol Vlv NV 8149AB -
BG HIS- All Letdown Orifice B Isol Vlv NV 8149BB
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v TABLE 7.4-1 (Sheet 3) l BG HIS- All Letdown Orifice C Isol Vlv fnr 8149CB BG HIS- All Letdown Ctmt Isol Vlv 4 8152A BG HIS- All Letdown Ctmt Isol Vlv 1 8188A AL HK-5B All SG D Aux Fw Ctrl Vlv Md Pmp B 4 AL HS-5 All EG D Aux Fw Ctrl Vlv Xfr Sw 4 AL ZL-5B All SG D Aux Fv Ctrl Vlv Posn 4 AL HK-8B All SG D Aux Fw Ctrl Vlv to Pmp 1 AL HS-6 All SG D Aux Fw Ctrl Vlv Xfr Sw 1 AL ZL-8B All SG D Aux Fw Ctrl Vlv Posn 1 AL HK-7B All SG A Aux Fw Ctrl Vlv MD Pmp B 4 AL HS-7 All SG A Aux Fw Ctrl Viv Xfr Sw 4 AL ZL-7B All SG A Aux Fw Ctrl Vlv Posn 4 AL HK-8B All SG A Aux Fw Ctrl Vlv to Pmp 1 AL HS-8 All SG A Aux Fw Ctrl Vlv Xfr Sw 1 AL ZL-8B All SG A Aux Fw Ctrl Vlv Posn 1 AL HK-9B All SG B Aux Fw Ctrl Vlv Md Pmp A 1
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TABLE 7.4-1 (Sheet 4)
Instrument Unit Sep.
No. No. Service . Group AL HS-9 All SG B Aux Fw Ctrl Vlv Xfr Sw 1 AL ZL-9B All SG B Aux Fw Ctrl Vlv Posn 1 AL HK-19B All SG B Aux Fw Ctrl Vlv to Pmp 4 AL HS-10 All SG B Aux Fw Ctrl Vlv Xfr Sw 4 AL ZL-10B All SG B Aux Fw Ctrl Vlv Posn 4 AL HK-llB All SG C Aux Fw Ctrl Vlv Md Pmp A 1 AL HS-11 All SG C Aux Fw Ctrl Vlv Xfr Sw 1 AL ZL-llB All SG C Aux Fw Ctrl Vlv Posn 1 AL HK-12B All SG C Aux Fw Ctrl Vlv to Pmp 4 AL HS-12 All SG C Aux Fw Ctrl Viv Xfr Sw 4 AL ZL-12B All SG C Aux Fw Ctrl Vlv Posn 4 AL FI-1B All SG D Aux Fw Flow 4 AL FI-2B All SG A Aux Fw Flow 1 AL FI-3B All SG B Aux Fw Flow 2 AL FI-4B All SG C Aux Fw Flow 3
- - - - - - _ _ _ . _ _ _ _ _ _ _ ~ - ~ _ - - - - - - - - - - - . . _ _ _
D 5#3Fs57 SNUPPS
. V TABLE 7.4-1 (Sheet 5)
Unit "Sep.
Instrument Group
! . No. No. Service AL PI-15B All Md Aux Fw Pmp B Disch Press NV AL PI-18B All Md Aux Fw Pmp A Disch Press NV AL PI-21B All Turb Driven Aux Fw Pmp Disch NV Press AL PI-25B All Md Aux Fw Pmp A Suct Press 1 Md Aux Fw Pmp B Suct Press 4 AL PI-24B All _
Turb Drive Aux Fw Pmp Suct 2 AL PI-28B All Press Md Aux Fw Pmp B 4 AL HIS-22B All Md Aux Fw Pmp A 1 AL HIS-23B All I
FC-ZL-312AD, All Afpt Trip & Throt Vlv Posn 2 AE, AF FC HIS-312B All Turb Driven Aux Fw Pmp Trip 2 A Throt V1v Aux Fp Turb Speed Gov Ctrl 2 FC HIK-313B All Turb Drvn Aux Fw Pmp Stm 2 AB HIS-5B All
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Isol Viv Turb Drvn Aux Fw Pmp Stm 7 AB HIS-8B All Isol Vlv Cond Stor Tank Level NV AP LI-4B All
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SNUPPS MSib 033I SI TABLE 7.4-1 (Sheet 6)
Instrument Unit Sep.
- No. No. Service Group AL HIS-39B All ESW to Md Aux Fw Pmp B 4 AL HIS-31B All ESW to Md Aux Fw Pmp A 1 AL HIS-32B All ESW to Turb Driven Aux Fw Pmp 1 AL HIS-33B All ESW to Turb Driven Aux Fw Pmp 4 AL HIS-34B All Cst to Md Aux Fw Pmp B 4 AL HIS-35B All Cst to Md Aux Fw Pmp A 1
'sL HIS-36B All Cst to Turb Driven Aux Fw Pmp 1 BB-TI-413X All RCS Cold Leg Temp Loop 1 NV BB-TI-423X All RCS Cold Leg Temp Loop 2 NV BL-TI-433X All RCS Cold Leg Temp Loop 3 NV BB-TI-443X All RCS Cold Leg Temp Loop 4 NV l
SE-NI-31C All Source Range Nuclear Inst NV SE-NI-32C All Source Range Nuclear Inst NV AE-LI-517X All S.G. A Level (narrow range) 4 AE-LI-528X All S.G. B Level (narrow range) 1
- - en
SNUPPS TABLE 7.4-1 (Sheet 7)
V Instrument Unit Sep.
- No. No. Serv #ce
. Group AE-LI-537X All S.G. C Level (narrow range) 4 AE-LI-548X All S.G. D Level (narrow range) 1 i
BG HIS-459B All RCS Letdown to Regen Hx NV BG HIS-460B All RCS Letdown to Regen Hx NV FC-HS-313 All Afpt Gov Ctrl Sel Sw 2 SE-NI-35C All Intermediate Range Nuclear NV Inst SE-NI-36C All Intermediate Range Nuclear NV Inst FC-ZL-315B, All AFPT Gov Vlv Position 2 317B j FC-ZL-312DB All Afpt Throttle Vlv Trip Mech 2 l
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'. SNUPPS
(((dggD 53) If' 7.5 SAFETY-RELATED DISPLAY INSTRUMENTATION The. information necessary to monitor the nuclear steam supply systems, the containment systems, and the balance of plant is displayed on the operator's console and the various control boards located within the control room. These indications include the information to control and operate the unit through all operating conditions, including anticipated operational occurrences and accident and post-accident conditions. Hot shutdown information is also displayed on the auxiliary shutdown control panel located outside the control room (refer to Section 7.4.3). This section is limited to the discussion of those display instruments which provide information to enable the operator to assess reactor status, the onset and severity of accident con'itions, and engineered safety feature system (ESFS) status and performance, or to enable the operator to intelligently perform vital manual actions such as safe shutdown and initiation of manual ESFSs. Reactivity control is monitored by sampling of the reactor coolant for boron.
The surveillance instrumentation, which includen indicators, annunciators, recorders, and lights, consists of specific instrumentation for the following functions:
- a. Reactor trip
- b. Engineered safety features
- c. Safe shutdown This section discusses instrumentation that is required for safety as well as instrumentation that is only indirectly related to safety. The safety-related display instrumentation provided in the control room is listed in Table 7.5-4 and 7.5-5.
This section also furnishes a summary of important display instrumentation provided to monitor system status and per-formance. The bypassed status indication is treated separate-ly to establish a clear definition of the system of bypass indication. The display instrumentation defined for bypass, status, and performance monitoring is not safety related (refer to Table 7.1-2, Sheet 2) since failure in no way degrades the operation of safety systems aad poses no threat to public health and safety.
Refer to Section 1.7 for drawings associated with auxiliary shutdown panel, safety-related display instrumentation, and main control board layouts and ESFS logic diagrams. 7.5-1 Rev. 7 9/81
SNUPPS DY31'sW '
Y TABLE 7.5-4 SAFETY-RELATED DISPLAY INSTRUMENTATION LOCATED ON THE CONTROL BOARD - (NSSS SCOPE OF SUPPLY)
Indicator Notes 1 and 2 Parameter Tag No. PAMS Separation Group I II WIDE RANGE T HOT LEG TI 413A X WIDE RANGE TO HOT LEG TI 423A X WIDE RANGE T COLD LEG TI 413B X WIDE RANGE T COLD LEG TI 423B X PRESSURIZER WATER LEVEL LI 459A X PRESSUh'ZER WATER LEVEL LI 460A X PRESSURI.?ER WATER LEVEL LI 461A X STEMA GEN. LOOP 3 PRESSURE PI 534A X STEAM GEN. LOOP 1 PRESSURE PI 514A X STEAM GEN. LOOP 2 PRESSURE PI 524A X STEAM GEN. LOOP 4 PRESSURE PI 544A X STEAM GEN. LOOP 1 PRESSURE PI 515A X STEAM GEN. LOOP 2 PRESSURE PI 525A X STEAM GEN. LOOP 4 PRESSURE PI 545A X STEAM GEN. LOOP 3 PRESSURE PI 535A X STEAM GEN. LOOP 1 PRESSURE PI 516A X STEAM GEN. LOOP 4 PRESSURE PI 546A X SYEAM GEN. LOOP 2 PRESSURE PI 526A X STEAM GEN. LOOP 3 PRESSURE PI 536A X STEAM GEN. LOOP 2 WATER LEVEL LI 529 X STEAM GEN. LOOP 3 WATER LEVEL LI 539 X STEAM GEN. LOOP 1 WATER LEVEL LI 519 X STEAM GEN. LOOP 4 WATER LEVEL LI 549 X STEAM GEN. LOOP 1 WATER LEVEL LI 518 X STEAM GEN. LOOP 2 WATER LEVEL LI 528 X STEAM GEN. LOOP 3 WATER LEVEL LI 538 X STEAM GEN. LOOP 4 WATER LEVEL LI 548 X STEAM GEN. LOOP 1 WATER LEVEL LI 517 X STEMA GEN. LOOP 2 WATER LEVEL LI 527 X STEAM GEN. LOOP 3 WATER LEVEL LI 537 X STEAM GEN. LOOP 4 WATER LEVEL LI 547 X CONTAINMENT PRESSURE N. R. PI 934 X CONTAINMENT PRESSURE N. R. PI 935 X CONTAINMENT PRESSURE N. R. PI 936 X CONTAINMENT PRESSURE N. R. PI 937 X STEAM GEN. W. R. WATER LEVEL LI 501 X STEAM GEN. '.;. R. WATER LEVEL LI 502 X STEAM GEN. W'.
R. WATER LEVEL LI 503 X STEAM GEN. W. R. WATER LEVEL LI 504 X
SNUPPS l
TABLE 7.5-4 (Sheet 2)
Indcator Notes 1 and 2 Parameter Tag No. PAMS Separation Group
- I II R. C. S. W. R. PRESSURE PI 405 X R. C. S. W. R. PRESSURE PI 403 X BORIC ACID WATER LEVEL LI 102 X R. W. S. T. WATER LEVEL LI 930 X R. W. S. T. WATER LEVEL LI 931 X R. W. S. T. WATER LEVEL LI 932 X R. W. S. T. WATER LEVEL LI 933 X 1 SAFETY INJECTION FLOW FI 917A X SAFETY INJECTION FLOW FI 918B X CONTAINMENT PRFer"EE W. R. PI 938 X CONTAINMENT PRESSURO W. R. PI 939 X R. C. S. EXCESS LETDOWN HEAT TI 137A X EXCHANGER R. C. S. EXCESS LETDOWN HEAT TI 137B X EXCHANGER R. C. S. EXCESS LETDOWN HEAT TI 138A X EXCHANGER R. C. S. EXCESS LETDOWN HEAT TI 138B X EXCHANGER BORIC ACID WATER LEVEL LI 104 X BORIC ACID WATER LEVEL LI 105 X BORIC ACID WATER LEVEL LI 106 X l
NOTES:
1
- 2. See Westinghouse process control block diagrams for the l applicable protection set.
1
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SNUPPS TABLE 7.5-5 9 SAFETY-RELATED DISPLAY INSTRUMENTATION LOCATED ON THE CONTROL BOARD - (BOP SCOPE OF SUPPLY)
Indicator Separation Parameter Tag No. Group AUXILIARY FEEDNATER-FLOW AL-FI-1A 4 AUXILIARY FEEDWATER-FLOW AL-FI-2A 1 AUXILIARY FEEDWATER-FLOW AL-FI-3A 2 AUXILIARY FEEDWATER-FLOW AL-FI-4A 3 CONDENSATE STORAGE TANK-PRESSURE AL-PI-37 1 CONDENSATE STORAGE TANK-PRESSURE AL-PI-38 2 CONDENSATE STORAGE TANK-PRESSURE AL-PI-39 4 TURBINE DRIVEN AUXILIARY FEED PUMP-SUCTION PRESS. AL-PI-26A 2 MOTOR DRIVEN AUXILIARY FEED PUMP A-SUCTION PRESS. AL-PI-25A 1 MOTOR DRIVEN AUXILIARY FEED PUMP B-SUCTION FRESS. AL-PI-24A 4 CONTROL ROOM AIR INTAKE-CHLORINE GK-AI-2 4 CONTROL ROOM AIR INTAKE-CHLORINE GK-AI-3 1 CONTROL ROOM AIR INTAKE-GASEOUS RADIOACTIVITY GK-RI-4* 4 CONTROL ROOM AIR INTAKE-GASEOUS RADIOACTIVITY GK-RI-5* 1
! CONTAINMENT-GASEOUS RADIOACTIVITY GT-RI-31* 4 f CONTAINMENT-GASEOUS RADIOACTIVITY GT-RI-32* 1 l CONTAINMENT-HYDROGEN GS-AI-10 4 I CONTAINMENT-HYDROGEN GS-AI- 19 1 CONTAINMENT SUMP / CONTAINMENT LEVEL LF-LI-10 4 CONTAINMENT SUMP / CONTAINMENT LEVEL LF-LI-9 1 CONTAINMENT PURGE-GASEOUS RADIO- GT-RI-33* 4 ACTIVITY CONTAINMENT PURGE-GASEOUS RADIO- GT-RI-22* 1 ACTIVITY CONTAINMENT SPRAY ADDITIVE TANK-LEVEL EN-LI-17 4 l CONTAINMENT SPRAY ADDITIVE TANK-LEVEL EN-LI-19 1 l FUEL BUILDING-GASEOUS RADIOACTIVITY GG-RI-28* 4 l
FUEL BUILDING-GASEOUS RADIOACTIVITY GG-RI-27* 1 CONTAINMENT-AIR TEMPERATURE GN-TI-61 4 CONTAINMENT-AIR TEMPERATURE GN-TI-60 1 CONTAINMENT-AIR TEMPERATURE GN-TI-63 4 CONTAINMENT-AIR TEMPERATURE GN-TI-62 1 CONTAINMENT-POST-ACCIDENT RADIATION GT-RI-60 4 l CONTAINMENT-POST-ACCIDENT RADIATION GT-RI-59 1 CONTROL BUILDING SUMP-LEVEL LF-LI-125 4 CONTROL BUILDING SUMP-LEVEL LF-LI-124~ 1 DIESEL GENERATOR BUILDING SUMP-LEVEL LF-LI-106 4 DIESEL GENERATOR BUILDING SUMP-LEVEL LF-LI-105 1 1 -
___ MS@ #35,5,5 V SNUPPS TABLE 7.5-5 (Sheet 2)
Indicator Separation Parameter Tag No. Group RHR PUMP ROOM SUMP-LEVEL LF-LI-101 4 RHR PUMP ROOM SUMP-LEVEL LF-L7-102 1 AUXILIARY BUILDING SUMP-LEVEL LF-L't-104 4 AUXILIARY BUILDING SUMP-LEVEL LF-LI-103 1 i
a l
- Digital display on radiation monitoring panel SP-067.
I
,. g
(. -
. SNUPPS is at a temperature below the reference nil ductility temperature
! (RNDT). The monitored system temperature signals are processed l
to generate the reference pressure limit program which is compared to the actual monitored RCS pressure. This comparison provides an actuation signal to an actuation device which will cause the PORV to automatically open, if necessary, to prevent
- pressure conditions from exceeding allowable limits. Refer to Figure 7.6-4 for the block diagram showing the interlocks for j
RCS pressure control during low temperature operation.
, a pressure and kamennhwt1 l IK;r'sMwvevtm E '_.az:)Q - the generating stationAvariables e j required for this interlock are channelized as follows:
r 1
- a. Pr tection Set I
(
! 1) Wid range RCS t erature from ~ t legs D b i
4 ce (2) W'd range RC p:tess e( 405) wi%
i
. Tnt.ed b. Prot ction S I
@ ) Wide r e 5 temp ature fr m co /leg I
- c. Pr tection et IV ,
L (1) Wide range RCS pressure (PT 403) i .
.D o protec lon pnej sejts wlce I a ge tepperatur iT, continuo signaAs, monito Anputs S temp aturt co ibon- l pne<Wo lehcA:?c en tx.1,sp ,.a u _x yp e,sp.u;y c v. cw ur Sr .
/ In - tectior t I, .e e tin RCS no .eg wide 7ge ~
tem ture harhq1s ill su contin u analog np t Wf ?Wcxe , -o an auc eerin devi rpas o n puw17 / // 3y A y--Lh Mr= *. + -n- m 2- w s.
acs of c ccyQe releded The lowest reading is selected d input to a function generator which calculates the reference pressure limit program, considering the plant's allow le pressure a 1ntemperature limits. Also available frpm protection set Eris the wide ,
range RCS pressure signal. > w ex ve -XEW ~ _
&rrz.s-W m_-g y ,_ _l The reference pressure from -
the function generator is ccmpared to the actual RCS pressure monitored by the wide range pressure channel. The error signal derived from the difference between the reference pressure and the. actual measured pressure will first annunciate i
a main control board alarm whenever the actual measured pres-sure approaches, within a predetermined amount, the reference pressure. On a further increase in measu_r_ed pressure, the error siglal will generate an W y M M actuation sienal.
The crua on gnaA a 1 . a.c A l 6 _
C .e t .ill ntrol ' .?"q:ma pne everpyc)ess a em cgr.ara prors-l e ,w v . .sA ., .. , .. A,p ce)s n. % ,,p,4 s j rm' =~~ l ~
- C.e ' R p: . ..# r./. .; .f. . . ,.tem: 'ture c enc t '3 t
l [, - - - - y '#.e . _i .._... _: .
- . s_ -rksk , ie:dation'
! 7 6-5
l*, .
\ .TMERT(R) (-!c page 74-s) 1 nse wa ro" r unnsmeM o' ens me^un kJ salprou l .
org gue gne{$d,,esmg .
N o/$1[nYis dedio
/he auchaneeni low fempem/are. Fran -/he hts a>de range
/cmpero7are chonneis fa//s auMm ,'he in of co/d owspessure opdahilily /here o/erfmg 49e opemfor -k arm He, Ras co overprmue mihgahon syslem onwh au/omahco/ly cpens lhe .bicek va/ve, uden ne block valve con /ro/ swikh As m Hie aukmo.ha. pasihon.
.TNSERT & (-lo page ?;6-S)
- a. PVessure and Tempenlare Inpuls lo mv4ssn -
(d four wde ronge Res femperahare -
signals clerimd Acm channok m o 7r'om A relaled profec/ ton seh -
(.?) One unde rarge Respressure asissi.
derwed frcm a channel in a Twin .
~
A refalelprolechon seh .
- b. Pressure and 7e~mperofure .Tnpuls k MV<sM .
(d /~ cur wide ran.ge Res -fempera/ure signals cienved from channels m a 17am 8 telaled pro /echon sef; (2.) Ce wde range Ms pasure signa./
denved from a shpanel m o. hm B re/oled prol"chon seh O
6
---,-,,-,n-., -
, - - - ,,,,--,,,__---.--,,-,_,,,,-.-n_n,- -
_..,__-n-, + - - - - , _ - - - . - - , - , - , - , - - - - . . , , - - - ,-c
. . SNUPPS a zusser @ 9 e_
si t.al t
- mper gur rea+ than .e rar e of once . 1 7aC 's i13 , riv . t ,ec . s tu 1 e at mal i _ _
op tin on . .ons ar uit o a fa. re 1. e o iproce s%r ors.1 f
The monitored generating station variables that generate the actuation _ signal for the E PORV are processed in a similar manner. Jn t::e case of FORV "B," ,the reference te=pera re i
' gen razea FAM&MvWxxWMrom w' e range
- e owest a 'onee e} =
l ld leg emp "ature, the uc neer' g dev ce t riving 'ts 4nput rom t. RC wide tempe t in pr *ectd n se II and the e aal measu ed pressu. signal '
avai = e from otection se IV. The e. re, the ant var able sed for V "B" ar d -ived om a rote i set t
( is .de enden of .e set frem whi . plant ar' ables us fo
-- NEY--
YS&ffhW$f$ N e Upon receipt of the actuation signal, the actuation device l will automatically cause the PORV to open. Upon sufficient -
RCS inventory letdown, the operating RCS pressure will decrease, clearing the actuation signal. Removal of this signal causes the PORV to close.
7.6.6.1 Analvsis of_I_nterlocks Many criteria presented in IEEE Standards 279 ~1971 and 338-1971 do not apply to the inteirlocks for RCS pressure control during ~
low temperature operation, because the interlocks do not perform a protective function but, rather, provide automatic j pressure control at low temperatures as a back-up to the~
j operator. However, although IEEE Standard 279-1971 criteria do not apply, some advantages of the dependability and. benefits l of an IEEE Standard 279-1971 design have acened by including sc1ceted elements, as noted above, in the protection sets and by organizing the control of the two PORVs (either of which can accomplish the RCS press _ure control function) into dual channels,In. w x' A W 2 A - __-
The design of the low temperature interlocks for RCS pressure l control is such that pertinent features include:
~
! a. No credible failure at the output of the protaction I set racks, after the output leaves the racks o interface with the interlocks, will prevent the associated protection system channel from performing -
its protective function because p .. ,
., ns_ s sa _. .
^
N. -- :. - e -
A of MC fxpss of Tlaxos 8 7Ye/n
/nkrbCh5ffMt';estingcapabilityforel(emena A see yrg ~,76-4).
- b. . .ne inter.i.ocks ai-hin (nct e:tternal tc) the pr:tection system is censistent with ohe tes ing principles and methods 1
7.6-6 i .-- _ -..- -_ - --_. . - - - _. _- . _ -
. -.. . . . = = - .. = - - . - - -.
e, . .
2NSEe7 @ (lopage 76-6) 1 is is cho provided -lo close Me block va ve au/cinahcally d he relief volve
~
i
( fads or shc.ks m Me opert poyhon l followmb Me Re some pressure planf drcpsfrawent,bs/o +and he resel pressure fcr +he rehef Volve. -
N&
is above Me cold overpressura-
} ion se/pomf, d
"- .m o . m. o .o q& . . .. a.. .
, . = = . . . - . . ,
~ '
- SNUPPS
.,%e blek valve e armed
~
^
[Q discussed in Section 7.2.2.2.3, item J. I should be noted that there is an annunciator which p ovides an alarm whe w W "-c"'w w -Ae w2 m
. w r e w coincident with a closed po/ition of the motor-operated (Mov) pressurizer relief 4 valve. This MOV is in the same fluid path as the PORV, with a separate MOV and alarm used with the secend PORV.
- c. A loss of offsite power will not defeat the pro-visions for an electrical power source for the inter-
! locks because these provisions are through onsite power, which is described in Section 8.3.
7.6.7 ISOLATION OF ESSENTIAL SERVICE WATER (ESW) TO THE AIR COMPRESSORS ,
l 1
7.6.7.1 Description As stated in.Section 9.2.1.2.2.1, ESW flow to the nonsafety-related air compressors and associated aftercoolers is maintained following a DBA.- Instnu:entation and controls are provided to automatically isolate each train of the ESW to the air com-pressors on high flow. ESW to the air compressors can also be isolated by remote manual means. .
Each control system (one per train of the ESW) utilizes a ,
differential pressure transmitter and bistable which senses flow through the associated isolation valve. On high flow (indicative of gross leakage in the nonseismic portion of the '
system), the control system automatically closes the isolation valve.
The isolation valve will remain in the closed position until the valve is manually reset by the operator in the control ,,
l room.
A means of remote manual isolation is provided in the control room. The status of each isolation valve is indicated by open and closed indicating lights in the control room.
The isolation valves are air operated and are designed to fail closed on the loss of air and electrical power.
1
- a. Initiating circuits Each isolation valve is automatically actuated by flow monitoring instrumentation. The isolation valves can also be closed via control switches in the control room.
- b. Logic .
I The logic diagram for the isolation of the ESW to the ~
l air compressors is provided in Section 1.7.
7.6-7
- ~ - - - -
. SNUPPS
- c. Conformance to other criteria and standards i conformance to other criteria and standards is indi-cated in Table 7.1-2.
7.6.9 FIRE PROTECTION AND pETECTION Fire protection and detection is discussed in Section 9.5.1.
NOT
^
l l
l .
t l
l .
t .
t S
1 S
7.6-12
.1Nse.er @ (-lo page 7t:,-/z) /of s 7 4:,./0 INTER 2.XXS R;R PRESSUR/ZER PRESSURE REDEC4YSTE191 l 74,. /6.I Desenohon ef Ressunzer Ressure l Ne' lief' Evslem The pmssunzer pressure relief (PPR) sys/em provides lhe followig: _
- a. dophibly for Res overpressure mihaahon du cc/d .shu/down, healap, and'coddown froin lo mmimize
-lhe polenhol r impairmg teoc/cr '
vessel integrily a: hen cperahng af '
er near 4he vessel ducYi/>ly limi/s-
- b. ohi/>ly for RGS depressunzalron 1
~
d,op//cwmg fi Ccndihon X,227,and2Eereirl .
- c. 3ki/cck lint a>r/h /he pressunzer -
FCRY.s and FCRY b/cc/c yolves .m aulo cenirol, c/csss the FcRY b/cck volves ord preveh .sionaI.1 ficm the pressunzer pressu"re ccnhol Syslem frcm cpsning .}he FORVs amen pressunzer penie n low.
76.102 Ak/tef,3vshm Desarrohan nf Ressurtzer Pressue Tnk>rlocks
.T4erlocks for }Ae PPR ,syslem con}rol }he.
openmg and closmg of Jhe pnessurizer PORYs ond she PORY block w/ves. _ These
.2NSERT@ (aambnued.) 2of6 In/erlocks prowcle -h9e following Emc/rens :
- a. Flessunzer peore confroI (refer
/c Sechon 7.7.1.s for a desanphon). -
- b. Res pressure conJrol darm low
-lem Seck,oera}ure operahon (r h ns s2.2 and 74.fa r s.
desanph6n).
- c. Res pwre ccn/ro/ lo ochieve and mamfom a cold skuldocon and Jo heo/ap asmg e ' menf lhal is required for (refer h> Appendix 64A for o iphon).
7he ronde ness .
urizer mler/cck pressurefanchons cotiliol arehtherived kom .
promxs xaramelers os shown en are
- 7. 2-1 sheef ll ond -Hie infer /cil se
~-
fur.ch,ons crecess para inpu/s required as well for as/dw kmp& erahre cpera- 4 hon as ,thown on Rqure 7; -
. 7he ~
funchons shown on F1
-/I:cre needed for }he PG'gure 7:&-4volves includeas*Ry b/cck well as Me pressunzer PCR's / lo neef boM inier/cck legie end inanual cperahon reoairemenh where monaal opera. hon is trisw M Me mam. ccnfrol beord. f ggwv2aneworsweete>#Fw #nd l
l l
1
. E 13.557-39
$ dm t#
_5 E .- 8 E EE a
2_-----------
.R
~5 g kB
- ,[ *- .-
e t .
=
u5a-
'~
_ g, "E .
- k - LA- W a_ w g
= -2 x , as! I s
- 3 : s ';E.
M
!E 5 *
=
E " ON-
_ bwE s E l -
g .g 3. w :3
=r
=
, "g .
A \ 6 'S
- E lw E 2= s e.s g f- Sh
- e= ,
5 w EW=
/
xF- \- b
=
u-
/ -
/
5_ E ..: Ej . .
s E QE C7 5
/ [q J l
=5 s .w
'l s' ?- '
- ,n_ Idl wa
" w* v3 -
- g - c$* 3 3 .s *E =
- N *
- 8-i::
s 555 \f! 5
~
E 3 . ;-E' + =E3 y. :R
\ y
g,'" ,. s 3-gg ?"
3 W
=
I g
l 1
/X' =y _ 's
.M *: *
- 3;
-- g i
5 *
= 5.
3
!==
ma= E_= 2-
?
=5"_E E o
EW
==
. SNUPPS
. FIGURE'7.6-4 DIAGR;J4 SHCWING GENERAT!NG PLANT VARIABLE PROCESSING FOR LOW TEMPERATURE INTERLOCK 5 FOR RCS PRESSURE CONTROL s_ -
em .D o _ _ .q;P * - - . _ . , - . - --
19,SE12 PnE ssuns2En wiot nANGE nCs wiOE n ANot PaEssunt2En PatssURE TE RAPE n ATuRE nCsPatstumE PREssung nEL3E7 PCV essa (NOTES 2 & se (seOTE 40 REufF a- ----m- **
.NTEnLocs .SeGasA,L, OTE CO8sinOL.swtTCH
,ON . CS, ________ __
,=i",'s*.. ,=;"/, ,. IOP<N I AUTalc==l , '
SHEET 3D SMEET 111 I AUCTIO8 low lsten l ll b .Ru "1
{j A inasse O CDLD i,,,) 9- (NOTE 31 OVEmPatstung
") 'T [ eseNnATaose ACTnaATsoed
() EON asCes INOTE 3) l l Ants lSLOC SLOCE WALVE 30009 PG PS M ese A CoasTnOLswtTCH -
(MTE si tOae escen A
l OPEN l AUTO lCLOSEl fesOTE 33 NOTE S:
U
- 1. THis stGalAL es THE OUTPUT FROne BesTABLE PS eseE ELECiniCAL ISOLATIO8e Is nEQuenED TO SasesO THis slGNAL 188T0 feet inAlee S l PnOTECTIOes CAtletET.
- 3. wios n AssoE aCs TEMPEn Aruns s4GNALs OEnsvfo enOes CnAssanELs sus Ainaus S mELATED PnOTECTIOce SET.
3 ANNumCIATeON IN MAase C09sinGL nOOes is OPES 8 ' CLOsE l mEOvenEO visible TO OPEnATOn AT MAsas CoasYnot SOAno.
SLOCK WALvt BeOT FULL OPE RI l 4. wios nAnsof nCs PnEssunE seGNAL DEnlVED "
7 ROM CHAmeNEL IN AinAsse 8 nELATED '"O 88
> PnOTECTiON SET.
- 6. status LIGHTS Musi SE PnOVIDED FOn E ACM POnv ANo E ACw POnv SLOCu vatvE AT THE M A4N CONTROL SOARD 10 seeDeCATE wMEN THE WALvE is FULLY CLOSED On FULLY OPEN,
- s. TME nC3 LOOP A8eD HOT LEG CR COLD gesOTE 33 ES$UPPS LEG AsstGaseet NTs FOn THE miDE nANot nCs TEasPEnATunt sicNAts F'EEU" 7 $4" Musi SE COWsisTEseT wif te THE yr g pen,ci a o,atrem Shousing nEOuenEnsENis POn avus ANo PAMs Leys T , . . _. _: for Presseariamr Presswo now e svatun
. snese 1 e
18.9411 PRESSuRutR PRESSURGER PRESSURE WutOE RANGE ACS wtOE RARIGE PRES $uRE RE LIE F TE MPE R ATURE ACS PRESSURE RELIE F SsONAL PCV - 444 A (NOTES 3 & SI (seOTE 43 tesTERLOCK (NOTE Tl CONTRCL SWITCH - - - - - - - - --
gnEFER TO (REFER TO FeOURE 7 44, FIGURE T 21 lOPEN l AUTO lCLOSEl SHEET 33 SMEET 11) $< , 4, g l AUCTIO**EER low A TRAsas A RCS COLO
~ r #III OVERPRESSURE
' T' h TI . MITIGATION ACTUATIOes
/
lesOTE 31 SON esCes OPEN,. C O.E -
l ARM i-OC.i KOCK VALVE 8000A [Ps Pg PCV 446A CONTROL SWutTCM %. 7 (NOTE Si (ON MCet A
lOPEN] AUTO lCLOSEl INOTE 3)
[ feOTES:
e
! 1. THis SIGN AL is THE OUTPUT FROM SISTABLE PS 466E ELECTRIC AL ISOLATION IS REOutRED - .i l TO BRING THet SaGNAL INTO THE TRAsas A '.
PROTECTION CAseseET.
- 3. WtOE R ANGE RCS TEasPERATURE SIGNALS DERIVED FROM CHAmasELS IN A TRAlfe P i il e RELATED PROTECTION SET.
- 3. AN8euNCIATIO8s IN MAIN COssTROL ROOM 18 REOUIRED VISISLE TO OPERATOR Af hsAIN gpggg , gLggg CONTROL SOARD.
- 4. WlOE R ANGE ACS PRESSURE StGNAL DERIVED FROM CHAN8sEL IIe A TRAles A RELATED SLOCK VALVE 800T FULL OPEN PROTECTION SET. 80004
$. STATUS LIOMTS MUST SE PROVtOED FOR E ACH PORV AfeO E ACM PORV SLOCK VALVE ,
AT THE MAIN COneTROL SOARD TO teeDeCATE WHEN THE V ALVE IS FULLY CLOSED 04 FutLY OPEN. g yps
[
S. THE RCSLOOP AND teOT LEG 04 COLD A LEG AlleGeeMENTS FOR THE WIDE Fagwe 7.$4.
8 ,, a3 RANGE RCS TEMPERATURE SaGauALS Treen A Fesacteenal D symm Siemeng asuST DE CONSISTENT ntTM THE
! REOueREastsetS FOR RVuS Aaso PAMS. Logic Requweenants for Freamsresar
' reummse Reeser SysessR S:
j
. e i
l __ ___ ___ _ __ _ _ _ __
~~
. ~
~
~ ~ -
. .. = . : ~~ .. ... ._
19136 1 PRESSURIZER LOW PRESSURE LEAD / LAG COMPENSATED PB P8 PB PS 455H } 456H { 457N { 458H {
& 0- -
O- 0- -
a 2/4 A ,
i ,
NOTES: 3r
- 1. FOR NOTATION AND DRAWING .
CONVENTION, REFER TO l FIGURE 7.21, SHEET 1 1 2. THIS LOGIC IS REDllNDANT 1 r
- t oJRIZER PRESSURE REL.EF INTERLOCK -
(REFER TO FIGURE 7.6-4, SHEETS 1 AND 2)
SNUPPS Figurs 7.6 4 Functional Diagram of Logic .
Requirement for Pressurizer Pressure Relief System Interlock Sheet 3
O
- o. -I,! T S r 3 l'8 8e1 ,d g8 i
= -
5 1 1 -
p-_ .
F/-i gis,g _
k 25 E I l f-
' i 8 *
> rs 8 lniI !
i I
1] -
l ,
h h i jl,-e 1.1].
g- ,, -
-l=>p 9 ll
--- qhpF- T 8l M ji!
=
- i
! !l,i!- Eg
,, s --
ll . E N ~'~ ~s *0 2lc I -------j ' l Pj ! I
_____.g' FC 'llih.lj-i .g , , l iy I u i -
.___?._.j.I .I ! !' r- ;' d 4- "I'l!!
~] , l-ENfl i
l
-!!!!! =
I l
i_
ji
_k_________.(( .
l[jj .-l3l Aw__.
Y
.t_u 1 i 8 ...J
's' l' ill j-- S !!!g! ------ lii!!
I I
lI!l------- jj;fi h-S .
p_________________________.jjj!j
- i. .
'I85!!
l IP __ I4$Hl ril ! ,
- .; .ne, b-b a$ -
Ib
~
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l 42 4 l " ylll
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- l Ill j.i!l !
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'2 !!!! !
i
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.illriy;ill!
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s s.:. . u . u
( . _ _ _ _ . . - _
4 RESPONSE 7D ICSB A6ENDA /TE/H 9* /3 (AND NUREG-0737., /7Em EK.S.lo) c The NRC has raised the question of whether the pressurizer power operated relief valves would be actuated for a turbine trip without reactor trip .
below a power level of 50% (P-9 setpoint). An analysis has been perforrted using realistic yet conservative values for the core physics parameters (primarily reactivity feedback coefficients and control rod worths), and a conservatively hl h Initial power, average reactor temperature (T g and pressurizer press re level to account for instrument inaccuracles.yg),*
The transient was initiated from the setpoint for the P-9 Interlock, namely 50% of the reactor full power level plus 2% for power measurement uncer-tainty. This is a conservative starting point, and-would bracket all trans-lents initiated from a lower power level. The core physics parameters used were the ones that would result in the most positive reactivity feedbacks (i.e. highest power levels). The steam dump valves were assumed to be act-uated by the load rejection controller.
Based upon the results from the analysis, the peak pressure reached in the pressurizer would be 2302 psia. The setpoint for the actuation of the pres-surizer power operated relief valves is 2350 psia. Even including the +20 psi pressure measurement uncertainty, there is still a margin of 28 psi ,
between the peak pressur.e reached and the minimum activation pressure for the pressurizer power operated relief valves. .
.3 e
e O
e I
gZ' e
t l- _
. . INSEET A An addihons/ onolysis has been perkrmed do defermine. Me enseyuenes ('specificaHy' Me hkelihwd af Me pmssunzer power-opgraled relie/ volves cpenne) of hewns a larhine
+n,o.due /t> c loss of condenser vacuum.
above.
The =ajor difference between thisM** analysis and the one presentedE2:1 -
VEN is that now the nor=al steam du=p sys:em is unavailable, and the stes= relief =us: be done through the at=osphetic relief valves. Since there is a longer delay ti=e before the atsospheric reliefs reach their se: point (in ce=parison to the nor=al steam du=p syste=) and their capacity is about one-half of the seca:: c' ,p sy s t e= , there is an increased likelihood : hat the pressuri:er PORVs will open.
Figure 1 shows the plant operating ranges for which the pressurizer PORVs will Open for a turbine trip due to a loss of condenser signal. Above 50%
pcwer, a turbine trip will cause a reactor trip (due to F-9 setpoint), and
- he pressuri:er PORY setpoint will not be reached. Below a power level of .
35%-a0% (depending on fuel burnup), the pressurizer spray rate is adequate to =aintain the pressuri=er pressure below the setpoint. Therefore, only -
in the narrow band between about 35% and 50% power will the pressuriser PORVs opsn for a loss of condenser. -
3ased upcn :he opercting histe.7 of current plants, the chancer of g'et:ing a cog. denser unavailable signal (and hence a turbine trip) is about 15e out of 10 cperating hours. D hS E S Assu=ing 98% plant availabili:v and a -
'0-year plant liferine, this works out to about 4 condenser unavailable turbine trips occurring duri.s :he no- Al life of a plant. Assuming an equal chance or having the plant operste an .rhere berveen 0% snd 100% power (an unrealis:i:
value. since they usuelly epersen either r.: a full a- no load levell the chances of having a :endenser unavailable signal generate a transient which would rosul: in the opening of the pressurizer ?ORVs is less than one per plant lifeti=e. -
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ENCLOSURE C TO SLNRC 81-82 LEVEL MEASUREMENT ERRORS RESULTING FROM POST-ACCIDENT ENVIRONMENTAL EFFECTS ON LEVEL INSTRUMENT REFERENCE LEGS F
Ic System Descriptions The following liquid level measuring systems inside containment are used to initiate safety actions.
- a. Steam Generator Narrow Range Water Level The steam generator narrow range water level detection system con-sists of four differential pressure measurement channels per steam generator, each with an open column reference leg, a condensing pot to ensure that the reference leg maintains a constant level, and a pressure transmitter. The upper taps are located at a reference elevation of 581 in. (approximate distance above tube sheet) and the lower taps at a reference elevation of 453.25 in. Tne vertical dis-tance between taps is 127.75 in. The top of the tube bundle is approximately 90 in. below the lower taps. The feedring is approxi-mately 22 in, above the lower taps. The differential pressure trans-mitters and the level indicators are Class IE and are qualified both to the post-accident environment and an SSE. I The SG narrow range water level is used for the following safety functions:
- Turbine trip and feedwater isolation on high-high SG water level.
- Reactor trip on low-low SG water level.
- Auxiliary feedwater initiation on low-low SG water level.
- Post-accident monitoring. In Emergency Operating Procedures narrow range water level is the basis for manual control of auxiliary feedwater flow to intact steam generators for pipe breaks inside containment. It is also an alternate basis (to auxiliary feedwater flow) for termination of safety injection,
- b. Steam Generator Wide Range Water Level Each steam generator has one wide-range differential pressure mea- i surement channel consisting of an open column reference leg with a
Page Two condensing pot, to ensure that the leg maintains a constant level, and a pressure transmitter. Tne upper tap is located at a reference elevation of 587 in. and the lower tap at a reference elevation of 22 in. The vertical distance between taps is 559 in. The differ-ential pressure transmitters and the level indicators are Class IE and are qualified for both the post-accident environment and an SSE.
.The SG wide range water level is used for the following safety functions:
- Post-accident monitoring. SG 1evel indication for pipe breaks outside containment. For pipe breaks inside containment, it has no specific function in Emergenci Operating Procedures.
- c. Pressurizer Water Level The pressurizer has three channels of differential oressure measure-ment, each with a differential pressure tran mitter and an open column reference leg with a condensing pot. The span is from the bottom of the pressurizer to the top of the straight shell. The differential pressure transmitters and the level-indicators are Class IE and are qualified for the post-accident environment or an SSE.
The pressurizer water level is used for the following safety func-tions:
1
- Reactor trip on high water level.
- Post-accident monitoring. Tennination or throttling of safety injection flow.
l l
- 2. Safety Function Setpoints
- a. SG Narrow Range Water Level TPe programmed operating water level in the steam generators is at 50% of the narrow range span at all power levels. This water level is maintained by an automatic control system that functions frem 0 to 100% power, whenever at least one of the feedwater pumps (either
Page three the small motor en pump or one or two large turbine-driven pumps) is operating. The control system is capable of maintaining level within 5% of the programmed level, that is, within 45 to 55%
of narrow range span.
Neglecting, for the monent, the effect of post-accident conditions on the SG reference legs, the high '.iigh setpoint would be at 78.1%
of narrow range and the low-low setpoint at 17.2% of narrow rance.
The potential errors in these setpoints are 113 .2%, which cor,ists of a 110% error allowance for post-accident effects (radiation and temperature) on the transmitter and 1.2% 3 statistical combination of other errors. Thus the possible lower limit of actual low-low k
trips would be 4.0% of narrow range span. Similarly the upper limit 3 of high-high trips would be 91.3% of narrow range span. This would result in the following margins:
Low-low level trip 1 4% above the bottom of the narrow range span.
Low-low level trip 124.6% below extreme expected swings in operat-ing level.
high-high level trip > 19.9% above extreme expected swings in operat-ing level.
- b. Pressurizer Water Level The programmed operating water level in the pressurizer varies from 25 to 60% of span as the power level varies from 0 to 100%. This water level is maintained by an automatic control system. The set-point for the high water level trip is at 92% of span.
- 3. Effect of Post-Accident Cor.ditions on Reference Leg Level Indication Systems
- a. Reference Leg Heatup High energy line breaks inside containment can result in heatup of level measurement reference legs. Increased reference leg water column temperature results in a decrease of the water column density
Page four with a consequent apparent increase in the indicated water level (i.e., apparent level exceeding actual level).
The following formula can be used to calculate the magnitude of this bias:
I L* cal -PL1 E=
H (Pf , cal - Pg, cal)
Where E = level error due to reference leg heatup, as a fraction of level span H = level span = vertical distance _
between pressure taps H
L
= height of reference leg (maxi-mum vertical distance from lower tap to water level in condensing pot on upper tap) 9 L, cal = water density at containment temperature and steam genera-tor or pressurizer pressure for which the level indication system was calibrated
& L
= water density in reference leg at the time of interest (Pf , cal- Pg , cal) = difference between saturated water density and dry -satu-rated steam density a+ e
- - ~ - . - --
Page fi e steam generator or pressurizer pressure for which the level indication system was calibra-ted.
This procedure is based on the assumption that the tubing from the upper and lower taps, below the elevation of the lower tap, have the same temperature at all times.
Figure 1 shows the level bias as a function of reference leg tempera-ture, assuming HL/H = 1.1 and the calibration conditions of: con-tainment temperature = 90 F, steam generator pressure = 1000 psia.
These are conservative values applicable to the SNUPPS plants.
- b. Reference Leg Boiling i
In addition to reference leg density change under subcooled condi-tions, boiling could conceivably occur in the reference leg following depressurization of any steam generator with high-containment tempera-l
- ture. This combination of conditions could occur only af ter a steam-l line or feedline rupture inside containment. If such boiling were to occur, it could cause a major bias in the indicated level for a short l time period, in the extreme case indicating 100 percent level when the vessel is actually empty.
Containment analyses performed by Westinghouse indicate that such
' soiling would not occur.
l l
- c. Coolant Density Changes A bias in indicated water level may also be introduced by changes in pressurizer or steam generator pressure, due to changes in the den-sity of the saturated water ano steam within those vessels. While prediction of the effects of rapid depressurization requires complex calculations for each specific case, the bias which would exist at
! low power under quiescent conditions can be calculated directly, i
l using the following formula.
l l
4 + emm me.=+ha we-ew mh- wm -w twe, +ie,,.*w-Page six l
E= I L, cal -il - f gcal + p g) .
(f , cal - f cal) f g L (Tf-fg) 2 H -
L (f f, cal - f gcal) H 4
Where E = level error due to density i
changes in both the vessel
]
and the reference leg, as a fraction of level span L = true water level in the vessel, above the lower
' level tap
' = saturated water density at ff the pressure of interest
! pg
= dry saturated steam density at the pressure of interest, and other symbols have the same meaning as in Section 2A.
For an example, Figure 2 shows the true water level as a function of steam generator pressure and indicated level, assuming the fol-lowing calibration conditions: Containment temperature = 90 F, steam generator pressure = 1000 psia, and the reference leg is at 90 F. Figure 3 is similarly calculated for a pressurizer, with the assumptions noted on the Figure.
Page seven
- 4. Description and Evaluation of Planned Modifications to Water Level Measurement Systems
_ a. Steam Generator Narrow Range Water Level Trips The low-low setpoint will be raised 11% of narrow range span to compensate for the effects of reference leg heatup. The bases for
~
this change are analyses specific to the SNUPPS plants.
Low-low steam generator water level is the primary means of tripping the reactor for a feedwater line break. This trip is backed-up by the high-1 containment pressure trip, which trips the reactor by initiating a safety injection signal. The primary means of tripping the reactor for a steam-line break is low steam-line pressure.
Diverse (back-up) trips for a steam-line break are hgh steam-line negative pressure rate, and the high-1 containmer.c pressure trip.
Low-low steam generator water level may under certain circumstances also be a diverse trip for a steam-line break, but an increase in feedwater flow could prevent this trip from occuring in the event of a steam-line break.
Calculations have been performed for a wide spectrum of feedwater line breaks inside containment. For a feedwater line break postula-ted upstream of the feedline check valve in a feedwater line, feed-U water supplied by the feed train, at a maximum temperature of 420 F, is released to containment. At the time that the actual water level in the affected SG reaches 4% of narrow range span (conservatively assumed to correspond to the low-low setpoint) the containment temperature is less than 150 F for any break size or operating power level.
For a feedwater line break postulated between the feedline check valve and the steam generator at full power, initially a 50-50 mix-ture of 420 F feedwater and saturated water at 1000 psi (545 F) l blows down to containment. When the feedring becomes uncovered, saturated steam from the affected steam generator blows down to containment. At the time. the actual water ler in the affected l
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Page eight 4
steam generator reaches 4% of narrow range span (assumed to cor-respond to the trip setpoint, as above) the containment tempera-ture is less than 200 F for any break size.
For a feedwater line break postulated between the feedline check valve and the steam generator at zero power, initially saturated water at 1100 psia (556 F) blows down to containment. When the feedring becomes uncovered, saturated steam from all four steam generators blows down to containment. At the time the actual water level in the affected steam generator reaches 4% of narrow range span the containment temperature is 292 F.
The reference leg heatup error for 292 F containment temperature i, 11% of narrow level span from Figure 1.
For a steam-line break inside containment, low-low SG is only a weakly diverse trip as previously discussed. The nominal set-point for the high-1 containment pressure trip will be 3.1 psig and the potential error is 11 .9 psi. Assuming release of 1000 psia saturated steam to containment, the containment temperature at 5 psig is 200 F.
Raising the steam generator narrow range low-low setpoint by 11%
of narrow range span will preserve the margin of actual water level >_
4% above bottom of narrow range span at the time of low-low level trip for any feedwater line break and although the margin between the trip setpoint and operating water level will be reduced, the discussion of Section 2.a, above, shows that spurious trips are unlikely to occur.
The high-high SG water level trip is not required for accident situa-tions that could cause significant errors in level indication. The setpoint of this trip will remain unchanged.
Page nh.
- b. Pressurizer Water Level Trip The pressurizer level trip is not required for accident situations that could cause significant errors in level indication. The set-
- point of this trip will remain unchanged.
- c. Post-Accident Monitoring - SG Level
~
The standard emergency operating instructions (E01's) for Westing-house reactors require use of the narrow range level indication for pipe breaks inside containment or steam generator tube ruptures.
Narrow range level is used as the basis for throttling auxiliary feedwater flow, but only after the affected steam generator has been identified. After the affected SG is identified and isolated, the E01's instruct the operator to maintain SG indicated level within the narrow range. If the water level is below this range, the operator need not know the level accurately because his objective will be to obtain the maximum possible auxiliary feedwater flow.
The presence of steam-generator level within the narrow-rrge Span is also one of the criteria (along with pressurizer pressucc, pres-surizer level, subcooling and auxiliary feedwater flow) for throttling safety injection flow. The E01's require either steam-generator level to be within the narrow-range or auxiliary feedwater flow rate to exceed a specified rate. The purpose is to keep the tube bundle covered or to limit the duration of any uncovery of the tube bundle.
[
l l Potential errvs in narrow-range level may amount to approximately 40% of span in the post-accident monitoring mode. That is, the actual level could be 40% below the ir.dicated level. Since the SG tube bundle is approximately 70% of narrow range span below the bottom narrow range tap, if the indicated water level is maintained within the narrow range, the tube bundle will always be covered.
This is a conservative basis for ensuring that heat removal via the steam generators and natural circulation capability within the l
reactor coolant loops exist.
Page ten When the indicated water level is below the narrow range span or for steam / feed-line breaks outside containment, the wide range level indication will be used by the operators. There is no need for the operators to know the precise level. ihey will know from other l indications if the heat removal function of the steam generators has been lost and will have emergency instructions that will relate the potential error in level indication to containment temperature.
- d. Post-Accident Monitoring - Pressurizer Level The standard emergency operating instructions (E01's) for Westing-house reactors require use of the pressurizer level indication to throttle and/or terminate safety injection following a high energy i line break, after which the pressurizer level is restored.
Reference leg heatup effects could introduce an error in level indication of 40%, with the actual level being lower than the indicated level. The operators will be provided with operating instructions to cover this phenomenon.
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