ML20236M339
| ML20236M339 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 07/31/1987 |
| From: | Andognini G SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| To: | Miraglia F Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20236M344 | List: |
| References | |
| GCA-87-263, TAC-64359, NUDOCS 8708110026 | |
| Download: ML20236M339 (48) | |
Text
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SACRAMENTO MUNICIPAL UTILITY DISTRICT Cl P. O. Box 15830, Sacramento C'A 95852-1830,(916? 452-3211 l AN ELECTRIC SYSTEM SERVINi3 THE HEART OF CALIFORN! A GCA 87-263 i
I JUL 311987 Associate Director for Projects Attn: Frank J. Miraglia, Jr.
U. S. Nuclear Regulatory Commission i Washington, DC 20555 l Docket No. 50-312 Rancho Seco Nuclear Generating Station {
License No. DPR-54 l PROPOSED AMENDMENT NO. 152 REVISION 2
References:
- 1) " Rancho Seco: AFH Minimum Flow Analysis," SMUD Document Z-FHS-I-00150, B&W Document 86-1167930.
- 2) "SMUD Minimum AFH Justification," SMUD Document ERPT-I-0018, B&W Document 51-1167962.
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Dear Mr. Miraglia:
By letter dated December 5, 1986, the District submitted to the Commission Proposed Amendment No. 152 to permit the operation of the Emergency Feedwater Initiation and Control (EFIC) system. Revision 1 to Proposed Amendment No. 152 was submitted by letter dated March 26, 1987.
The District hereby submits Revision 2 to Proposed Amendment No. 152.
Revision 2 provides clarification of the proposed limiting conditions of operation and additional surveillance requirements for the EFIC system.
Revision 2 also revises the required auxiliary feedwater flow rate to a new value of 475 gpm. This reduction in required auxiliary feedwater flow is based on an analysis which shows that on a loss of main feedwater accident with an assumed auxiliary feedwater flow rate of 475 gpm, the primary and secondary system transient will remain within acceptable limits (Ref.1). The validity of the analytical method used is demonstrated for this specific application in Reference 2. This validation addresses concerns identified in a meeting held on May 27, 1987, between District representatives and NRC staff.
Resubmitted in their entirety are the Description of Proposed Changes, the associated Safety Analysis, and the No Significant Hazards Evaluation (Enclosure 1), the proposed technical specifications (Enclosure 2), the Design Bases Report (Enclosure 3), and the System Description (Enclosure 4).
Revisions in Enclosures 1, 3 and 4 are clearly marked by revision bars. Note that the only changes to the Design Bases Report and the System Description concern the revised auxiliary feedwater flow rate.
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p m 88n u n e RANCHO SECO NUCLEAR GENERATING STATION C 14440 Twin Cities Road, Herald, CA 95638-9799;(209) 333 2935
_-_-_-_______________D
o ^ GCA 87-263 Frank J. Miraglia, Jr. gg33 Pursuant to 10 CFR 50.91(b)(1), the Radiological Health Branch of the California State Department of Health Services has been informed of this proposed amendment by mailed copy of this submittal.
Because this is a revision to Proposed Amendment No. 152, no additional I license fee is required.
I I
If you have any questions concerning this submittal, please contact i Mr. Ron Colombo at (916) 452-3211, extension 4236.
Sincerely,
< %y /f a l ".
. Ca Andogn ni Chid Executive Officer, Nuclear Enclosures cc w/atch: i G. Kalman, NRC, Bethesda (2)
A. D'Angelo, NRC, Rancho Seco J. B. Martin (2)
HIPC (2)
INP0 Sworn to and subscribed before me this d @ day of July, 1987.
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UNotary Public l
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n- ..r ENCLOSURE 1 1
Description of Proposed Changes, Associated Safety Analysis, and
" No Significant Hazards Evaluation".
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FACILITY CHANGE SAFETY ANALYSIS ENCLOSURE 1 PROPOSED AMENDMENT NO. 152, Rev. 2 PAGE 1 0F 15 l DESCRIPTION OF CHANGES:
Proposed Amendment No. 152 incorporates changes to the Rancho Seco Technical Specifications required because of modifications to the auxiliary feedwater system and the addition of Emergency Feedwater
! Initiation and Control (EFIC).
Additionally, this Proposed Amendment reduces the minimum total i feedwater flow (from "at least 780 gpm" to 475 gpm) to a pressur- J{
ised steam generator (1050 psig) required by section 4.8 of the Rancho Seco Technical Specifications.
l REASON FOR CHANGE:
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Purpose:
Several significant transients at nuclear power plants with Bab-cock & Wilcox supplied NSSS which occurred during the 1970s were caused by inappropriate, post reactor trip, steam generator feed-water and/or steam pressure control. Reactor Coolant System over-
! coolings at Rancho Seco and Crystal River 3, undercoolings at Davis-Besse and TMI-2, and others were investigated by the NRC in j the spring of 1980. The transients and suggested plant altera-tions are summarized in NUREG 0667, " Transient Response of B&W Designed Reactors." In the summer of 1980, following issuance of NUREG 0667, and in between TMI Short-Term Lessons Learned (NUREG 0578) and the Clarification of TMI Action Plan Requirements (NUREG 0737), SMUD agreed in discussions with the NRC to install "EFIC" and to upgrade the AFW system to substantially comply with the NRC Standard Review Plan for Auxiliary Feedwater Systems (NUREG 0800, Section 10.4.9).
A conceptual design for EFIC and its related plant modifications was submitted in draft form to the NRC in October of 1980 and a preliminary Safety Evaluation Report based on a point by point l comparison with SRP 10.4.9 was received from the NRC in January 1 1981. The salient features of the design at that time were to provide:
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- Isolation of Main Feedwater (MFW) and AFW to prevent contin-ued feeding of a Steam Line Break (SLB) inside containment.
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't t FACILITY CHANGE SAFETY-ANALYSIS ENCLOSURE 1 PROPOSED AMENDMENT NO. 152, Rev. 2 PAGE 2 0F 15
- Failsafe control of ADVs to prevent "mid-range" failure on loss of control power and to prevent a common mode failure which would open ADVs on both main steam lines.
Following the District's initial commitment to install.the EFIC- !
changes, features of the design were found to accommodate the ;
licensing requirements of NUREG 0737 II.E.1.1, II.E.1.2, and II.K.2.10 (AFW Reliability, Safety-Grade AFW Initiate, and Antici-patory Reactor Trip Loss of MFW), and portions of Reg Guide 1.97 (Class 1 Steam Generator Level and Pressure Indication).
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The reduction of the minimum AFW flow' rate to 475 gpm when dis-charging to a steam generator at full ~ pressure is necessitated by l the installation of two flow-limiting venturis in the AFW system.
The flow limit will prevent excessive steam generator tube im- /b pingement velocity and will increase the margin to mitigate exces-sive cooling which might occur for some control or valve failures. )
EVALUATION AND BASIS FOR SAFETY FINDINGS:
EZ11amg2 Subsystems. Commonanta Affected EFIC is a four channel, safety grade, seismic Class 1 AFW initia-tion and control system. The design basis for EFIC is discussed in the Design Basis Report for ECN A-5415. EFIC installation affects the Main Steam System (MSS-), Feedwater System (FWS),
Auxiliary Feedwater System (AFWS), Reactor Protection System (RPS), Safety Features Actuation System (SFAS), Control Room Pan-els (H1SS, H1RC, H2YS, H2PS, and H2SF), Safe Shutdown Panel (H2SD), Integrated Control System (ICS), Interim Data Acquisition and Display System (IDADS), and Safety Parameter Display System (SPDS).
I Safety Functions 21 Affected Systems /Componanta The purpose of EFIC is to perform specific' safety-related primary protective functions in response to various plant conditions. It is a logic, control, and electrical switching system designed to provide the following:
- 1. Initiation of auxiliary feedwater (AFW),
- 2. Control of AFW flow to maintain steam generator level at i appropriate setpoints,
- 3. Steam generator level rate increase control when required to minimize overcooling,
- 4. Isolat' ion of the main feedwater lines of a depressurized steam generator,
f f FACILITY CHANGE SAFETY ANALYSIS ENCLOSURE 1 FROPOSED AMENDMENT NO. 152, Rev. 2 PAGE 3 0F 15 '
- 5. The selection of AFW flow to the appropriate steam genera-tor (s) under conditions of steamline break, main feedwater or emergency feedwater line break downstream of the last check valve, and
- 6. Control the Atmospheric Dump Valves (ADVs) independent of ICS or other safety grade control. Control should minimi=e chal-1enges to the main steam safety valves, and allow cooldown to be controlled from the main control room or the Appendix "R" shutdown area.
l The Auxiliary Feedwater System assures an adequate feedwater sup-ply to the steam generators to remove reactor decay heat during periods when the normal feedwater supply and/or the electrical l supply to vital auxiliaries has been lost (USAR section 10.2.2.2).
The pumps are actuated automatically upon indication of main feedwater pump trip (ARTS), loss of all reactor coolant pumps, low "
SG pressure, low SG 1evel, and SFAS. The capacity of the Auxil-iary Feedwater pumps is sufficient to ensure that adequate feed-water flow is available to remove decay heat and reduce the Reac-tor Coolant System temperature to less than 300 degrees F when the Decay Heat Removal System may be placed into operation (Technical Specification section 4.8 Bases).
l Effecta 2n Safety Functions EFIC will require changes to the descriptions in chapters 6, 7, 8, and 10 of the USAR. Installing EFIC does not change the analysis in Chapter 14 but provides additional margins because EFIC is a fully-Class 1 Control and Initiation System and is, therefore, a more reliable system. An example of this increase in margin is the steam line break analysis which relies on the main steam failure logic (MSFL) to isolate MFW from the steam generator.
EFIC replaces the non-safety-grade MSFL with a safety-grade system that isolates MFW and AFW to a depressurized steam generator.
The minimum AFW flow rate of 475 gpm is less than that required by section 4.8 of Technical Specifications (Auxiliary Feedwater Pump Periodic Testing). This results in a reduction in the margin of safety as defined in the Bases of Technical Specifications, and [L therefore, involves an Unreviewed Safety Question. Calculations (SMUD calculation Z-FWS-IO150) have shown, however, the AFW flow rate of 475 gpm is acceptable and has no effect on the safety function of the AFW system.
Analysis 21 Effects 2n Safety Functions Proposed Amendment 152 includes the Technical Specification chan-ges required by the incorporation of EFIC and other modifications to the AFW system. The proposed Technical Specification changes $h are primarily additions and clarifications to the limiting condi-i
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l l FACILITY CHANGE SAFETY ANALYSIG ENCLOBURE i l PROPOSED AMENDMENT NO. 152, Rev. 2 PAGE 4 0F 15 tions for operation and surveillance standards. The EFIC Auxili-ary Feedwater System Description and the Design Basis Report for f( '
EFIC provide a description of the changes being made.
The NRC has reviewed the District's proposed Auxiliary Feedwater System Upgrade (which includes EFIC) and the NRC's evaluation concluded in their April 7, 1983 and September 26, 1983 SERs that ,
"The proposed AFWS upgrade represents a considerable and accept- I able improvement over the existing design." The SER further stated l "We therefore conclude that, until a staff position is developed regarding the need for further modifications to improve AFWS reliability, operation of Rancho Seco, with the proposed upgraded AFWS design, is ecceptable." j The District has modified the EFIC design to incorporate " lessons learned" from Crystal River 3 and ANO 1 without changing the basic functional goal to provide a highly reliable AFW system.
The failure modes for EFIC are discussed in the DBR for ECN A-5415 and Casualty Events are discussed in Section 6.0 of the EFIC .
System Description. The failure or casualty events discussed in the EFIC System Description include a discussion of the recovery I procedures and design features to mitigate the effects of the assumed event. The failures discussed include Loss of Main Feed-water (LMFW), LMFW with loss of offsite power, LMFW with loss of all AC (onsite and offsite), plant cooldown, turbine trip with and .
without bypass, main feed line break, main steam line break (MSLB) l and AFW line break, small break LOCA, and fire inside and outside the control room. The failure modes discussed in the DBR also include a discussion of AFW valve failure, failure of fiberoptic cables between channels, failure of RPS inputs to EFIC, failure of SFAS input to EFIC, failure of EFIC trip interface equipment, power sources failures for EFIC and EFIC-related hardware, and EFIC control failures.
The events discussed in the EFIC system description include design basis events and hypothetical events for Rancho Seco that have already been analyzed in the USAR. The design of the EFIC con-trols will allow a minimum of 10 minutes before operator action is required over the full spectrum of decay heat rates.
The discussion indicates that the required initial actions are to verify or confirm AFW initiation and flow and steam generator levels to ensure EFIC is functioning as required. EFIC is de-signed to minimize overcooling following a loss of MFW event.
However, this feature of EFIC is not designed to meet single failure criterion. For a SBLOCA, the operator will have to select the ECC level setpoint for steam generator level. For all cases, the operator can take manual control of EFIC (AFW flow, steam generator level, etc.). For all Design Basis Events discussed in USAR Chapter 14, the addition of EFIC does not change the USAR analysis.
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e i FACILITY CHANGE SAFETY ANALYSIS ENCLOSURE 1 i PROPOSED AMENDMENT NO. 152, Rev. 2 PAGE 5 OF 15 The failure modes discussed in the DBR for ECN A-5415 address EFIC system failures. The following is a summary of the failure modes discussed in the DBR:
- 1. AFW valve failure - AFW is controlled by a parallel combination of series sets of valves (parallel flow paths to each steam generator). A failure of one valve ;
will affect only one of the paths in the parallel set. l The existing design does not have Class 1 parallel flow paths to each steam generator.
- 2. Failure of fiberoptic cables between channels - the fiberoptic communication between EFIC channels is de-signed so that a single failure (such as loss of all fiberoptic cables going into one EFIC cabinet) will not result in a failure of EFIC functions to actuate as required. A failure of any cable causes the signal to go to an actuated state.
Should a single event affect more than one channel, it !
could inadvertently cause actuation of either AFW in-itiation or MFW isolation. Initiation of AFW will only f supply water to the steam generators if the steam gener-ator levels are low (starts AFW pumps, but AFW valves are closed due to no demand signal). Isolation of MFW can lead to plant trip. However, for this improbable event AFW is available to provide cooling.
It.should be noted that there are currently several failure modes which exist that lead to LMFW (e.g., ,
failed NNI turbine header pressure sensor). In this ,
l case EFIC represents no additional plant failure modes. _
- 3. Failure of RPS inputs to EFIC - EFIC receives actuation signals from RPS for MFW pumps tripped and RCPs tripped.
For both cases, EFIC looks at the RPS inputs as four channels and actuates based on a one-out-of-two-taken-twice logic. Thus any single failure of the RPS inputs to EFIC will not prevent actuation of EFIC function or cause inadvertent EFIC actuation.
- 4. Failure of SFAS Input to EFIC - The SFAS inputs to EFIC are designed so that a single failure will not stop EFIC i from initiating AFW when SFAS actuates. There are two !
channels of initiate signals sent'to EFIC from SFAS.
There are 2 signals per channel, each signal delivering i a half-trip to the AFW trip module. A loss of power in ;
a SFAS channel will prevent the channel from initiating its corresponding EFIC channel. A signal f ailure in a i SFAS channel could cause a half-trip in its correspon-ding EFIC channel. An actuation of either channel of SFAS is sufficient to actuate one train of AFW, thereby i .
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f 9 FACILITY CHANGE SAFETY ANALYSIS ENCLOSURE 1 PROPOSED AMENDMENT NO. 152, Rev. 2 PAGE 6 OF 15 ensuring AFW initiation and control even assuming a concurrent single failure. Thus a single failure in the SFAS input to EFIC will not result in EFIC failing to initiate, or cause an inadvertent initiation.
- 5. Failure of EFIC trip interface equipment (TIE) - EFIC l actuates various components through the trip outputs of the train A and B TIE cabinets. The outputs of train A are redundant to train B, therefore, a single failure will not cause the failure of more than one train of AFW.
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- 6. Failure of power sources for EFIC and EFIC-related hardware - EFIC AFW power sources are discussed in sec-tion III.C.8 of the DBR for ECN A-5415. It describes the upgraded AFW system as a two-train system with either train capable of supplying the required AFW to both steam generators. It concluded that with channel-ised power and logic, a single failure will neither prevent feed or isolation of AFW flow. The DBR also discusses specific failures and their affects. All of these failures assume concurrent loss of offsite power.
They are:
- a. Failure of Diesel Generator GEA or GEB No AFW components are powered from these diesel generators. However, main steam system branch isolation valves are. Without GEB the normally closed HV-20565 would fail in its last position.
If closed, its EFIC function would be correct. If open, and a major steam leak were occurring, both main steam lines would depressurise. In this event, P-318 would not function using the turbine driver. EFIC would feed both steam generators.
To avoid overcooling, operator action would be re-quired either to close HV-20560 (powered from GEA) or to manually regulate AFW flow,
- b. Failure of Diesel Generator GEA2 or GEB2
- 1. Failure of Diesel Generator GEA2 Without GEA2 power, the AFW pump P-319 would not operate. P-318 is sufficient for all cooling requirements and would be available in either its turbine, or motor driven mode.
Without GEA2 power, MFW block valves HV-20529 and HV-20530 would not function. The EFIC MFW isolation function would still be provided by MFW isolation valves HV-20515 and HV-20516.
I f f FACILITY CHANGE SAFETY ANALYSIS ENCLOSURE 1 PROPOSED AMENDMENT NO. 152, Rev. 2 PAGE 7 OF 15
- 2. Failure of Diesel Generator GEB2 Without GEB2, the motor driver for P-318 will not be available. P-318 should still be func-tional on its turbine driver and P-319 would ,
l still be functional.
Without GEB2, HV-20515 and HV-20516 would fail in their last position. However, without '
I offsite power, the condensate pumps would probably fail and flow through these valves will not occur. In any case, isolation of MFW would still occur via the MFW control and block valves.
- c. Failure of EFIC and AFW indication in the Control I Room Power for control circuits and for backlighting of pushbuttons which control EFIC comes from the EFIC ,
i channel affected e.g., if power to EFIC Channel "A" l
' is lost, the "A" channel EFIC control circuits on HISS will go dark and control will be non-functional.
The Class 1 analog indication on H1SS requires two inputs to be functional; signal and power. If the signal is lost, the display will go "off scale low", i.e., the digital readout will be at its lowest possible value, and the bar graph will flash a single LED in the lowest position. If the 120 VAC power is lost, the indicator will go dark.
Power to the Channel "A" indicators is from the same battery-backed inverted power which powers EFIC Channel "A". Power to the Channel "B" indica-tors is from the same battery-backed inverted power which powers EFIC Channel "B".
Since all Class 1 indications except AFW pump dis-charge pressure have redundant indicators from a different channel, the only process indication lost on loss of a single power source would be one of the pump discharge pressures. Control lights to the back-lighted pushbuttons and the ammeter would be backup indication showing pump operation.
- 7. EFIC Control Failure - The following is a discussion of EFIC control failures. It should be noted that for these failures, the rate of change of RCS and secondary system parameters is not different than would be expec-ted for similar control failures to the. existing AFW and ADV controls.
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FACILITY CHANGE BAFETY ANALYSIS ENCLOSURE i PROPOSED AMENDMENT NO. 152, Rev. 2 PAGE 8 OF 15
- a. Atmospheric Dump Valve Fails Closed l
If the Turbine Bypass Valves do not control the main steam pressure, the pressure would increase !
and be controlled by the code safety valves. If !
the other steam generator is available, and has pressure control, RCS cooling would proceed through it and steam pressure in the impacted generator would follow saturation pressure consistent with RCS hot leg temperature,
- b. Atmospheric Dump Valve Fails Open l
l The energy release would cause main steam pressure to decrease with a resulting decrease in steam generator secondary temperature. The RCS tempera- l ture would decrease. The best response is for the operator to isolate the open ADV(s) using the motor operated ADV isolation valves. However, if steam generator pressure drops below 600 psig, EFIC will isolate MFW and AFW to the affected generator.
control point would shift rapidly to the parallel l control valve.
- d. AFW Valve Fails Open l
The energy required to heat the cold AFW to satura-tion will cause a temperature and subsequent pres-sure decrease in the steam generator. Operator action to isolate the open AFW valve using the series-aligned motor operated isolation valve is the best operator response. Actual valve position indication is available to identify the errant valve. AFW flow will be limited to 1000 gpm by the flow venturis installed in the AFW lines.
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If only one SG is impacted, EFIC will automatic-ally isolate AFW to that steam generator if pres-sure drops below 600 psig. In the event that the excess AFW develops to an overfill condition, the MFW overfill protection and annunciation would alert the operator to the need to isolate the errant valve.
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1 1 FACILITY CHANGE SAFETY ANALYSIS ENCLOSURE 1 FROPOSED AMENDMENT No. 152, Rev. 2 PAGE 9 0F 15
! 1)1oss of power to EFIC "A" or "B" channel, 2)1oss of a control module within EFIC "A" or "B" channel,
- 3) failure of a pressure or level sensing circuit, and 4) f ailed signal tc. a single device. A single failure cannot simultaneously cause failure of control signals from both channel "A" and channel "B". Control failures for either channel would be similar. Therefore, only failures of channel "A" i
will be discussed below.
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Loss of power to channel "A" would cause the ADV(s) on one main steamline to fail closed, and one AFW j l
' control valve to each steam generator to fail open. J During normal plant operation no change in opera- )
l If AFW has been initiated, the i
tion would result.
ADV closure would initially play a minor role be-cause cooling from excess AFW flow would eventually dominate secondary pressure. Manual closure of the series AFW isolation valves is required. Following I repressurization, the failure of the ADV(s) will f become apparent and the course of action is as described in 7.a above.
Loss of one of the two control modules within EFIC channel "A" will cause either a control valve to the "B" steam generator to fail open (See 7.d) or a control valve and the ADV(s) of the "A" steam generator to fail open and closed respectively.
This latter failure becomes a subset of loss of channel power.
Failure of a pressure sensor signal, though pos-sible in either direction, would be expected to fail low. This would cause the ADV(s) on one steam generator to fail closed and one AFW valve on the j same steam generator to fail closed (due to Feed j Only Good Generator or F.O.G.G. logic). Manual I control of both valves through EFIC would still be possible.
Failure of a low range level sensor, though pos-sible in either direction, would be expected to fail low. If it failed low, and AFW had been initiated, one AFW control valve would fail open (See 7.d). If it failed high, one AFW control valve would fail closed (see 7.c).
Failure of a wide range level sensor, though pos-sible in either direction, would be expected to i
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FACILITY CHANGE SAFETY ANALYSIS ENCLOSURE 1 PROPOSED AMENDMENT NO. 152, Rev. 2 PAGE 10 0F 15 fail low. This would lead to like scenarios for a failed low level sensor, but only if all RCPs were not running.
A failed signal to a single controlled component could cause a valve to open or close. Those events are described in 7.a to 7.d. However, due to the nature of the 4-20ma control circuits used, fail-ures which produce 4ma or less are the expected failure modes. Therefore, the expected fail state for a single component would be closed for an ADV(s) or open for an AFW control valve.
- f. EFIC Sensing Line Failures i
The single EFIC control failure discussion assumes loss of only one sensor or loss of power to all sensors in one channel. EFIC sensors may share process sensing lines with other EFIC channel sen-sors or other system sensors depending upon speci-fic sensor installation details. This includes steam generator level and pressure instrumentation.
In both cases, the design uses some shared instru- A 1 ment taps. 2 The lower OTSG taps are each shared by 2 EFIC channels !
of both narrow and wide range level indication and one l channel of NNI level indication. The middle (narrow I range) and upper (wide range) OTSG taps are each shared by 2 EFIC channels. The lower level taps are T'eed and have two root valves, one for the EFIC channel tubing and one for the NNI tubing. Excess flow check valves are installed downstream of each of the root valves on the lower taps. These ensure that a tubing failure on either EFIC or NNI sensing line will not affect the other level sensing line on that tap. The middle and upper level taps also have root valves.
The failure of a bottom level tap will cause the trans-mitters to fail low, while failure of a reference leg (top level) tap will cause the transmitters to fail high. The failure of the individual taps or associated instrument tubing will result in the following:
Lower OTSG Tap -
l a) Failure of Lower EFIC Tubing -
Failure of lower EFIC tubing would cause the b
shared EFIC channels on that tap to indicate level l
i offscale low. This would cause AFW initiation and full AFW flow to the affected steam generator l
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FACILITY CHANGE SAFETY ANALYSIS ENCLOSURE 1 PROPOSED AMENDMENT NO. 152, Rev. 2 PAGE 11 OF 15 while MFW is still flowing. The control room j operator would take action to terminate this The operator has several indications that j event.
this failure has occurred. They include:
- 1) Both narrow and wide range EFIC level indica- l /h tion on the affected channel would be off-scale low while the unaffected channel would show normal levels for that range of opera-tion.
- 2) Both EFIC channel AFW flow indicators would /h show high, steady flow to the affected steam generator and demand indication and actual position indication for the valve on the affected channel would show the valve to be wide open.
b) Failure Between Tap and Excess Flow Check Valve -
A failure in the short length of pipe between the steam generator tap and the excess flow check valves would affect one channel of NNI startup l level and full range level as well as the two channels of EFIC steam generator level. The ef-l feet on the NNI and possible consequential control I action is not changed from similar potential fail- /h j ures in the existing plant and could result in excessive MFW flow to the affected steam genera-l tor. The concurrent affect on EFIC would be as described above for an EFIC sensor tubing failure.
Additionally, the EFIC MFW overfill protection circuitry would be available to automatically isolate MFW to the affected steam generator al-though due to the incident, only two EFIC channels would be available to initiate the automatic iso-lation function.
Middle OTSG Ian (top of narrow range level indication) -
failure of this tap would not cause automatic EFIC actuation. However, both EFIC channels on the affected tap would not be available for auto initiate of AFW on low steam generator level. The unaffected channel of EFIC would provide proper feed control to the affected i steam generator with no operator action required. The ;
operator would be alerted that this problem has occurred (without AFW initiate) because affected channel narrow range level indicators would be reading offscale high. j 1
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I I FACILITY CHANGE BAFETY ANALYGIS EHCLOGURE 1 PROPOSED AMENDMENT NO, 152. Rev. 2 PAGE 12 OF 15 Top OTSG Tap (top of wide range level indication) -
failure of this tap would cause the affected EFIC chan-nels to indicate high steam generator level and cause i one EFIC channel to isolate MFW to the affected steam generator. The narrow range indication would actuate AFW on low steam generator level. If at least one RCP was running, EFIC would control AFW via the narrow range indication. ,
If EFIC goes to the natural circulation mode (no RCP I running) the affected EFIC channel would not feed the ,
affected steam generator. However, the unaffected EFIC channel would properly feed the steam generator without immediate operator action required. ;
If this failure caused the steam generator pressure to i drop below 600 psig, EFIC would isolate this generator via the vector logic (FOGG). Operator action would be required to return this steam generator to service.
The operator has several indications that this failure has occurred. They include:
- a. The affected channels wide range EFIC level indica- l tion would be offscale high while the unaffected channel indication would show normal for that range of operation.
- b. A MFW isolation valve or MFW control valve station to the affected steam generator would indicate l closed with the remaining MFW isolation or control valve station indication open.
Eight pressure transmitters (pts) are employed by EFIC to indicate steam generator pressure. Four of the pts are installed on 4 separate, independent taps. The other four pts are installed in 2 taps (2 pts per tap).
The shared pressure taps are on the main steam sample lines (one on "A", and one on "B").
SG Pressure Ian - Failure of a shared pressure tap would cause both pts on that tap to indicate low steam genera-tor pressure. EFIC would then initiate AFW and isolate MFW to the affected steam generator. One EFIC channel would continue to control level in the steam generator because only one AFW control path (2 series valves) would be isolated. The other paralleled AFW control path would still provide controlled AFW to the affected steam generator. Throughout this failure mode, the unaffected steam generator would still be fed by the MFW
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F 1 1 FACILITY CHANGE SAFETY ANALYSIS ENCLOSURE 1 PROPOSED AMENDMENT NO, 152. Rev. 2 PAGE 13 0F 15 l
system, with a fully operational AFW system and EFIC
! system available to control steam generator level if MFW !
l is lost. No immediate operator action would be required because EFIC would still regulate AFW to ensure there is no overcooling or undercooking of the RCS. Full manual ,
AFW control and indications would be available to the operator in the control room. This includes all indica-tion and control provided by EFIC except for steam
! generator pressure for the affected channels and steam generator (one steam generator pressure indicator on H1SS panel).
I The failure modes discussed above show that a single failure of the " shared" instrument taps affects only one steam generator and that the affects can be mitigated by the operator in the control room. Tap or instrument l
i line failures will not deprive the operator of necessary indications or annunciation pertaining to AFW control or status of the secondary side of the steam generator.
The failure of a shared instrument tap will not cause undercooking of the NSSS. However, logic required to automatically initiate AFW for certain failure modes l
(failure of the OTSG middle level tap) is reduced,to 2 out of 2. Also, for certain single failures (failure of the OTSG low level tap), operator action is required to reduce AFW flow to mitigate overfeed. This is similar to some single active failures which will require opera-tor action to diagnose and isolate feedwater coolant flow to prevent overcooling (such as loss of a fail open AFW control valve or its power supply). Again, even with these shared tap failure modes, EFIC represents a significant plant enhancement compared to the existing system design.
- 8. Instrument Air System Failure - Four independent Class 1 backup air supply systems are provided to assure power is available to EFIC-related air-operated valves in the event of the loss of normal air supply. One system supplies power for the MFW, Startup Feedwater, and AFW control valves feeding the "A" OTSG; another system supplies power for same valves feeding the "B" OTSG. j Two systems supply power for ADVs; one for the ADVs on the "A" main steam line and one for the ADVs on the "B" ;
main steam line. Each system is sized to provide at {
1 east two hours of air supply.
1 A Babcock & Wilcox report to the B&W Owner's Group (Auxiliary /
Emergency Feedwater System Review, March 1987) identified several {
AFW design bases which need improvement in light of operating f{
experience. The findings include: ;
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l i I-FACILITY CHANGE SAFETY ANALYSIS ENCLOSURE 1 PROPOSED AMENDMENT NO. 152, Rev. 2 PAGE 14 0F 16 (
t "Many EFW systems are sized with very high flow papacity ,
relative to the need for heat removal. Capacity that is in ;
excess of heat removal needs can cause reactor coolant !
overcooling when all EFW trains are actuated. The over-cooling results from steam quenching and uttem depressuriza- j tion." ,
j One of the findings and recommendations of the rd/ich was: I "To lower the susceptibility for overcooling caused by ex- a cess flow capacity new bases and calculational methods that 1 permit flow reduction are presented...' j Two flow-limiting venturis are to be installedtin the AFW system to reduce the maximum AFW flow to less thars 1000 gpm. Calcula-tions have shown that presently, AFW flow from both pumps to a single depressurized steam generator'is capable pf exceeding the l administrative limit of 1800 gpm. Tils lim 1.t is' enforced to !
prevent excessive flow-induced vibration of the steam gnnerator tubes. The installation of the flow venturis will reduce this maximum available AFW flow to a single, depressurized steam gener-ator to approximately 1000 rpm, while AFW flow f rom a single pump ,
to a pressurized steam generator (1050 psig) will be at leaat 475 4 gpm. This capacity is less than thet required by the Qases of Technical Specification section 4.8, however, calculation Z-FWS- 4 10150, " Rancho Seco: AFW Minimum Flow Analysis" has determined A i
that a flow rate of 475 spm is acceptable. l
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The purpose of the calculation was to provide the engineeri.ag analysis to justify a minimum AFW flow rate lower than the current >
Technical Specification value of 780 gpm. This is based c:i the ,
conservative (no credit taken for Anticipatory Reactor Trim) loes / ,' i of feedwater transients from 102% of design power (2772 Plet). . T2 n acceptance criteria used were that RCS pressure should not excer:d- 7 2750 psig, that the pressurizer should>not go solid, and that flow to the pressurizer quench tank due to pressurizer safety valve actuation should not exceed. 10,000 lbm of saturated steam. The ,
analysis of a loss of feeduater, with no anticipatory reactor l trip, and with a maximum AFW flow og 475 gpm, mot all of the acceptance criteria and vas accepttble. Additional informatica .
about the reduction in ATA flow rate is given in Engineering ;!
Report ERPT-0018, "SMUD Minimum AFW Justification".
The loss of main f eedwater represents the higdest heat _ removal capability required of the Auxiliary Feedwater Systerifor accident i mitigation. The result of a loss of main feedwater accident and l resulting AFW performance is discussed in section 14.1.2.8 of the l USAR, k22a 21 Electr(s Rower. i a
Based on the analysis, the function of the AFR system (to remove j decay heat and reduce the RCS temperature to less than 300 degrees F when the Decay Heat Removal System is placed into operation) is 1 -
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FACILITY CHANGE SAFETY ANALYSIS ENCLOSURE 1 f PAGE 15 OF 15 PROPOSEL AMENCHENT Noi 152 Rev. 2 o / e j
unafle,cted by<the redu'etion in AFW flow rate. [h j je ; i ,! L ,
!. hmmarz ./ \ * -
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( 4 Proposed > Amendment No'. 152 incorporates changes to the Rancho Seco j
., Technical (Grecifications required because of modifications to the g' 5 AnxiAiary Feeswater Statem and the addition of Emergency Feedwater 3 '
I:iitiettion >and Control (EFIC) .
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'EFIC is bounded by the design basis and the safety analysis as i described in the Rancho Seco USAR. EFIC provides additional ]
margin to the existing USAR Chapter 14 analysis. The September 2 6 ,' 1983 NRC SER documenti.the acceptability of the design basis.
The system, as proposed, is bounded by the NRC-approved SER dated September 26, 1983 , However, the re: duction in required AFW flow, until reviewed,and approved by the NRC, involves an Unreviewed Safety Question. /h l EFICdoesnotincreasetheprob/.bilityofoccurrenceortheconse-qtences of an accident or malfunction of equipment important to l safsyE previously evaluated inithe SAR, because the September 26, !
1963 NRC SER documents the aepeptability of the design basis. It )
g is also an upgrade of existirg1 plant systems and enhances SAR
,' accident analyses.
I , EFId dos /s' not create the possibility for an accident or malfunc-tion [
because the'Sertember 26, 1983 NRC SER documents the acceptability of the design basis.
1
-EFIC does not reduce the vergin of safety as defined in the bases /h for ths Technical Srjecifications, because the September 26, 1983 H3C, SER documents the acceptability of the design basis. It
$1ncreaset the margin of safety by upgrading the reliability of the (xisting (plant systams. The reduction in the minimum AFW cflow rato (under wor'st case conditions) below that required in the present section 4.8 of Technical Specifications reduces the margin ;
- of safety as defined in the basis for Technical Specifications.
'Fowever, the reduction of the required AFW flow rate, coupled with l 1he. installation of flow-limiting venturis is an increase in the mardis of saioty because it provides the necessary balance between f(
protsetion against design basis undercooking events and protection ag inst non-design basis overcooling events.
Bodauso Technical Specification Bases must be revised to reflect this chan'ge in AFW flow rates, Proposed Amendment 152, Rev. 2 involves an Unreviewed Safety Question as defined by 10 CFR 50.59 (a)(2).
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DISCUSSION OF CHANGES I The following section provides a discussion of the.
proposed changes to the Technical Specification hnd a-comparison between the existing specification and the proposed specification.
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_ _ . _ _ _ _ . _ _ _ _ _ _ . _ .__ ___ _ _ _ _ . _ _- _ . _ _ _ _ . _ _ _ _______________._________m._____.____ __
- 1. Existina Specification:
1.2.10 Remain Critical I A technical specification that requires that the reactor shall not remain critical shall mean that an uninterrupted normal hot shutdown will be completed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
New Specification:
1.2.10 Remain Critical l
A technical specification that requires that the reactor shall not remain critical shall mean that an uninterrupted normal hot shutdown procedure will be completed within j 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> unless otherwise specified.
1.20 Vector Logic ,
1 A set of circuitry in each channel of the Emergency Feedwater Initiation and Control (EFIC) which, once Auxiliary Feedwater (AFW) i has been initiated, determines whether AFW to l a steam generator should be allowed or '
terminated and the signal output for each l EFIC channel to the AFW valves associated with that channel.
Discussion:
(
The revision to the definition " remain critical" clarifies that certain specifications have action times different from that required by the existing :
I definition. The addition defines the term vector logic as used in the EFIC system.
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- 2. Existina Specifications:
3.1.1.2 Steam Generator A. One steam generator shall be operable whenever the reactor coolant average temperature is above 280 degrees-F.
New Specification 3.1.1.2 Steam Generator A. Two steam generators shall be operable whenever the reactor coolant average l temperature is above 280 degrees-F except l as described in 3.1.1.2.B.
B. With one or more steam generator (s) inoperable due to excessive leakage per l
3.1.6.9, bring the reactor to cold l shutdown conditions within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. i l
C. With one or more steam generator (s) l inoperable due to steam generator l defective tube (s), restore the inoperable j generator (s) to operable status prior to l increasing the reactor coolant average I temperature above 200' degrees-F.
Bases:
When the reactor is not critical but TAV is above 280 degrees-F, one steam generator provides sufficient heat removal capability for removing decay heat. However, j single failure considerations require that both steam generators be operable.
Discussient This change requires both steam generators to be operable and defines the corrective action for inoperable steam generator (s). This revision is consistent with analysis supporting the EFIC system by requiring both steam generators to be operable to catisfy single failure considerations.
3
- 3. Existina Specifications:
3.4.1 The reactor shall not remain above 280 f degrees-F with irradiated fuel in the pressure l vessel unless the following conditions are met:
[ 3.4.1.1 Capability to supply feedwater to one steam l
generator at a process flow rate corresponding j l
to a decay heat of 4-1/2 percent full reactor l power from at least one of the following means. t A. A condensate pump and a main feed pump or B. A condensate pump or C. An auxiliary feedwater pump.
The required flow rates are: !
Feedwater temperature Required flow degrees-r gpm ,
40 743 60 756 1 90 780 3.4.1.2 Two steam system safety valves are operable per steam generator.
3.4.1.3 The turbine bypass system to the condenser j shall have one valve operable or the j atmospheric dump system shall have a minimum ;
of 1 of 3 valves operable per steam generator. j 3.4.1.4 A minimum of 250,000 gallons of water shall be available in the condensate storage tank. l 3.4.2 In addition to the requirements of 3.4.1, the reactor shall not remain critical unless the ,
following conditions are met: -
l 3.4.2.1 Seventeen of the eighteen main steam system l safety valves are operable. !
i 3.4.2.2 When two independent 100 percent capacity {
auxiliary feedwater flow paths are not available, the capacity chall be restored ;
within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or the plant shall be placed i in a cooling mode which does not rely on steam t generators for cooling within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
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3.4.2.3 When at lesst one 100 percent capacity auxiliary feedwater flow path is net available, the reactor shall be made subcritical within four hours and the facility .
placed in a shutdown cooling mode which does {
not rely on steam generators for cooling within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
Bases: l l
The feedwater system and the turbine bypass system are i normally used for decay heat removal and cooldown above 280 degrees-F. Feedwater makeup is supplied by operation of a condensate pump and main feedwater pump.
In the event of complete loss of electrical power, )
feedwater is supplied by a turbine driven auxiliary I feedwater pump which takes suction from the condensate j
, storage tank. Steam relief would be through the l
! system's atmospheric relief valves. ]
i If neither main feed pump is available, feedwater can be l supplied to the steam generators by an auxiliary feedwater pump and steam relief would be through the l turbine bypass system to the condenser. i a
In order to heat the reactor coolant system above 280 l degrees-F, the maximum steam removal capability required j t is 4-1/2 percent of rated power. This is the maximum j decay heat rate at 30 seconds after a reactor trip. The !
I requirement for two steam system safety valves per steam generator provides a steam relief capability of over 10 percent per steam generator (1,341,938 lb/h). In j addition, two turbine bypass valves to the condenser or two atmospheric dump valves will provide the necessary capacity.
The 250,000 gallons of water in the condensate storage tank is the amount needed for cooling water to the steam generators for a period in excess of pne day following a complete loss of all unit ac power.tl>
Theminimumreliefcapacity(g{17steamsystemsafety valves is 13,329,163 lb/hr. This is sufficient capacity to protect the steam systeg ynder 3
the design overpower condition of 112 percent.1 /
References:
(1) FSAR paragraph 14.1.2.8.4 (2) FSAR paragraph 10.3.4 (3) FSAR Appendix 3A, Answer to Question 3A.5 5
l New Specifications:
3.4.1 The reactor coolant system shall not be brought or remain above 280 degrees-F with irradiated fuel in the pressure vessel unless the following conditions are met:
A. Capability to remove decay heat by use of two steam generators as specified in 3.1.1.2.
B. One atmospheric dump valve per steam generator shall be operable. :
C. A minimum of 250,000 gallons of water ,
r shall be available in the condensate !
storage tank.
l D. Two main steam system safety valves are operable per steam generator.
E. Both auxiliary feedwater trains (i.e.,
pumps and their flow paths) are operable.
F. Both trains of main feedwater isolation on each main feedwater line are operable.
G. Four independent backup instrument air bottle supply systems for ADVs and MFW, SFW, and ATW valves are operable. ,
l With less than the above required components operable be on decay heat cooling within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. I 3.4.2 The reactor shall not be brought or remain critical unless the following conditions are met:
A. Capability to remove decay heat by use of two steam generators as specified in 3.1.1.2.
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i B. One atmospheric dump valve per steam I i
generator shall be operable except that:
(1) with only one atmospheric dump valve f operable, restore'an inoperable valve for I the other steam generator within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> j or be in hot shutdown within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and i on decay, heat cooling within the next 12 l hours; (2) with no atmospheric dump 1 valves operable, restore at least 1 !
inoperable valve within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in j hot shutdown within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and (
on decay heat cooling within the next 12 l hours. l C. A minimum of 250,000 gallons of water shall be available in the condensate storage tank I except that with less than the minimum volume, restore the minimum volume within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in hot shutdown within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and on decay heat cooling within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
D. Seventeen of,the eighteen main steam safety l valves are operable except that with less. j than the minimum number of valves, restore the inoperable valves within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in hot shutdown within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and ,
on decay heat cooling within the next 12 )
hours.
1 E. Four turbine throttle stop valves are l operable except that with less than the minimum number of valves, restore the inoperable valves within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in hot shutdown within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and on decay heat cooling within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
F. Both auxiliary feedwater trains (i.e., pump and their flow path) operable except that; (1) With one auxiliary feedwater train inoperable, restore the train to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in het shutdown i within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and on decay heat
! cooling within the next-12 hours; (2) With
! both auxiliary feedwater trains inoperable, the reactor shall be made subcritical within four hours and the reactor shall on decay heat cooling within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
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. , i G. Both trains of main feedwater isolation on each main feedwater line are operable except that; (1) With one main feedwater isolation train inoperable, restore the train to j operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in hot '
shutdown within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and on decay heat cooling within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />; (2) With both main feedwater isolation trains inoperable, the reactor shall be made suberitical within four hours and the reactor shall be on decay heat cooling within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
H. Two independent backup instrument air bottle supply systems (one per steamline) for ADVs are operable except that:
- 1) With one system inoperable, restore the system to operable status within 7 days or be in hot shutdown within j the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and on decay heat cooling within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
- 2) With two systems inoperable, restore l at least one system within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> i or be in hot shutdown within the next i 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. With one system restored to I operable status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, follow 3. 4. 2.H. (1) . I I. Two independent backup instrument air I bottle supply systets (one per feed water l line) for MFW, SFW, and AFW control valves are operable except that with either one or both system (s) inoperable, I restore the inoperable system (s) within 7 1 days or be in hot shutdown within the I next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and on decay heat cooling within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
Bases:
The feedwater system and the turbine bypass system are normally used for decay heat removal and cooldown above 280F. Main feedwater is supplied by operation of a condensate pump and main feedwater pump. If neither l main feed pump is available, feedwater can be supplied to the steam generators by an auxiliary feedwater pump.
Steam relief capability is provided by the system's l atmospheric dump valves.
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1 The auxiliary feedwater system is designed to provide !
sufficient flow on loss of main feedwater to match decay I' heat plus Reactor Coolant Pump heat input to the Reactor i
Coolany4fystembeforesolidpressurizeroperationcould occur.
l The 250,000 gallonc of water in the condensate storage tank is sufficient to remove decay heat (plus Reactor l Coolant pump heat for two pumps) for approximately 13 I
hours. This volume provides sufficient water to remove the decay heat for approximately 5.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> and to subsequentlycooltheplanttotheggysystempressureat a cooldown rate of 50 degrees-F/hr.
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Theminimumreliefcapacity(g{17steamsystemsafety valves is 13,328,153 lb/hr. This is sufficient l
capacity to protect the steam systeg ynder the design overpower condition of 112 percent.l 3' Both trains of main feedwater isolation on each main feedwater line are operable. Train A of main feedwater isolation is comprised of main feedwater control valves, main feedwater block valve and startup feedwater control valve. Train B of main feedwater isolation is comprised of main feedwater isolation valves.
Four independent Class 1 back-up air supply systems are provided to assure power available to certain air operated valves in the event of the loss of normal air supply. One system supplies power for the MFW, Startup Feed Water (SFW), and AFW control valves feeding the "A" OTSG; another system supplies power for same valves feeding the "B" OTSG. Two systems supply power for ADV's with one for the ADVs on the "A" main steam line and one for the ADVs on the "B" main steam line. Each system is sized to provide at least 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of air supply.
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References:
(1) B and W Document 32-1141727-00, " Heat Removal Capability of SMUD CST," March 1984.
(2) FSAR paragraph 10.3.4 (3) FSAR Appendix 3A, Answer to Question 3A.5 (4) B and W Calculation 86-1167930, " Rancho Seco: AFW Minimum Flow Analysis," (SMUD Calculation No. Z-FWS 101 9
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l Discussion:
This specification defines the additional components required to be operable by the installation of EFIC and-the amount of time components can be. inoperable prior to taking corrective action. The requirements for both steam generators and both auxiliary feedwater trains to be operable is consistent with the approach as described.
in item 1.
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- 4. Existina Specification i 3.5.1.1 Startup and operation are not permitted unless f the requirements of table 3.5.1-1, Columns A and B are met: .
1 3.5.1.2 In the event the number of protection channels I operable falls below the limit given under Table 3.5.1-1, Columns A and B, operation shall be limited as specified in Column C.
In the event the number of operable Process l Instrumentation channels is less than the i Total Number of Channel (s), restore the j inoperable channels to operable status within !
7 days, or be in at least hot shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. If the number of operable channels is less than the minimum channels operable, either restore the inoperable channels to operable within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in i at least hot shutdown within the next 12 )
hours. ]
Table 3.5.1-1 ,
1 INSTRUMENTS _ OPERATING _ CONDITIONS (A) (B) (C)
Operator Action if i Minimum Conditions of j Total Number Channels Columns A and B i Functional Unit of Channels Operable Cannot be Met.
- 1. Low Main Feedwater See Section Pressure: Start 3.5.1.2 Motor Driven Pump and Turbine Driven Pump 2 1
- 2. Phase Imbalance /
Underpower RCP:
Start Motor Driven and Turbine Driven Pumps 2 1 1
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New Specifications:
3.5.1.1 Startup and operation are not permitted unless the requirements of Table 3.5.1=1, Columns A and B are met.
I l 3.5.1.2 In the event the number of protection channels l l operable falls below the limit given under i Table 3.5.1-1, Columns A and B, operation l
shall be limited as specified in Column C.
In the event the number of operable Process Instrumentation or EFIC system channels is l less than the Total Number of Channel (s), I l
restore the inoperable channels to operable status within 7 days, or be in at least hot shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. If the I number of operable channels is one less than the minimum channels operable, either restore the inoperable channels to operable within 48 (
hours or be in at least hot shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. If the number of operable channels is two less than the minimum j channels operable, the reactor shall be made l subcritical within four hours and on decay heat cooling within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
3.5.1.7 For calibration or maintenance of an Emergency l Feedwater Initiation and Control (EFIC) channel, a key operated " maintenance bypass" l switch associated with each channel will be used which will prevent the initiate signal i from being transmitted to the Channel A and B i trip logic, only one channel shall be locked (
into " maintenance bypass" at any one time.
l 3.5.1.8 If a channel of the RPS is in bypass, it is permissible to bypass only the corresponding channel of EFIC.
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Table 3.5.1-1 INSTRUMENTS _ OPERATING _ CONDITIONS (A) (B) (C) )
Operator Action if i Minimum Conditions of Total Number Channels Columns A and B l Functional Unit of Channels Operable Cannot be Met l l
Emergency Feedwater Initiation and Control (EFIC) System
- 1. AFW Initiation
- a. Manual 2 2 See 3.5.1.2.
- b. Low Level, SGA or B 4/SG 3/SG See 3.5.1.2. May b (Note 2) (Note 1) bypassed below 750 I psig OTSG pressure
- c. Low Pressure 4/SG 3/SG See 3.5.1.2. May SGA or B (Note 1) bypassed below 750 (Note 3) psig OTSG pressure
- d. Loss of MFW 4 3 See 3.5.1.2. Loss Anticipatory (Note 1) of MFW Anticipator Reactor Trip Reactor Trip is effectively bypass in RPS below 20 percent power.
- e. Loss of 4 RC 4 3 See 3.5.1.2. May Pumps (Note 1) bypassed below 750 psig pressure.
- f. Automatic Trip 2 2 See 3.5.1.2.
l Logic
- 2. SG-A Main Feedwater Isolation
- a. Manual 2 2 See 3.5.1.2.
- b. Low SGA pressure 4 3 (Note 1) See 3.5.1.2. May b (Note 3) bypassed below 750 psig OTSG pressure
- c. Automatic Trip 2 2 See 3.5.1.2.
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i l Table 3.1.1-1 (Continued) q INSTRUMENTS OPEPATING CDNDITIONS (A) (B) (C) l Operator Action if l Minimum Conditions of '
i Total Number Channels Columns A and B )
Functional Unit of Channels Operable Carnot be Met I
- 3. SG-B Main Feedwater i Isolation i
- a. Manual 2 2 See 3.5.1.2.
- b. I04 SGB 4 3 (Note 1) See 3.5.1.2. May be Pressure bypassed below 750 l (Note 3) psig OTSG pressure.
- c. Autcmatic Trip 2 2 See 3.5.1.2.
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, 4. ATW Valve Ccmmands (Vector)
! a. Vector Enable 2 2 See 3.5.1.2 l b. Vector Module 4 3 See 3.5.1.2 (Note 4) l c. Control Enable 2 2 See 3.5.1.2 l l d. Control Mcdule 2 2 See 3.5.1.2 i
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- Nete 1 The number of minimum channels operable may be reduced to 2 provided one of the inoperable
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channels is in a tripped state.
1 Note 2 Low level AFW initiation has a maximum of a 10.0 )
second delay.
Note 3 Low pressure AFW initiation has a maximum of a 3.0 second delay.
Note 4 SG Pressure Difference AFW Valve Command (Vector) has a maximum of a 10.0 second delay.
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Bases:
Every reasonable effort will be made to maintain all safety instrumentation in operation. A startup is not permitted unless three power range neutron instrument channels and two channels each of the following are operable: four reactor coolant temperature instrument channels, four reactor coolant flow instrument channels, l
four reactor coolant pressure instrument channels, four pressure-temperature instrument channels, four flux-imbalance flow instrument channels, four power-number of pumps instrument channels, and four high reactor building pressure instrument channels. The safety features actuation system must have two analog channels functioning correctly prior to startup. EFIC system instrumentation as required by Table 3.5.1-1 must be operable.
There are four reactor protection channels. Normal trip logic is two out of four. Required trip logic for the power range instrumentation channels is two out of three. The EFIC trip logic is two times one-out-of-two taken twice. Minimum trip logic on other instrumentation channels is one out of two.
The EFIC system is designed to automatically initiate AFW when:
- 1. all four RC pumps are tripped,
- 3. The level of either steam generator is low,
- 4. either steam generator pressure is low, or ;
( RCS pressure).
l The EFIC system will isolate main feedwater to any l steam generator when the pressure goes below 600 psig.
The EFIC system is also designed to isolate or feed AFW according to the following logic:
If both SGs are above 600 psig, supply Anf to both SGs.
If one SG is below 600 psig, supply AB7 to the other SG.
If both SGs are below 600 psig but the pressure difference between the two SGs exceeds 100 psig, supply arf only to the SG with the higher pressure.
If both SGs are below 600 psig and the pressure difference is less than 100 psig, supply arf to both SGs.
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At cold shutdown conditions, all EFIC initiate and isolate functions are manually or automatically bypassed. When pressure in both steam generators is greater than 750 psig, the following bypassed initiation signals will have been automatically reset:
- 1) loss of 4 RC pumps, 2) low oteam generator pressure, and 3) low steam generator level.
Since the EFIC receives signals from the RPS, it is ;
important that only corresponding channels be placed in !
" maintenance bypass". If a channel of RPS is in ,
maintenance bypass, the corresponding channel of EFIC l can be bypassed. An interlock feature also prevents bypassing more than one EFIC channel at a time. These interlocking features allow the EFIC system to take a single failure in addition to having one channel in ;
I maintenance bypass.
Various RPS test features can inhibit initiate signals to the EFIC system and degrade the EFIC system below i
acceptable limits if the RPS channel is not in bypass.
Therefore, no testing should be performed on a RPS I instrument string which supplies an output to EFIC l
without placing that RPS channel in bypass.
The EFIC system is designed to allow testing during power operation. The EFIC system can be tested from its input terminals to the actuated device controllers j without placing the channel in key locked " maintenance l bypass." A test of the EFIC trip logic will actuate one l The two relays are I of two relays in the controllers.
tested individually to prevent automatic actuation of the component.
i Each EFIC channel key operated maintenance bypass switch j is provided with alarm and lights to indicate when the maintenance bypass switch is being used. ,
l Discussion:
The existing instrumentation that controls the Auxiliary i Feedwater System has been replaced by the EFIC system instrumentation. Also, the revisions in Proposed Amendment No. 150 are no longer required since that instrumentation has been replaced by EFIC system instrumentation. This specification defines the operating l requirement on the EFIC system.
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- 5. Dcistirn Specifications: I i
3.5.3 h e safety features actuation setpoints and permissible bypasses shall be as.follcws:
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l Functional Unit Action Setpoint 'l I.oss of all RC Pumps Starts Auxiliary Nct Applicable Feedwater Pumps I.aw Feedwater Pressure Starts Auxiliary 2750 psig l Feedwater Pumps l 1
New Specifications: i l
3.5.3 Se safety features actuation setpoints and !
pam4==ible bypasses shall be as follows: i l
Functional Unit Action Setpoint High Reactor 9141 ding Reactor B141 ding spray pressure
- valves *** $30 psig Reactor Building spray pumps *** $30 psig High pressure injection j and start of Reactor '
l B141 ding cooling and l- Reactor 9141 ding isolation. $4 psig Iow pressure injection, EFIC AEW initiate $4 psig Iow reactor coolant High pressure injection system pressure ** and start of Reactor Building cooling and Reactor Build!ng isolation. 21600 psig I.cw pressure injection, EFIC ATU initiate 21600 psig
- May be bypassed during Reactor BH16tng leak rate test.
- May be bypassed below 1850 psig and is automatically reinstated abcVe 1850 psig.
- Five-minute time delay.
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3.5.6 EMERGENCY FEEDWATER INITIATION AND CONTROL SETPOINTS Applicability This specification applies to the emergency feedwater initiation and control (EFIC) setpoints.
Objective To provide for automatic initiation and control of auxiliary feedwater and automatic isolation of main feedwater.
i Specification l Thh emergency feedwater initiation and control setpoints and bypasses shall be as follows:
Functional Unit Action- -Setpoint Initiates AFW >9 inches
- a. Low SG Level
- d. SFAS Actuation Initiates AFW N/A (1) l
Loss Of MFW
- f. Vector Logic Isolates Faulted SG Various (3)
Bypass Permissive <750 psig
- 9. Shutdown Bypass i (1) Refer to Specification 3.5.3 for SFAS setpoint (2) Refer to Table 2.3-1 for RPS setpoint (3) Refer to Bases below for description of vector setpoints 18
1 3.5.6 (continued)
Bases _ i The EFIC system is designed to automatically initiate AFW when:
- 1. all four RC pumps are tripped,
- 3. the level of either steam generator is low,
- 4. either steam generator pressure is low, or
The EFIC system will initiate main feedwater isolation to any steam generator as the pressure goes and stays below a minimum set point of 575 psig. 1 The EFIC system is also designed to isolate or feed AFW according to the 1 following vector logic. Setpoints are nominal and subject to instrument inaccuracies:
- If both SGs are above 600 psig, supply AFW to both SGs
- If one SG is below 600 psig, supply AFW to the other SG j
- If both SGs are below 600 psig but the pressure difference between the two SGs exceeds 100 psig, supply AFW only to the SG with the higher pressure
- If both SGs are below 600 psig and the pressure difference is less than 100 psi.g, supply AFW to both SGs I
!l At cold shutdown conditions all EFIC automatic initiate and isolate functions are manually or automatically bypassed. Prior to a pressure of greater than 750 psig in both steam generators, the following bypassed initiation signals automatically reset: 1) Loss of 4 RC pumps, 2) low steam generator pressure, 3) lou steam generator level.
Bypassing of automatic AFW initiation on Loss of 14FW Anticipatory Trip or SFAS actuation is controlled by bypass permissive logic within tiie RPS and SFAS, respectively.
p_iscussion:
Loss of all RC Pumps and low Feedwater Pressure actuation setpoints are deleted from Specification 3.5.3 since the signals now provide input to initiate the EFIC system. The SFAS actuation signal will result in EFIC AFW initiating. Specification 3.5.6 is added to provide, the applicable EFIC setpoints. Nominal values for EFIC vector logic setpoints are provided in the Bases of Specification 3.5.6.
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- 6. Existirn Specifications:
Table 4.1-1 INSTRUMDTP SURVEILLANCE REQUIREMENTS Channel Description Check Test Calibrate Remarks
- 48. Auxiliary Feedwater Start Circuit l a. Ehase Imbalance /
! UrdeIpower RCP S N/A R
- b. I.cw Main Feed-water Pressure N/A M R NEW SPECIFICATION \
Table 4.1-1 INSTRUMEtTP_SURVEIIIANCE_ REQUIREMENTS l
Channel Description Check Test Calibrate Remarks
- 48. Deleted l
- 68. AFW Initiation
- a. Manual N/l M N/A
- b. Low Invel (1) Include time SGA or B S M (1) R (1) delay module l
- c. Icw Pressure (1) Include time SGA or B S M (1) R (1) delay mcdule j
- d. Loss of MEW Anticipatory Reactor Trip S M R
- e. Ioss of 4 RC 18b.
Table 4.1-1(Continued)
INSTRUMDTP SCRVEIUMCE REvUIRDOTrS Channel Description Check Test Calibrate Remarks
- 69. SGA Main Feedwater Line Isolation
- a. Manual N/A M N/A
- b. Autcmatic Trip l Iogic N/A M N/A l
l 70. SGB Main Feedwater Line Isolation l a. Manual N/A M N/A
- b. Automatic Trip l Isvel N/A M N/A
- 71. AEW Valve Commands (Vector) l
- a. AFW Initiation Automatic Trip logic Tripped N/A M N/A
- b. SGA Pressure (1) Include time Iow S M(1) D(1) delay module
- c. SGB Pressure (1) Include time <
Icw S M(1) D1) delay module j
- d. SG Pressure i Difference (1) Include time SGA Pressure > S D(1) R(1) delay module SGB Pressure SGB Pressure > (1) Include time SGA Pressure S M(1) R(1) delay module
- 72. AEW Control Valve Control
- a. Manual / Auto in Manual N/A M N/A
- b. AFW Initiation Automatic Trip Irgic Tripped N/A M N/A 19
Table 4.1-1 (Continued)
I INSTRLKENT_SURVETIIRG_ REQUIREMENTS Channel Description Check Test Calibrate Remark 1
- 73. SG Imel Control Setpoint Selection l
- a. Manual / Auto in Manual N/A M N/A
- b. AEW Initiation ]
I.cgic Tripped N/A M N/A
- c. I. css of 4 BC Pumps S M N/A l 74. ADV Control Valve Control
- a. Manual / Auto N/A M N/A in Manual l ]
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- 75. Backup Instrument Air SLTply System
- a. Pressure D N/A N/A S = Each shift M = Monthly P = Prior to each starmp if D = Daily Q = Quarterly not done previous week W = Weekly SY = Semiannual R = Once during the refueling interval Discussion:
This change deletes the surveillance requirements for j the existing control functions for the auxiliary I feedwater system and replaces them with the new control functions for the EFIC system. !
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! 7. Existina Specifications:
Table 4.1-2 MINIMUM ECUIRO7I'_ TEST _FRECUENCY Item Test Frequency l
- 6. Turbine steam stop valves Rovement of each valve Monthly Nes Specifications:
Table 4.1-2 MINIMUM ECUIINENT TEST FRECUENCY Item Test Frequency
- 6. Turbine 'Ihrottle Stop Valves Movement Monthly of each valve
- 16. Main Feedwater Isolation Valves
- a. Main Feedwater Isolation Valves- Functional Each refueling interal
- b. Main Feedvater Block Valves Functional Each refueling interval
- c. Startup Feedwater control Valves Functioral Each refueling interval
- d. Main Feetater Control Valves Functional Each refueling interval
- 17. Turbine 'Ihrottle Stop Valves Cycle Each refueling interval
- 18. Backup Instrument Air Supply Functional Each refueling interval System Discussion:
This new specification defines the surveillance requirements included for satisfying the operability requirements of specifications 3.4.2.E, 3.4.2.G, 3.4.2.H, and 3.4.2.I.
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. _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ - _ _ - _ _ _ _ ~
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- 8. Existina Specifications:
4.8.1 Monthly on a staggered test basis at a time ;
when the average reactor coolant system )
temperature is 2 305 degrees"F, the l turbine / motor driven and motor driven !
auxiliary feedwater pumps shall be operated on l recirculation to the condenser to verify proper operation. Separate tests will be i performed in order to verify the turbine l driven capability and the motor driven j l capability of auxiliary feedwater pump P-318. j l The monthly test frequency requirement shall i be brought current within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after the !
average reactor coolant system temperature is 2 305 degrees-F for the motor driven pumps.
The turbine driven capability shall be brought current within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of obtaining 5 percent reactor power. ,
Acceptable performance will be indicated if the pump starts and operates for fifteen minutes at a discharge pressure of greater than or equal to 1050 psig at a flow of greater than or equal to 780 gpm. This flow will be verified using tank level decrease and pump l differential pressure.
1 I 4.8.2 At least one per 18 months during a shutdown: l l \
- 1. Verify that each automatic valve in the I flow path actuates to its correct position upon receipt of each auxiliary feedwater actuation test signal.
- 2. Verify that each auxiliary feedwater pump i starts as designed automatically upon receipt of each auxiliary feedwater ,
l actuation test signal. 1 4.8.3 All valves, including those that are locked, l sealed, or otherwise secured in position, are I
to be inspected monthly to verify they are in the proper position.
4.8.4 Prior to startup following a refueling shutdown or any cold shutdown of longer than 30 days duration, conduct a test to demonstrate that the motor-driven AFW pumps can pump water from the CST to the steam generator.
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l 4.8.5 Provide a dedicated individual during surveillance testing who will be in communication with the control room. This individual shall be stationed near any (locally) manually realigned valves that would inhibit injection into the steam generators, when only one auxiliary feedwater train is available.
l 4.8.6 Component Tests A. Testing At least quarterly, when the average j reactor coolant system temperature is l
greater than or equal to 305 degrees-F, inservice testing of Auxiliary Feedwater l System pumps and valves shall be performed in accordance with Section XI of the ASME Boiler and Pressure Code and applicable Addenda as required by 10 CFR 50, Section 50.55a (g) (6) (i) .
The quarterly test requirement shall be brought current within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after the average Reactor Coolant System temperature is greater than or equal to 305 degrees-F.
i B. Flow Path Verification Following inservice testing of pumps and valves as required by paragraphs 4.8.1 and 4.8.2, required flow paths shall be demonstrated operable by verifying that each valve (manual, power-actuated or automatic) in the flow path that is not locked in position is in its normal operating position.
Bases:
The monthly test frequency will be sufficient to verify that the turbine / motor driven auxiliary feedwater pumps are operable. Verification of correct operation will be made both from the control room instrumentation and direct visual observation of the pumps.
The OPERABILITY of the auxiliary feedwater system ensures that the Reactor Coolant System can be cooled down to less than 305 degrees-F from normal operating conditions in the event of a total loss of off-site power.
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l Each electric driven auxiliary feedwater pump is capable of delivering a total feedwater flow of 780 gpm at a pressure of 1050 psig to the entrance of the steam generators. The steam driven auxiliary feedwater pump is capable of delivering a total feedwater flow of 780 gpm at a pressure of 1050 psig to the entrance of the steam generators. This capacity is sufficient to ensure that adequate feedwater flow is available to remove l decay beat and reduce the Reactor Coolant System l temperature to less than 300 degrees-F when the Decay 1
l l Heat Removal System may be placed into operation. j l
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New Specifications,1 4.8.1 Monthly on a staggered basis at a time when ;
! the average reactor coolant system temperature l l is 2305 degrees-F, the turbine / motor driven and motor driven auxiliary feedwater pumps shall be operated on recirculation to the condenser to verify proper operation.
Separate tests will be performed in order to verify the turbine driven capability and the motor driven capability of auxiliary feedwater pump P-318.
The monthly test frequency requirement shall l be brought current within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after the l average reactor coolant system temperature is i l 2305 degrees-F for the motor driven pumps.
The turbine driven capability shall be brought i current within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of obtaining 5 percent reactor power.
Acceptable performance will be indicated if the pump starts and operates for fifteen minutes at a flow rate sufficient to assure 475 gpm of flow to the steam' generator at a discharge pressure sufficient to drive that flow through the most restrictive flow path to a single steam generator which is at a pressure of 1050 psig.
The monthly testing of the auxiliary feedwater pum and valves shall be performed in accordance with t inservice inspection requirements of Specification 4.2.2.1.
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4.8.2 At least once per 18 monthst
- 1. verify that each automatic valve in the l flow path actuates to its correct position upon receipt of each auxiliary feedwater actuation test signal.
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- 2. Verify that each auxiliary feedwater pump starts as designed automatically upon receipt of each auxiliary feedwater actuation test. signal.
4.8.3 All auxiliary feedwater system valves, including those that are locked, sealed, or otherwise secured in position, are to be inspected to verify they are in the proper '
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position following surveillance performed l pursuant to Specifications 4.8.1, 4.8.2 and 4.8.4. j 4.8.4 Prior to startup following a refueling 4 shutdown or any cold shutdown of longer than I 30 days duration, conduct a test to demonstrate that the motor-driven AFW pumps can pump water from the CST to the steam generator.
Bases:
l The monthly test frequency will be sufficient to verify that the turbine / motor driven and motor driven auxiliary l feedwater pumps are operable. Verification of correct operation will be made both from the control room instrun.entation and direct visual observation of the .
pumps. !
The OPERABILITY of the auxiliary feedwater system ensures that the Reactor Coolant System can be cooled 3 down to less than 305 degrees-F from normal operating (
conditions in the event of a total loss of off-site power. ,
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The electric driven auxiliary feedwater pumps are capable of delivering a total feedwater flow of 475 gpm at a pressure of 1050 psig to the entrance of the steam generators. The steam driven auxiliary feedwater pump a is capable of delivering a total feedwater flow of 475 l gpm to the entrance of the steam generators over the steam generator operating range of 800 psig to 1050 psig. This capacity is utilized as analytical input to the Loss of Main Feadwater Analysis which is the design l basis event for AFW flow requirements. j 1
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Discussion:
This revision incorporates a reduced auxiliary feedwater flow rate of 475 gpm and 2) Proposed Amendment No. 148 which would permit auxiliary feedwater periodic testing either during plant operation or shutdown conditions. The definition of acceptable 9rformance for the auxiliary feedwater pumps has ~een clarified to better define the pressure requirement associated with the 475 gpm flow. The prescription that testing of AFW pump and valves is j performed in accordance with Specification 4.2.2.1 l requirement is also added. Specification 4.8.5 was a l
deleted since the AFW flow test valves can now be automatically operated from the control room, an individual is no lenger required to be stationed at the valves during surveillance testing. Also, existing l Specification 4.8.6 was deleted. Since the NRC issued Amendment 80 dated April 14, 1986 requiring monthly testing, the NRC also left in the requirement for the quarterly testing of the same pumps.
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Basis for No Significant Hazards Determination The proposed change does not involve a significant hazards consideration because operation of Rancho Seco in accordance with this change would not:
- 1. Involve a significant increase in the probability or consequences of an accident previously evaluated. The l September 26, 1983 NRC SER documents the acceptability of the design basis for the EFIC system.
- 2. Create the possibility of a new or different kind of accident from any previously analyzed.
l The transient analyses performed demonstrates that sufficient auxiliary feedwater will be provided for all transients which require auxiliary feedwater for mitigation.
- 3. Involve a significant reduction in a margin of safety.
The revised design basis event for sizing the AFWS is a loss of main feedwater with'no loss of offsite power.
The reactor trip with no loss of offsite power and reactor trip due to loss of offsite power represent the limiting conditions for determining AFWS flow requirements. Our transient analysis documents.the acceptability of using the reduced design basis flow.
This reduction in required auxiliary feedwater flow does not represent a significant reduction in a margin of safety,in that, this flow reduction significantly reduces the potential of. steam generator structural damage due to excessive auxiliary feedwater flows at low steam generator pressures while continuing to ensure sufficient auxiliary feedwater for all accident scenerios.
( _ _ __- ____ - _ _ _ -
ENCLOSURE 2 Proposed Technical Specifications Amendment No. 152 l
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