ML20154R546
| ML20154R546 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 03/21/1986 |
| From: | MISSISSIPPI POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML20154R545 | List: |
| References | |
| TAC-60158, TAC-61038, TAC-61039, NUDOCS 8603310060 | |
| Download: ML20154R546 (25) | |
Text
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- 1. OLCR-NLS-85/06 (Supplemental Submittal)
SUBJECT:
Facility Operating License NPF-29, page 9.
DESCRIPTION: It is proposed to change Operating License Condition 2.C.(28) to read as follows:
MP&L shall have on its nuclear operations staff, one or more corporate. management officials or advisors (who- may be either permanent employees or contracted consultants) who have substantial commercial nuclear power plant operating manage-ment experience and who will advise on all decisions affecting safe operation of the plant.
This requirement shall be in effect until the plant has accumulated at least 6 months at power levels above 90% of full power.
JUSTIFICATION: On November 14, 1985 MP&L submitted proposed changes to the GGNS Unit 1 Operating License. As part of those changes, it was proposed to delete Operating License Condition 2.C.(28),
Advisor to Vice President.
License Condition 2.C.(28) of Facility Operating License NPF-29 requires MP&L to provide an advisor to the Vice President, Nuclear Operations until the plant has accumulated for at least six months at power levels above 90% of full power. This license condition was imposed in June 1982, due to the lack of operating plant experience by corporate management. In March 1985, MP&L filled the pcsition Vice President, Nuclear Operations with Mr. O. D. Kingsley, Jr. who has significant, responsible operational experience.
Mr. Kingsley is a graduate of Auburn University with a B. S.
degree in Engineering Physics. He served in various line-of ficer capacities in the. nuclear submarine service for five years before being employed by Alabama Power Company. The following are some of the positions held during his 14 year tenure there.
Senior Engineer Assistant Plant Manager, Farley Nuclear Plant Plant Manager, Farley Nuclear, Plant Assistant Manager - Nuclear Generation Manager, Nuclear Engineering and Technical Support Director, Nuclear Plant Support - Southern Company In addition to the commercial nuclear power experience of Mr. O. D. Kingsley, Jr., the President and Chief Operating Officer of MP&L, Mr. W. Cavanaugh, III, has extensive commercial nuclear power experience.
8603310060 DR 860321 ADOCK 05000416 PDR J16PMI86031103 - 2 E--------------.------___._.__-._-.____________.-.-
Prior to assuming his present position with MP&L in early 1984, Mr. Cavanaugh worked for Arkansas Power and Light Company for 15 years. He was employed by tha't company in April, 1969, and worked on the design, construction, licensing, and operation of Arkansas Nuclear One Units 1 and 2. The following are some of the positions he subsequently held during his tenure there:
Assistant Plant Superintendent of Arkansas Nuclear One Production Project Manager of Arkansas Nuclear One Manager of Nuclear Services Assistant Director of Power Production Executive Director, Generation and Construction Vice President, Generation and Construction Senior Vice President, Energy Supply.
From April through August 1983, Mr. Cavanaugh was on loan to Louisiana Power and Light Company and served there.as a Senior Vice President - Nuclear Operations with direct responsibility for Waterford 3. He is a graduate of Tulane University with a Bachelor of Science Degree in Mechanical Engineering and served as an officer in the U. S. Naval Submarine Nuclear Program prior to employment with Arkansas Power and Light.
MP&L has had contractors _ fill the position of Advisor to the position of Vice President, Nuclear Operations for over three years with approximately 65 days of that time while the plant was above 90% of full power (first averaged greater than 90%
for a full day on May 12, 1985). MP&L requested deletion of License Condition 2.C(28) in light of the extensive commercial nuclear power experience of its existing Vice President, Nuclear-Operations and its President and Chief Operating Officer in a letter to the NRC dated November 11, 1985 (AECM-85/0360).
Based on subsequent conversations between the NRC staff and MP&L, MP&L now proposes that Operating License Condition 2.C.(28) not be deleted but reworded to allow credit to be taken for the present MP&L management experience.
This proposal supplements and modifies that part of the previously proposed change to the Operating License (submitted on November 14, 1985) dealing with Operating License Condition 2.C.(28). A notice on the significant hazards considerations of the previously. proposed change was published in the Federal Register on December 3, 1985. Because the intent and content-of this proposed modification are essentially unchanged with regard to significant hazards considerations, HP&L believes the December 3, 1985 Federal Register notice suffices to meet the requirements of 10CFR50.91. j i
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SIGNIFICANT HAZARDS CONSIDERATIONS:
The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated because there is no significant change in the intent of the license condition since the underlying requirement of assuring that corporate management has an appropriate level of operating experience or is advised by persons who do until that experience is attained.
The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated because the change in no way affects the design or the procedures involving day to day operation of the plant.
The proposed change does not involve a significant reduction in I the margin of safety because the change is administrative in nature.
Therefore, the proposed change involves no significant hazards considerations.
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av (b) Final evaluations and recomendations from the TDI Owners $ d',
Group Prcgram applicable to GGNS Unit 1 and MP&L's actions in response to this program for the standby diesel gener- 2L ators shall be submitted for NRC review and approval prior 'd j to startup following the first refueling cutage. de it (26) Turbine Dise Integrity (Sutton 10.2.1, SER SSER #1) i Prior to aceuu O p n h e ar r a -rw e a u e ,o m S ring cach cfueling%cutapwMP&L shall ultrasonically inspect d the bores and keyways of the low pressure turbine discs for indications of crackinf. All unacceptable indications and their dispositions shall be reported prior to startup for the next cycle of operation. These inspections shall continue en a Eo;ooo hour mal until the potential for turbine disc cracking has been assessed and an acceptable alternate inspection schedule has been established.
(27) Circulating Water System (Section 10.4.5 SER)
MP&L shall not fill the Unit 2 circulating water system (includ-ing the natural draft cooling tower basin) until Unit I flooding concerns related to this system are resolved to the satisfaction of the NRC staff.
(28) AdvisortoVicePresident(Section13.1.1,SER,SSER#2,SSER#4, SSER #5)
EP hall provide one or more additional staff members, prejoog direct o Vice President principally in charge of far ,
operations, who av substantial comercial nu power plant l INSE8T - o)erating management ex nce and wh act as advisors to tie vice president on all dec a fecting safe operation of the plant. The addit yioPsTaff membe be permanent employees or contracted consultants, but they h Rbe retained in this, advisory position until the plant has operaterf t
leasf6 months at power levels above 90% of full power.
(29) 0'perating Shift Advisor (Section 13.1.2 SER)
At least one individual on each operating shift shall have sub-stantive previous BWR operating experience, including startup and shutdown of a BWR and under conditions that one might expect to encounter during the initial startup and power escalation at the Grand Gulf plant. This individual is not required to be licensed on Grand Gulf Unit I and need not be an MP&L employee, but as a minimum shall be retained on a contract basis to'act as a consultant or advisor to the GGNS shift crew. Such an experienced person shall be assigned to each operating shift until the plant achieves and demonstrates i full power operation. l
INSERT TO OPERATING LICENSE CONDITION 2.C.(28)
MP&L shall have on its nuclear operations staff, one or more corporate management officials or advisors (who may be either permanent employees or contracted consultants) who have substantial commercial nuclear power plant operating mangement experience and who will advise on all decisions affecting safe operation of the plant. This requirement shall be in effect until the plant has accumulated for at least 6 months at power levels above 90% of full power.
J16PMI86031103 - 5
- 2. (NPE-86/04)
SUBJECT:
Technical Specifications Table 3.3.7.2-1 and 4.3.7.2-1 pages 3/4 3-64, 65.
DISCUSSION: The proposed change results from a design change to install a Strong Motion Accelerometer (SMA) on a piping support for the-injection line to the reactor in the high pressure core spray system. This SMA will provide an acceleration time history for a reactor piping location and confirm the post-seismic evaluation for this plant equipment.
As done on several recent Technical Specification changes involving design changes to the plant, it is requested that the NRC issue the change with an open effective date and require that MP&L notify the NRC within 30 days of the effective date.
of implementation of the affected technical specification changes. This design change is scheduled for implementation not later than startup following the first refueling outage.
JUSTIFICATION: Currently, there are five (5) Triaxial Time History Accelerographs (SMA's) installed in GGNS Unit 1. One SMA is located on the Unit I containment base slab such that it measures the input vibratory motion of the base slab. A second
! SMA is located in Unit I containment attached to the drywell wall at El. 150'-6" on the same containment azimuth as the base slab SMA. A third SMA is located in the Unit I auxiliary building attached to one of the standby gas treatment system filter train supports which is seismic Category I equipment. A fourth SMA is located in the standby service water pump house A, which is an independent Category I structure. The fifth SMA is located in the free field approximately 250 feet from any station structure, with axes oriented in the same direction as the containment building accelerometers. All accelerometers, including the one proposed in this submittal, have their principal axes oriented identically, with one horizontal-axis parallel to the major horizontal axis assumed in the seismic analysis.
This proposed change results from.a design change to add a sixth SMA on a reactor support by restart from the first refueling outage as required by Operating License Condition 2.C. (7) . This SMA will be located on a piping support for the injection line to the reactor in the high pressure core spray system. This will allow data collection on a reactor piping support during seismic events.
MP&L believes that the installation of the specified seismic instrumentation in the reactor containment structure along with other Category I structures, systems, and components constitutes an acceptable program as described in the Standard Review Plan (NUREG-0800 Rev. 1 - July 1981) to record data on seismic ground motion as well as data on the frequency and amplitude J16NLSIM86021901 - 2
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relationship of the seismic response of major structures and systems. A prompt' readout of pertinent data at the control room can be expected to yield sufficient information to allow for post event evaluation. With at least one of the two Seismic Triggers indicating greater than 0.0lg acceleration, all Seismic Monitoring System gnsaette magnetic-tape recorders start and record the Seismic acceleration data from the SMA's. Data obtained will be sufficient to determine that the seismic analysis assumptions and the analytical model used for the design of the plant are adequate and that allowable stresses are not exceeded under conditions where continuity of operation is intended. Provision of such seismic instrumentation complies with Regulatory Guide 1.12.
SIGNIFICANT RAZARDS CONSIDERATION:
The design change associated with this proposed technical specification change will provide additional assurance that adequate post seismic event data is captured for analysis.
The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated because it adds required improvements not currently listed in the technical specifications. The purpose of the SMA's is to provide data for post seismic event evaluation und to justify continued plant operation. This instrumentation is not utilized for transient mitigation or detection in the present accident analysis.
The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. This change will add a SMA to the high pressure core spray system piping supports, but will in no way affect the operation of that system. The additional SMA provides data only for post seismic event evaluation.
The proposed change does not involve a significant reduction in l the margin of safety because the addition of another SMA increases the data collection ability for seismic analysis.
Therefore, the proposed change involves no significant hazards considerations.
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INSTRUMENTATION TABLE 3.3.7.2-1 SEISMIC MONITORING INSTRUMENTATION l
MINIMUM MEASUREMENT INSTRUMENTS INSTRUMENTS AND SENSOR LOCATIONS RANGE OPERABLE
- 1. Triaxial Strong Motion Accelerometer
- a. Containment foundation .0.001 to 1.0g 1 l
- b. Drywell 0.001 to 1.0g 1 ,
- c. SGTS Filter Train 0.001 to 1.0g 1
- d. SSW Pump House A 0.001 to 1.0g 1
- e. Free Field 0.001 to 1.0g 1 C. b 4=- TYp'q fupport o.m1 +o 1.og 1 l
- 2. Triaxial Peak Recording Accelerograph
- a. Containment Dome 0.01 to 2g 1
- b. Auxiliary Building Foundation 0.01 to 2g 1
- c. Diesel Generator 11 0.01 to 2g 1
- d. Control Building Foundation 0.01 to 2g 1
- e. Control Room 0.01 to 2g 1
- f. Reactor Vessel Support 0.01 to 2g I
- g. Reactor Recirc. Piping 0.01 to 2g 1
- h. Main Steam Piping 0.01 to 2g 1
- i. LPCS Spray Line 0.01 to 2g 1
- j. HPCS Spray Line 0.01 to 2g 1
- k. SSW Pump House B 0.01 to 2g 1
- 3. Triaxial Seismic Switches
- a. Containment Foundation (SSE) 0.025 to 0.25g 1*
- b. Containment Foundation (OBE) 0.025 to 0.25g la
- c. Drywell (SSE) 0.025 to 0.25g 1* !
- d. Drywell (OBE) 0.025 to 0.25g 1*
- 4. Vertical Seismic Trigger
- a. Containment Foundation 0.005 to 0.05g 1*
- 5. Horizontal 5eismic Trigger
- a. Drywell 0.005 to 0.05g 1*
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- With control room annunciation, j GRAND GULF-UNIT 1 3/4 3-64 AMENDMENT Mo,
INSTRUMENTATION TABLE 4.3.7.2-1 SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL FUNCTIONAL CHANNEL INSTRUMENTS AND SENSOR LOCATIONS CHECK TEST CALIBRATION
- 1. Triaxial Strong Motion Accelerometer
- a. Containment Foundation M SA R
- b. Drywell M SA R
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- 2. Triaxial Peak Recording Accelerograph
- a. Containment Dome NA NA R
- b. Auxiliary Building Foundation NA NA R
- c. Diesel Generator 11 NA NA R
- d. Control Building Foundation NA NA R
- e. Control Room NA NA R
- f. Reactor Vessel Support NA NA R
- g. Reactor Recirc. Piping NA NA R
- h. Main Steam Piping NA NA R
- 1. LPCS Spray Line NA NA R
- j. HPCS Spray Line NA NA R
- k. SSW Pump House B NA NA R
- 3. Triaxial Seismic Switches
- 4. Vertical Seismic Trigger i
- a. Containment Foundation M SA R
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- 5. Horizontal Seismic Trigger
- a. Drywell M SA R l l
MaqbME>rr hdo.
GRAND GULF-UNIT 1 3/4 3-65 3 1
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- 3. (NPE-86/05)
SUBJECT:
Technical Specifications Tables 3.3.3-1, 3.3.3-2, and 4.3.3.1-1; pages 3/4 3-28, -31, -34, and -35.
DISCUSSION: The proposed change to Technical Specifications Tables 3.3.3-1, 3.3.3-2 and 4.3.3.1-1 adds instrumentation, setpoints and surveillance requirements.for ADS actuation instrumentation, and changes the name of the currently identified ADS Timer to ADS Initiation Timer. The instrumentation to be added to these tables are ADS Bypass Timer (High Drywell Pressure) and Manual Inhibit switches. These changes'to the Technical Specifications are needed to supplement a design change required by Operating License Condition 2.C.(33)(f).
Mississippi Power & Light Company (MP&L) is now in the process of developing a design change to satisfy the license condition.
This design change is scheduled for implementation not later t
than startup following the first refueling outage. As with i several recent Technical Specification changes involving design changes to the plant, it is requested that the NRC issue the change with an open effective date and require that MP&L notify the NRC within 30 days of the effective date of implementation of the affected technical specification changes.
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Operating License Condition 2.C.(33)(f) requires MP&L to modify the Automatic Depressurization System (ADS) logic in accordance with Option 4 of the BWR Owners' Group Evaluation of NUREG-0737, Item II.K.3.18, prior to startup following the first refueling. '
. outage. The Option 4 modification is the addition of a timer I
that bypasses the high drywell pressure permissive if the reactor water level is low for a sustained period and. changes i
the low reactor pressure vessel (RPV) water level trip set I point to the top of the active fuel. No change to the low-low-low water level trip is proposed because the trip is provided by the wide range instrumentation and the lowest level measured by
- this instrumentation is 373 inches above vessel zero. As shown I
in Bases Figure B 3/4 3-1 of the Technical Specifications, the j level at the top of the active fuel is 366.3 inches above vessel zero. The RPV water level 1 trip setpoint is 382.7 inches above vessel zero.
In addition to the modifications, the License Condition requires MP&L to provide justification for the timer delay settings, revise the emergency procedures for use of the manual inhibit
- switch, and submit proposed Technical Specification surveillance procedures for the timer and switch.. The modifications and justification for the timer delay settings are addressed below.
The emergency procedures for use of the manual inhibit switch
- will be revised and availabic for NRC review and concurrence prior to completion of the modifications. The propoaed technical specifications surveillance requirementa are addressed with this submittal.
J12PCOL86030401 - 1
Through selected safety / relief _ valves, the ADS functions as a backup to the operation of the high-pressure coolant systems.
The ADS depressurizes the vessel so that low-pressure systems may inject water into the reactor vessel. ADS is activated automatically upon coincident signals of low water levels (Level 1 & 3) in the reactor vessel and high drywell pressure. In addition, a low-pressure emergency core cooling system pump running permissive must be satisfied prior to ADS initiation. A time delay of approximately 105 seconds after receipt of the coincident low water level and high drywell pressure signals allows time for the automatic blowdown to be reset manually if the operator believes the signals are erroneous or if the water level can be restored. For transient and accident events that do not directly produce a high drywell pressure signal (e.g.,
stuck-open relief valve or. steam line break outside containment) and are degraded by a loss of high-pressure coolant systems, manual initiation of the ADS is required to provide adequate core cooling. The present ADS logic design does not satisfy the criteria of Item II.K.3.18 of NUREG-0737 to eliminate the need for operator action because it has not been demonstrated that the high drywell pressure signal would be present for all situations requiring automatic ADS actuation.
The proposed modification will provide a bypass of the drywell high pressure signal after a set time delay and a manual inhibit function. This will further automate the ADS system by providing automatic ADS initiation, if required, for events such as a break external to the drywell or a stuck open safety relief valve. The modification also provides the capability to more easily inhibit ADS operation in accordance with Emergency Procedure Guidelines (EPGs). The manual inhibit switch allows the operator to inhibit ADS operation without repeatedly pressing the reset pushbutton as is required with the current design. One manual inhibit switch will be provided for each division (16 2) of the ADS actuation instrumentation. Each switch will activate a white indicating light and an annunciator to alert the operator of the inhibit action. The pressure relief function and the manual ADS or individual safety relief valve control will not be affected by operation of the manual inhibit switch.
JUSTIFICATION: Modifications: The design modifications required by the Licenae l Condition comply with NUREG-0737 by extending automatic l initiation of ADS to those transient events which do not result in a release of steam to the dryweJ1 but which may require depressurization of the reactor pressure vessel to maintain adequate core cooling. Of the eight design alternatives evaluated by the BWR Owners' Group, the one selected for implementation at CCNS is identified as Option 4. (See Item II.K.3.18 in Supplement 4 to the Safety Evaluation Report).
This design alternative bypasses the high drywell pressure portion of the current logic after a specified time interval, l
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and adds a manual switch which allows the operator to inhibit automatic ADS initiation. The high drywell pressure signal is bypassed by the installation of a " bypass" timer which is actuated on low RPV water level - Level 1. (The presently installed timer initiates ADS after receipt of coincident signals as discussed above.) When the high drywell. pressure bypass timer times out, the high drywell pressure trip is bypassed and ADS is then initiated on a Level I signal alone (coincident with a Level 3 signal and an ECCS pump running and after the 105 second timer runout). The additional logic would not affect the high drywell pressure / low RPV water level initiation sequence insofar as it responds to pipe breaks inside the drywell. It should be noted that once the bypass timer times out, the bypassing of the high drywell pressure would be sealed in and the bypass timer would not automatically reset.
The bypass timer is automatically reset if the Level 1 signal clears before the timer times out.
Timer Delay Settings: The' delay setting for the bypass timer is a double bounded setpoint in which the lower limit is based on the time needed to allow recovery of RPV water level above Level 1 and minimum SLCS injection time during an Anticipated Transient Without Scram (ATWS) event. The upper limit is based on the avoidance of excessive fuel cladding heatup using 10CFR50, Appendix K models.
The upper analytical limit for the bypass timer delay setting was established based on the limiting ECCS event which does not result in a high drywell pressure trip. A review of the FSAR accident analyses determined that the limiting event is a
, main steam line break outside of containment assuming a high pressure core spray failure. Using a conservative maximum bypass timer delay setting of 10 minutes, a performance analysis resulted in a calculated fuel element peak cladding temperature (PCT) of 1862*F which is still well below the criteria of 2200*F as given in 10CFR50.46. The attached figures 1 through 4 document the results of the analysis and reflect the effects on system pressure, water level, heat transfer coefficients, and PCT.
For operations after modification of the ADS initiation logic during the first refueling outage until the second refueling outage, operator action is required to manually inhibit auto-matic ADS initiation under extreme circumstances of a severe ATWS event in accordance with Emergency Procedure Guidelines.
After the SLCS modification is implemented during the second refueling outage, this operator action will no longer be necessary.
It should be noted that' the lower analytical value for the bypass timer setting takes into consideration the SLCS design as modified for ATWS (i.e., two pump operation and injection of 86 gpm through the llPCS line) and the time for recovery of level 1 J12PCOL86030401 - 3
using HPCS during the ATWS. The current GGNS design uses a nominal 43 gpm (SLCS) flowrate injected at the bottom of the core. The SLCS modification (s scheduled for implementation during the second refueling outage. In the interim, MP&L will evaluate the lower analytical value for the bypass timer setting to be used after the SLCS modification. If the 1cuer limit for the ATWS modification is determined to be necessary, MP&L will propose the necessary changes for implementation prior to startup following the second refueling outage.
With appropriate allowances for loop accuracy, calibration accuracy, and drift, the Technical Specification values for the high drywell pressure bypass timer delay setting were determined to be 9.4 minutes for the maximum allowable value and 9.2 minutes for the nominal trip setpoint..
Attachment I to this submittal is an update to the logic diagrams for the Emergency Core Cooling System Actuation Instrumentation reflected in Table 3.3.1-1. These' logics were originally submitted in a letter to Mr. Harold R. Denton f rom Larry F. Dale dated May 8, 1984 (AECM-84/0093).
SIGNIFICAST HAZARDS CONSIDERATIONS:
I a The proposed amendment does not:
- 1) involve a significant increase in the probability or consequences of an accident previously evaluated. The limiting accident analysis affected by this change is the steam line break outside of containment assuming HPCS failure. A reanalysis of this event determined that after the water level reaches Level 1 and with the maximum bypass timer setting of 10 minutes, a PCT of 1862*F would occur.
This PCT is substantially below the criteria of 2200*F as given in 10CFR50.46. Further, the modifications result in an enhancement of the ADS and do not affect performance of the intended safety function. The modifications extend the automatic initiation of ADS to encompass those transient events which do not produce a high drywell pressure signal but may require depressurization of the reactor vessel to maintain adequate core cooling.
- 2) create the probability of a new.or different kind of accident from any accident previously evaluated. The modifications eliminate the need for operator action for th1se events which do not produce a high drywell pressure signal but may require ADS actuation. Although operator action may be required for an extremely unlikely ATWS event, the modifications will allow the actions specified in the emergency procedure guidelines to be performed more reliably.
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Although new instrumentation will be added for the ADS modification, the instrumentation provides operational enhancements and does not inhibit a valid high drywell pressure signal. juum pressure relief function and the manual ADS or individual relief valve control are not affected by operation of the manual inhibit switch.
Further, the manual inhibit twitch represents an improvement in that the switch allows the operator to inhibit ADS operation without repeatedly pressing the reset button.
- 3) involve a significant reduction in a margin of safety because margins of safety are not adverselv affected. The only margin of safety considered is the peak cladding temperature. The criterion of 2200*F as g' iven in 10CFR50.46 is met with significant conservatism since the PCT for the limiting event is 1862'F. Further, the modifications eliminete the need for operator action for those events which do not produce a high drywell pressure signal but may require ADS actuation.
Margins of safety are not affected by the lower limit for the bypass timer delay setting. This limit is determined only by the limiting A7WS evenc and will be a plant-specific ATWS feature unique tc CGNS following ATWS modifications during the second refueling outage.
Therefore, the proposed change dces not involve significant hazards considerations.
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2 TABLE 3.3.3-1 E
EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION O
E MINIMUM OPERABLE APPLICABLE -
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c!- TRIP FUNCTION CHANNELS PER OPERATIONAL z TRIP FUNCTION I *} CONDITIONS ACTION Z A. DIVISION I TRIP SYSTEM w 1. RHR-A (LPCI MODE) & LPCS SYSTEM
- a. Reactor vessel Water Level - Low Low Low, Level 1
- b. Drywell Pressure - High 2((b) 1, 2, 3, 4* , 5* 30
- c. 2 b) 1,2,3 30 LPCI Pump A Start Time Delay Relay
- d. Manual Initiation 1 1, 2, 3, 4* , 5* 31 I/ system (b) 1, 2, 3, 4*, 5* 32
- 2. AUTOMATIC DEPRESSURIZATION SYSTEM TRIP SYSTEM "A"#
- a. Reactor Vessel Water Level - Low Low Low, Level 1
- b. Drvwell fressure -Jinh 2 1,2,3 30 2 1,2,3 c.
d.
GOS h 2nifia%w Th*ed 1 1,2,3 30 31 Keactor vessel Water Level - Low, Level 3 (Permissive) 1 1,2,3 w e. LPCS Pump Discharge Pressure-High (Permissive) 31
% 2 1,2,3 31
- f.
LPCI Pump A Discharge Pressure-High (Permissive) 2
- g. Manual Initiation 1,2,3 31 w -
=
_ _ 2/svstem 1.2~3 32 m 8. DIVISION 2 TRIP SYSTEM (h. ADS Bypas.tr Therer (High DryweH lr'eisure) 2. I, s, 3 21 I. RHR B & C (LPCI IEIDE)Q #enuai Jrdrikit -
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I i2# #3 32.
- a. Reactor vessel Water Level - Low, Low Low, Level 1
- b. 2 1,2,3,4*,5* 30 Drywell Pressure - High i
- c. LPCI Pump B Start Time Delay Relay 2 1,2,3 30 +
- d. Manual Initiation 1 1,2,3,4*,P 31 1/ system (b) 1, 2, 3, 4 * , $* .
32 <
- 2. AUTOMATIC DEPRESSURIZATION SYSTEM TRIP SYSTEM "B"#
- a. Reactor Vessel Water Level - Low Low Low, Level 1 t b.. Drvwell Pressure ._ H i * - 2 1,2,3 , 30
) c. 2 1,2,3 30 l
3 d. [ ADS Neactorhvesser Jnifia%ir water WaLevelred
- Low, Level 3 (Permissive) 1 1,2,3 31'
- e. 1 1,2,3 31 LPCI Pump 8 and C Discharge Pressure - High (Permissive) 2/ pump 1, 2, 3
- f. Manum 1 Initiatian w / system _
31 '
E. s. Aos nypss TiNor Oiph &l Presun) 1. 2. 3 _ - u i
3 h. Manual InMbit ,,,,3 ,1 m I r,1,3 'l 21 t
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TABLE 3.3.3-2
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EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUNENTATION SETPOINTS ALLOWA8LE
@ TRIP FUNCTION TRIP SETPOINT VALUE 4; A. DIVISION 1 TRIP SYSTEM g 1. RHR-A (LPCI MODE) AND LPCS SYSTEM Q a. Reactor Ves:.el Water Level - Low Low Low, Level 1 > -150.3 inches * > -152.5 inches
- b. Drywell Pressure - High 5 1.39 psig
- c. 5 1.44 psig LPCI Pump A Start Time Delay Relay < 5 seconds < 5.25 seconds
- d. Manual Initiation HA NA
- 2. AUTOMATIC DEPRESSURIZATION SYSTEM TRIP SYSTEM "A"
- a. Reactor Vessel Water Level - Low Low Low, Level 1 > -150.3 inches *
- b. > -152.5 inches Drvwell Pressure _ Hinh 31.39psig 31.44psig
- c. LAD 5 Mmee loifiafienJwed i
- d. - Reactor Vessel Water I.evel-Low, Level 3 i 105 seconds i 117 seconds I
> 11.4 inches * > 10.8 inches
- e. LPCS Pump Discharge Pressure-High 145 psig, increasing . 125-165 psig, increasing
- f. LPCI Pump A Discharge Pressure-High. 125 psig, increasing 115-135 psig, increasing
- g. Manual Inttiation -
- MA -- NA -_ -
B. DIVISION 2 TRIP SYSTEM . p.,,uNnNbit A g 1. RHR E AND C (LPCI MODE) ~ ^
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2 -
- a. Reactor Vessel Water Level - Low Low Low, Level 1 > -150.3 inches *
- b. Drywell Pressure - High > -152.5 inches c.
31.39psig 31.44psig LPCI Pump B Start Time Delay Relay < 5 seconds < 5.25 seconds
- d. Manual Initiation HA NA
- 2. AUTOMATIC DEPRESSURIZATION SYSTEM TRIP SYSTEM "B"
- a. Reactor Vessel Water Level - Low Low Low, Level 1 > -150.3 incles*
- b. > -152.5 inches Drywell_ Pressure Hioh _ 31.39psig 31.44psig
- c. (_ ADS 44eef Inifiefs,on Wmer) < 105 seconds
- d. Reactor Vessel Water Level-Low, Level 3 < 117 seconds I 11.4 inches
- I 10.8 inches
- e. LPCI Pump 8 and C Discharge Pressure-High 125 psig, increasing 115-135 psig, increasing
- f. Nanual Initiation ~
g, ADS Byy.s Timer Difkhrt) pressure) + g,w$de s's wen,Q C. DIVISION 3 TRIP SYSTEM
- 1. HPCS SYSTEM
- h. N aw wI InhiEI* ^ - #A NA j t
? a. Reactor Vessel Water Level - Low Low, Level 2 >-41.6 inches * >-43.8 inches L
- b. Drywell Pressure - High 5 1.39 psig i 1.44 psig
- c. Reactor Vessel Water Level - High, Level 8 < 53.5 inches * < 55.7 inches !
- d. Condensate Storage Tank Level - Low I O inches F -3 inches
- e. Suppression Pool Water Level - High 7 5.9 inches 7 7.0 inches
- f. Manual Initiation RA NA t
{ : o TABLE 4.3.3.1-1 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS -
CHANNEL OPERATIONAL E CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH .
$ TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRED E A. DIVISION I TRIP SYSTEM ~
[ 1. RHR-A (LPCI MODE) AND LPCS SYSTEM
Low Low Low, Level 1 S M R("a) 1, 2, 3, 4* , 5*
l b. Drywell Pressure - High. S M R I) 1, 2, 3
- c. LPCI Pump A Start Time Delay Relay NA Q 1, 2, 3, 4*, 5*
- d. Manual Initiation M(b) 1, 2, 3, 4*, 5*
MA R Q.
- 2. AUTOMATIC DEPRESSURIZATION SYSTEM TRIP SYSTEN "A"#
- 4' a. Reactor Vessel Water Level -
- Low Low Low, Level 1 1,2,3
, S M R((a,))
i Y b. Drvwell Pressu -H S M R 1, 2, 3 -
- 2i c. (ADS _44eee Initiaban Tnwer NA M Q 1,2,3
l' Low, Level 3 S M RI ") 1, 2, 3
- e. LPCS Pump Discharge Pressure-High 5 .M R(,) 1,2,3
- f. LPCI Pump A Discharge Pressure-High S R(,) 1, 2, 3
- g. NA M(b)
R NA 1, 2, 3 Mjaup1 Initiation
.i'b f B. DIVISION 2 TRIP SYSTEM I 3 1. RHR B AND C (LPCI M00E) io I s a. Reactor Vessel Water Level -
- - Low Low Low, Level 1 5 M RI *) 1, 2, 3, 4*, 5*
l 3 b. Drywell Pressure - High S M RI *) 1, 2, 3
!$+ c. LPCI Pump B Start Time Delay Relay NA Q 1, 2, 3, +, 5*
l
- d. Manual Initiation NA M(b)
R 1, 2, 3, 4*, 5*
I Q z __
s_
i . h. Aos ss Tio er crywe Mssure)(Higir NA >1 Q 23 3
- i. Manuni InMMt NA R MA I,1> 3
l TABLE 4.3.3.1-1 (Continued)
EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION. SURVEILLANCE REQUIREMENTS
$ OPERATIONAL E CHANNEL c3 CHANNEL FUNCT10NAL CHANNEL CONDITIONS FOR WHICH .
E TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRED
? 8. DIVISION 2 TRIP SYSTEM (Continued)
E- 2. AUTOMATIC DEPRESSURIZATION SYSTEM ,
Q TRIP SYSTEM "B"#
- e a. Reactor Vessel Water Level -
l Low Low Low, Level 1 5 M R 1,2,3
- b. Drvwell Pressure-Hiah S. M R 1, 2, 3
! c. LADS T4ee,. Ini_fiation_ Timed NA M Q 1,2,3
- d. Reactor vessel Water Level - R(,)
Low, Level 3 S M 1,2,3 i
- e. LPCI Pump B and C -Discharge Pressure-High S R(,) 1, 2, 3
- g. M(b)
- h. f. Manual Initiation NA R NA 1,2,3 C. . DIVISION 3 TRIP SYSTEM R
- 1. HPCS SYSTEM
Low Low, Level 2 1,2,3,4*,5*
Tf M b. Drywell Pressure-High##
S 5
M M
R R(,) 1, 2, 3 1, 2, 3, 4* , 5*
i c. Reactor Vessel Water S M R ,
. Level-High, Level 8
- d. Condensate Storage Tank ,
]
Level - Low S M R(,) 1, 2, 3, 4* , 5*
- e. Suppression Pool Water i
Level - High S M R(,) 1, 2, 3, 4* , 5*
R Ib) i Manual Initiation ## NA 1, 2, 3, 4*, 5*
i D.
f.
LOSS OF POWER NA l3 Division 1 and 2
- 1. '
l a. 4.16 kV Bus Undervoltage NA MI ') R 1, 2, 3, 4**, 5**
i n. (Loss of Voltage)
- b. 4.16 kV Bus Undervoltage NA M(,) R 1, 2, 3, 4**, 5**
.l 3 1m (80P Load Shed) ia c. 4.16 kV Bus Undervoltage. NA MI ') R 1, 2, 3, 4**, 5**
]4 (Degraded Voltage)
- 2. Division 3. 4**, 5**
- a. 4.16 kV Bus Undervoltage NA NA R 1, 2, 3, lgD
-(Loss o _
[ e ^l,'_ % ll q Gisk xs m a w NA R NA ! > *> 3 i
- h. Manual inhibit
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l DEFINITIONS FOR
( " CHANNELS", " TRIP SYSTEMS", AND " TRIP FUNCTIONS" FOR EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION TABLE 3.3.3-1 (Continued)
?
ADS Trip Systems
' ADS "A" Trip Unit Parameter Logic ,
3B21-LIS-N691A RPV Level - Lvl 1 l l 21-LIS-N695A *RPV Level - Lvl 3 -
_ __ _.s
,i_______ __
i B21-PIS-N694A Drywell Pressure - HI l I
- j. gB21-K114A ADS Bypass Timer (9.2 Min. )_j---Either l l
lB21-534A Manual Inhibit -Bo tir
~ It '#
IB21-K5A I ADS Initiation Timer (105'Sec.)~
IB21-HS-M629A Manual Initiation -Bo th- l Energize LPCS Pump Disch Pr. - HI
~
( lE21-PIS-N652 i Pilot Either IE12-PIS-N655A LPCI Pump A Disch. Pr. - HI l Sol. A I
B21-LIS-N691E RPV Level - Lvl 1 Both - and Open I B21-PIS-N694E Drywell Pressure - Hi *~ #6 l
l ,B21-K114E ADS Bypass Timer (9.2 Min.)_}-Either- Both- ,
Valves ,
,B21-HS-534A Manual Inhibit Either-g TRIP eB21-HS-M629E Manual Initiation Both- FUNCTION
-E lE21-PIS-N653 LPCS Pump Disch. Pr. - HI ' *#
g IE12-PIS-N,656A LPCI Pump A Disch. Pr. - HI l "g l l I TRIP SYSTEM o i -__ ____ _i l *'One Channel (Typical of 16 shown on this page) l 5
? ?%#
- n nn l' @
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DEFINITIONS FOR "
" CHANNELS", " TRIP SYSTEMS", AND " TRIP FUNCTIONS" FOR EMERGENCYLCORE COOLING SYSTEM ACTUATION INSTRUMENTATION TABLE 3.3.3-1 (Continued)
Trip Unit Parameter Logic ,
iB21-LIS-N691B RPV Level - Lvl 1 1
- E-21-LIS-N695B *RP Level - Lvl 3- .
lB21-PIS-N694B Drywell Pressure - HI -Bo th- i lB21-K114B ADS Bypass Timer (9.2 Min.) p Either-iB21-534B. - Both Manual' Inhibit -
Either- 8
'B21-K5B ADS Initiation Timer (105 Sec.)
lB21-HS-M629B Manual Initiation Both< 1 Energize iE12-PIS-N655B l Pilot LPCI Pump.
LPCI Pump CB Disch.Pr.
Disch. Pr. - HI ]I
.H Either e Sol. B
{E12-PIS-N655C eB21-LIS-N691F RPV Level - Lvl.1 .. Both - and Open
~ lB21-PIS-N694F , Drywell Pressure - Hi 8 S ADS Bypass Timer (9.2 Min.)_j--Either- Both ' Valves IB21-K114F-B21-HS-534B Manual Inhibit Either- , TRIP
,',B21 HS M629FManual e - -
Initiation Both- i FUNCTION
'{
r.
'E12-PIS-N656B LPCI Pump B Disch. Pr. - HI LPCI. Pump.C Disch. Pr. - HI } Either i N
- o. lE12-PIS-N656C a
$ i TRIP SYSTEM c, w_ __ ._ ___ _ _____ __ ___a ,,g 7
- One Channe1 (Typical of 16 shown on this page) y t'5 "
~
E-
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n SS 8"
w-
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r '
l.
DEFINITIONS FOR
" CHANNELS", " TRIP SYSTEMS", AND " TRIP FUNCTIONS" FOR EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION TABLE 3.3.3-1 (Continued)
HPCS System Trip Unit Parameter Logic E22-HS-M616 Manual Initiation I
.'lB21-LIS-N673C RPV Level - Lvl 2 i -Either-l
, ilB21-LIS-N673L RPV Level - Lvl 2
- l -Bo th ; Initiates 3-t i- .
. l HPCS and iB21-LIS-N673G RPV Level - Lvl 2 Either- 1 Starts I
IB21-LIS-N673R RPV Level - Lvl'2- Any I Div. 3 lB21-PIS-N667C Drywell Pressure HI g Eithar q--Bo One i Diesel-1 iB21-PIS-N667L Drywell Pressure HI th---
1
,B21-PIS-N667G Drywell Pressure HI ; Either- l Close HPCS
,B21-PIS-N667R ' Drywell Pressure HI g Pump Disch
.B21-LS-N674C RPV Level - Lvl 8 g Both - Valve eB21-LS-N674L 'RPV Level - Lvl 8 8E22-LIS-N654C Cond Stg Tk Lvl - Low l Switch
,E22-LIS-N654G Cond Stg.TK Lvl - Low I HPCS
'E22-LIS-N655C Supp. Pool Wtr Lvl .High Any 1 Pump
-lE22-LIS-N655G Supp. Pool Wtr Lvl - High - One. 8 Suction
.g
, l l from CST
- 1 I ; to Supp.
R I Pool 4 l 1 TRIP
,. @. 8 I FUNCTION.
I o l TRIP SYSTEM I
m>>
m 1. .
- yQ" g
- One Channel-(Typical of 15 shown on this page) *7$
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. v-l @-
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_ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ - _ _ _ _ _ - _ - _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - _ _ - _ _ _ _ _ _ _ - - - _ __