ML20248B886
| ML20248B886 | |
| Person / Time | |
|---|---|
| Site: | 05000605 |
| Issue date: | 08/02/1989 |
| From: | Marriott P GENERAL ELECTRIC CO. |
| To: | Chris Miller NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), Office of Nuclear Reactor Regulation |
| References | |
| 057-89, 57-89, NUDOCS 8908090359 | |
| Download: ML20248B886 (22) | |
Text
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e e GE Nuclear Energy N '[ ,' ~~ + ') .
August 2,1989 MFN No. 057-89 Docket No. STN 50-605
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Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Attention: Charles L. Miller, Director Standardization and Non Power Reactor Project Directorate
Subject:
Submittal of Responses to Additional Information as Requested in NRC I4tter from Dino C. Scaletti, Dated May 16,1989
Dear Mr. Miller:
Enclosed are thirty four (34) copies of further responses to the subject Request for Additional Information (RAI) on the Standard Safety Analysis Report (SSAR) for the Advanced Boiling Water Reactor (ABWR). These responses pertain to Chapters 7 and 8.
It is intended that GE will amend the SSAR with these responses in a future amendment.
Sincerely,
/
'k P. W. Marriott, Manager Licensing and Consulting Services cc: D.R.Wilkins (GE)
F. A. Ross (DOE) ,
J.F. Quirk (GE)
D. C. Scaletti (NRC)
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L 8908090359 890302 PDR ADOCK 05000605 .
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CHAPTER 7 RESPONSES i i
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QUESTION j 420.002 (7) Identify the topical reports that will be provided to support any j aspects of the design that are substantially different relative to i designs previously reviewed by the staff. Subjects addressed in these topical reports should include but not necessarily be limited to the following: 1 The applicant's.overall design verification program, covering . )
development of the functional requirements, criteria, specifications, i design, manufacture, test, and qualification methods and procedures; this should include a V6V plan for software design 3 verification / validation. l i
RESPONSE
420.002 In response to this question, refer to Appendin '7A, Section 7A 7, under ]
the heading: " Items 7A.5(1) and 7A.5(2)". Detail information may be j found in the ABWR Design Specifications referenced in Section 1.1.3. j l
QUESTION 420.011 (7.6.1.1) Identify the topical reports that will be provided to support i any aspects of the design that are substantially different relative to !
designs previously reviewed by the staff. Subjects addressed in these topical reports should include but not necessarily be limited to the i following:
Wide Range Neutron Monitor design basis. (NEDO.31439, May 1987) If this system is not part of the ABWR (Section 7.6.1.1 indicates it is not) provide justification for its exclusion.
RESPONSE
420.011 In the ABWR, the system " Wide Range Neutron Monitor" is implemented and is named the "Startup Range Neutron Monitor" abbreviated as "SRNM". ,
With reference to 7.6.1.1, the "SRNM" subsystem design description can l be found in the topical report "The Nuclear Measurement Analysis &
Control Wide Range Neutron Monitoring System (NUMAC-WRNMS)", NEDO 31439, May 1987. However, this topical report does not contain ABWR plant. specific design parameters, such as total number and locations of SRNM used in the reactor core, SRNM trip setpoints, etc. -Such plant specific information can be found in the NMS design specification (see Section 1.1.3).
i QUESTION '
420.012 (7.4.2.2.2) Identify the topical reports that will be provided to ;
support any aspects of the design that are substantially different relative to designs previously reviewed by the staff. Subjects addressed in these topical reports should include but not necessarily be limited to the following:
10CFR50.62 (ATUS) conform &nce. Specifically address the manually initiated SLCS conformance (7.4.2.2.2(1)) to the ATWS rule (50.62(4)) of' automatic initiation.
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.. c RESPONSE H
'l 420.012 No special. topical reports will be issued to address'the ABWR ATWS c,.
' design. The conformance of 10CFR50.62 for the ABWR design is discussed in Section 15.8 of the SSAR. Additional discussions are provided in the responses to IntC Questions 440.103,1440.104 and 440.115. 1 1
.')
QUESTION. .
. .l 420.014 (7.1.2.3.9) Address the effects ofLStation Blackout on the.HVAC required to maintain functional electronics.
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i RESPONSE-420.014 No HVAC systems are operational during Station. Blackout. Preliminary
-analysis has shown that-the limited amount of equipment available:and used during Station Blackout reduces the heat generated to a point where temperatures remain within the qualified limits;of the operating- .. . .i equipment. It is an interface ~ requirement for the applicant to perform i a temperature heat rise analysis for the Station Blackout scenerio applied to the control room in consideration of the environmental' i temperatures unique to the plant, location. j i
QUESTION 420.015 (7.4) Address the redundancy and diversity of the power supplies for ARI.
RESPONSE
420.015 Section 7.4.1.1 has been revised as shown in the attached mark-up of-that section of the SSAR. Currently, the ARI function is implemented both hydraulically and electromechanically. The new design, in addition to the usage of the "FMCRD Run.In" function (i.e., electromechanical), ,
also includes the ARI valves 1(i.e., hydraulic) that can cause reactor. f shut down, independently and diversely from the Reactor Protection l System. i The design of the ARI function has therefore been expande'd to: meet: all !
of the requirements of GE's Licensing Topical Report'(LTR)
NEDE-31906-P.A, titled " RESPONSE TO NRC ATWS RULE,' 10CFR.50.62".
Redundancy and diversity of power supplies are also addressed in this :
UTR.
i QUESTION
( 420.016 (7 4) Address the decision to make the AR1 non-1E instead of 1E system.
RESPONSE
420.016 The question of non.1E vs 1E has also been addressed in the GE LTR NEDE 31906-P-A. The design of the ARI function meets the requirements of this LTR, which is endorsed by the ABWR Standard Plant.
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QUESTION 420.019 (7.1) The submittal describes an intelligent multiplexed digital system as the implementation for the logic of the safety system. Figure 7.1-1 shows a system that is highly interconnected. Show how this l interconnection satisfies the independence criteria in accordance with I IEEE Std 603 and IEEE Std 379.
RESPONSE
420.019 Figure 7.1-1 was intended to show only the divisional interfaces for the self-test subsystem of the SSLC, not the SSLC logic itself. However, the figure has been revised (per attached) in accordance with the design change which eliminated the on-line interconnecting concept for the self-test subsystem.
The SSLC Iogic has inter-divisional fiber-optic links to facilitate the 2/4 coincident voting capability. However, such links are unidirectional and their only failure mechanism is an erroneous logic signal to the voting processor. The remaining channels would revert to 1/3 (unbypassed) or 2/3 voting depending on the state of the logical failure. This is the same affect as any other failure within a given channel and is consistant with the single failure criteria defined in IEEE Standards IEEE 603 and IEEE 379.
QUESTION 420.021 (7) Describe the safety computer system's interface to any non-safety computer systems and other plant instrumentation. Describe if ;
information transfer from 1E to N.1E computers is via broadcast or handshake.
RESPONSE
420.021 The ABWR does not use a central " safety computer" to initiate any safety function. Individual Class 1E microprocessors are used in place of CMOS (CESSAR II design) in the logic. The important distinction is that the ABWR uses a modern form of digital computer device (i.e.,
microprocessors) for the same reasons relays and solid. state devices were used in earlier designs (i.e., making simple logic decir, ions); not for making complex calculations for which protective cetion is dependent, Interactions between electrical divisions and with non.1E functions are performed via fiber-optic cable. Thus, electrical independence is maintained. Information transfer from 1E to non-1E devices is via broadcast only.
QUESTION 420.023 (7) Provide a table of conformance to IEEE 384, indicating where credit is taken for isolstion or separation, what devices or methods are used, and the basis of isolation device qualification. If specific types of components have not been chosen, provide specification level information including testing acceptance criteria.
RESPONSE
.f 420.023 IEEE 384 is addressed in Table 7.1-2, as endorsed by. Regulatory Guide
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' 1'. 7 5.. Since'the requirements of this guide envelope and endorse-IEEE.
] 1 384, it is not necessary to discuss IEEE 384 separately. Individual -
systems analysis sections discuss,the' degree of conformance, and ;
exceptions (if any).
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In general,. electrical isolation is accomplished through the use of :]
fiber-optic cable.
d QUESTION ,
l 420.036 (7.1.2.6.6) If the CAMS system is only a monitoring system; why is it l not always on instead'of waiting,for a LOCA.to monitor radiation?
RESPONSE
420.036 The CAMS radiation monitoring subsystem; continuously monitors the-total j gamma dose rate in both the drywell and the suppression chamber during normal plant operation, shutdown and, accident.(LOCA)-conditions. This: '
subsystem is manually turned on during normal plant operation for I
continuous monitoring. In the event that-this subsystem has been turned off for any reason, the LOCA signal (high drywell pressure or low j reactor water level L1) will. automatically turn it.back.on. CAMS '
radiation monitoring subsystem is always continuously monitoring for .
radiation during all plant conditions. See subsection 7.6.1.6 for more j information. j I
f QUESTION 1
420.042 (7.3.1.1.1.1) Section 7.3.1.1.1.1(3)(f) states that separation. prevents s a single design basis event from disabling core cooling. 'This section 1 should note that this event must be considered in conjunction with"an i additional single failure. i 1
RESPONSE :!
420.042 Section 7.3.1.1.1.1(3)(f) has been revised to read: " Separation within~
the emergevey core cooling system is such that no single design basis ~l event, in tunjunction with an additional single failure, can prevent 1 core cooling when required."' i
.............................................................................. i QUESTION 420.058 (7) Beyond the redundancy requirements levied by single failure i criteria, provide information to demonstrate sufficient diversity in the .l I&C system to preclude common mode failures. '
RESPONSE
420.058 In response to this question, refer to Appendix 7A, Section 7A.7 under.
the heading: " Items'7A.5(4) and 7A.6(4)",
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l QUESTI04 !
420.080 (7.1.2.3.1; dettion 7.1.2.3.1(1)(c) states that no operator action is j reqaired for 10 mi.tutes following LOCA. Section 6.3.1.1.1(3) states, '
that no operator action is required for 30 minutes after an accident.
Section 6.3.2.8 also states 30 minutes. Clarify which statement is the 1 design basis. Same question @ 7.3.1.1.1.4(3)(i) and 7.3.1.1.1.2(3)(1).
RESPONSE 'i 420,080 Sections 7.1.2.3.1(1)(c), 7.3.1.1.1.2(3)(i), and 7.3.1.1.1.4(3)(i) cf j the SSAR have been revised from 10 minutes to 30 minutes.
i QUESTION j 420.081 (7.1.2.3.1) itation 7.1.2.3.1(1)(c) states that operator action is not i
required. 1.recibe what operator actions are desired but not required for the- f! ret period of time (10 or 30 minutec) for various accident scenarios.
RESPONSE .I' 420.081 The 30 minutes that no operator action is required is a backup to expected operator actions performed normally in response to encountered j events. Normally an operator would take actions in accordance~with the symptom based Emergency Operating Procedures.
The following discussion outlines the expected operator response to the I
accidents identified, '
For this discussion, accident scenarios can be classed in the general ca egories of large breaks or medium /small treaks inside' containment or breaks outside containment.
lARGE BREAKS INSIDE CONTAINMENT bo drywell pressure will rise rapidly and all ECC5 systems will automatically initiate on high drywell pressure within a few seconds, including diesel generator automatic startup. RCIC and HPCF begin j injection without reactor pressure vessel (RPV) depressurization. The RHR initiates by starting the pump, but RHR injection requires that the. ;
RPV be depressurized to below a specified pressure. The operator would confirm these events, and attempt manual starts for any subsystem that l did not automatically start.
With a large break the RPV will depressurize rapidly, allowing the RH2 in the. flood' 'g mode to automatically inject 'into the RPV along with' the 1 RCIC and HFc . The operator should monitor reactor water level and primary containment parameters such as suppression pool water level and '
temperature, drywell pressure and temperature, and containment pressure and temperature. If reactor water level increases as a result of ECCS injection, RCIC and HPCF injections will automatically shutoff when RPV.
1evel increases to level 8 if no optrator act'on is taken. RHR does not '
l have an automatic shutoff feature or a level 8 signal. To prevent filling tto stermlines,-the cperato should stop some or all RHR injection prior to RPV water level r. aching level 8. .!
Once a RHR loop is not required for RPV level control, it can be applied to other tasks. These tasks can be direct suppression pool cooling, ;
drywell pressure / temperature' control (drywell spray), containment j pressure / temperature control (wetwell spray), and-suppression pool level.
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control, as directed by the symptom-b: sed Emergency Operating Procedures.
MEDIUM /SMALL BREAKS INSIDE CONTAINMENT High drywell pressure will increase rapidly, but slower than for large breaks, and RCIC, HPCF and RHR will automatically initiate on high drywell pressure. RCIC and.HPCF injection into the RPV begins without
- RPV depressurization, but for RHR, only the pump starts for inittation.
The operetor must confirm these high pressure systems are operating and try manual starts if they are not.
The operator must monitor the reactor water level to determine if the RCIC and HPCF can maintain or increase the reactor water level on their own. If the water level is decreasing and the operator does nothing, ADS valves automatically open to depressurize the RPV at level 1-(one) so the RHR can begin flooding. The operator should control RPV pressure and level in accordance with the symptom-based Emer5ency Operating Procedures to avoid automatic ADS initiation.
If reactor water level can not be maintatcsd with RCIC and HPCF, the operator r,hould use the SRVs to depressurize the RPV to a pressure where RHR can ;uject water into the RPV. As the RPV depressurizes, the HPCF will increase its flow rate which will slow or stabilize the dropping reactor water level. If the reactor water level continues to drop, the operator can increase the depressurization rate by opening more SRVs to permit low pressure RHR flooding injection-Once a RHR loop is not required for reactor water level c on m it can be applied to the same tasks identifted for the large break description above. If the reactor water level should begin to rise, the operator should anticipate the reflooding possibility and respond as described in j the large break description above. '
If the teactor water level begins to increase af ter the initial Figh pressure ECCS response occurs, then the operator should stop flooding the RPV before the level 8 is reached to prevent filling the steamlines.
Even to a lessor degree than for a large break, the drywell or wetwell sprays are not required during the first 30 minutes for medium /small breaks. If the RHR system is not required to maintain reactor water level, it can be used to control suppression' pool water level, suppression pool tNr.3 rature, and containment pressure in accordance with the symptom er'ad EOPs.
BREAKS OUTSIDE CONTAINMENT These breaks a;e detected by the Leak Detection and Isolation System and initiate various isolation responses and potentially a scram, depending on the established logic combinations for the many monitored signals.
The drywell pressure will not increase to initiate the ECCS. For the cases where the leak detection and isolation system isolated the MSIVs, the decay heat will be vented to the suppression pool through the SRVs.
Thus the event ia reduced to a transient " isolation" event. If the level control system can not automatically makeup the reactor water level, the reactor water level will decrease due to steam venting to tne suppression pool. The operator can manually initiate RCIC and HPCF, as required, to maintain reactor water level in accordance with the symptom-based Emergency Operating Procedures. As the RHR system is not required for RPV makeup for an isolatable break outside the containment i
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QUESTION 420.'0 87 (response 440.113) The response noted that RIP trips have mostly been caused by noisc in the adjustable speed drive (ASD). Describe the changes that have been made to reduce the susceptibility of the RIP's or the reduction in noise of the ASD's.
RESPONSE
420.087 The primary cause of faulty trips of Adjustable Speed Drives -(ASD) in the European plants with Reactor Internal Pumps (RIPS) is due to electromagnetic interference (EMI) and harmonic effects on the improperly installed ASD power and control cables. Several preventive changes have been made to the ASD design and installation requirements to reduce their occurrence. These changes can be summarized as below:
(1) Interconnection cables within the ASD are designed with considerate-on for EKI effects. Factory tests will be performed on tho ASD equipment to demonstrate that EMI effects are within the tolerance limits of the ASD control functions. Proper isolation between the power and the control cables is specified to eliminate propagation of EMI noises. Both the ASD equipment requirements specifications and the installation specification reflect the requirements of the "Special Wire and Cable Specification" (see Sec tion 1.1.3) for interconnection of static converter device.
(2) The ASD is designed to ensure that current and voltage distortions due to harmonics on the output pr wer waveform is tolerable by both the l connected lo .js and the input po ier distribution system. Phase-shifted l isolation t) msformers are utilized to reduce harmonic contents on the power supply circuits. These specifications have been defined as interface requirements on the affected equipment designs. !
...................................................... ..... .... ,,,,,,,,,,,, 1 QUESTION 420.096 (15A.6) The safety system auxiliaries (Figure 15A.6 1) shculd be mootfied to include any HVAC required to assure continued operation of the electronics.
RESPONSE
420.096 The electronics for safety syst. ems are located in either equipment rooms or the control room. The cooling of safety systems equipment rooms is already shown in Revision B of Fi Eure 15A.6 1. The boxes titled
" REACTOR BUILDING COOLING WATER SYSTEM" appear in 5 places on the figure. They are accompanied with notes on the right designating the i
areas which they cool. The HVAC system for the control room is also shown in Revisian B of Figure 15A.6 2 (top center).
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,' QUESTION . .
l 420.105 (7.1) The current criteria.for ATWS' capabilities is the NRC'ATWS Rule 't 10CFR50.62. ~ The existing EWR plant designs have been provided with a.
Safety Evaluation of the Topical Report (NEDE-31096.P) which contains an Appendix A " Checklist for Plant Specific Review of Alternate Rod Injection System (ARI). No topical reference was found in the 4 submittal. j Indicate if this checklist is applicable to this design and how the l compliance to the'ATWS rule is to be achieved.
RESPONSE
420.105 The Appendix A." Checklist for Plant Specific. Review of Alternate Rod' .
Injection (ARI)" contained in topical report NEDE-31096.P-A (Response to. ,
-NRC ATWS Rule, 10DFR50.62) is applicable to the current ARI function, j It was used,as a guide in the development of the design. 1 As Discussed in the response to Question 420.15, the ARI design has been-modified since:the previous submittal'to incorporate'the ARI valves inL compliance with the topical report.
QUESTION 420.118 (15.2.4.5.1) Describe when appropriate operator action c. eeconds is required to prevent significant radiological impact.
RESPONSE
420.118 The reference to "... operator action (seconds).." in:15.2.4.5.1(1) was 3 is in error and has been corrected per attached mark.up'of this {
subsection. 1 With respect to offsite radiological impact, there are no operator actions required within " seconds" to prevent significant radiological impact. Operator actions for transients are normally based upon (1)' l returning the plant to a normal condition or (2) are'taken to prevent- i damage to the plant. Those actions necessary to. prevent.significant. j radiological impact are automated as part.of the normal' plant safety j systems. ;
1 QUESTION i 420.120 (7.3.2.1.2(3)(c)) List all exemptions to the requirement rathe'r than l providing an example. '
RESPONSE
420.120 The list of specific devices which cannot.be fully operated for test during plant operation, or tested by other than continuity tests without. ,
degrading plant operability or safety,l includes (but- is not necessarily '
limited to) the following: i
- 1. SRV solenoid pilot valves i
- 2. HP/LP. interlocked valves'for RHR
- 4. SLC injection valves
- 5. Boron tank suction valves ,
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I QUESTION 420.122 (15.2.2.2.1.4) Is the instrumentation required for the operator to verify bypass valve performance and relief valve operation IE or N-1E?
RESPONSE
420.122 Bypass valve position sensing equipment is all non-1E.
The safety relief valves are designed with two methods of detecting their position. Linear variable differential transformers (LVDTs) are the primary sensors. These are mounted on the valves, and qualified along with the valves themselves as safety-related devices. Signals from the LVDTs are transmitted through the MUX interface to the annunciator and computer. Both of these devices are non-1E. However, SRV position is identified by Regulatory Guide 1.97 as a post-accident monitoring parameter, and is designated as Type D, Category 2, in Table !
i 7.5-2. Thus, display instrumentation is provided in accordance with Regulatory Guide 1.97 for SRV position.
The other method of detecting SRV position (or leaking) is accomplished 1 with temperature sensors in the tailpipes for each of the valves. l Signals are routed to a common temperature recorder which continuously l i
cyc1es through the full set of SRVs and records each's temperature on a moving chart. This instrumentation is all non-1E. The SRVs and associated instrumentation are shown on the nuclear boiler system P&ID, Figure 5.1-3. I In addition to the two methods described above, the Class-1E suppressiu.
pool temperature monitoring (SPTM) system provides an indirect methol of detecting relief valve activity. The SPTM is described in Sections 7.6.1.7 and 7.6.2.7. l j
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c, 23A6100AF REV A !
.. Standard Plant s 7.4 SYSTEMS REQUIRED FOR SAFE injection of " neutron absorber into SHUTDOWN the reae '# ally called upon to do
<*# 8 'rovides the neces- I
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7.4.1 Description r #
0 $ tion tempera.
f's gj The system This section eramines and discusses the in- *p#*/ *# 8 e/ /o## t The inter-strumentation and control aspects of the f[-/
lowing plant systems and functions desir assure safe and orderly shutdown of the ' g M +g $ p**'
/ #/*p. / c*je g in Figure (1) Alternate rod insertion functio [Aj '
o' y < #/ / g (2) Standbyliquid controlsys
- 8 g manually shut f .# ' #,/ /,.'
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(3) Reactor shutdown cooling .
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/gj[/ to p/,o by the control rod {
(4) Remote shutdown system (RSS, 4' 4.+ # *
/ ..*jt o , system is considered a 1 ystem.' The standby liquid 5 g/,/,[/ / ,s0ho See Subsection 7.1.2.4 which addre ,
g .,ss equipment, instrumentation, sign basis information required by St #p# */j t .ols essential for injection of the a absorber solution into the reactor IEEE 279. ogo'p designed to withstand Seismic Category I p earthquake loads. Any nondirect process 7.4.1.1 Alternate Rod Insertion Function-Instrumentation and Controls equipment, instrumentation, and controls of
@ 1;ruRI16 G the system are not required to meet Seismic Category I requirements; however, the local The alternate accomplished by the r rod insertion control and (AK)informatio and control room mounted equipment is system'(RC&IS) a he fine-m ion control d located in seismically qualified panels.
system. Th' function pr ide.
drive'(FMCRD) an' alternate me [ hod of drivifIg control r(3) - sPower intcSources Ihe core w h is diver , from the ydraulic ram sysi m.- The power supply to one motor-operated injec-tion valve, storage tank discharge valve, inc/ dst 1 The RC&lS(,.uhthgetige 4 rgng, unction and of injection pump is powered from Division the FMCRD motors [re not7equired Ior safety, nor 1,480VAC. The power supply to the other are these compone'nts qualified in accordance with motor operated injection valve, storage tank.
i safety criteria. However, the FMCRD components outlet valve, and injection pump is powered associated with hydraulic scram are qualified in from Division 11,480VAC. The power supply accordance with safety criteria. to the tank heaters and heater controls is el M connectable to a standby AC power source.
The strhrystem inherent diversitygrovides The standby power source is Class 1E from an mitigation of the conse sences of ATWS (ant ici- onsite source and is independent of the p.aled transient withouj scraf de- offsite power. The power supply to the main s-i control room benchboard indicator lights and
[( taisecnfin ed .7.7.d.2.
desefiption 1offfM&jng.__/._give -
in the level and pressure sensors is powered from a Class IE instrument bus.
7.4.1.2 Standby Liquid Control System-Instrumentation and Controls (4) Equipment The SLCS is a special plant capability event (1) Function system. No single active component failure The instrumentation and controls for the of any plant system or component would SLCS are designed to initiate and continue necessitate the need for the operational 7A 1 Amendment 2
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ABM p-ge' 23A6100AF REV.A
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, Standard Plant m The HPCF pump motors and injection valv- divisions end, if'necessary, th'e ADS and es are provided with manual override low pressure flooder mode of the R'HR.
controls which permit the operator man- The locked-out loop c:.n be manually re-ual control of the system following a started by unlocking the switch and LOCA. placing in t'ac " start" position.
During test operation, the HPCF pump. (e) Actuated Devices discharge is routed to the suppression .
pool. Two motor-operated valves are in- All motor-operated valves in the HPCF stalled in the test lines for each system are equipped with remote manual loop. The piping arrangement is shown functional test feature. The entire in Figure 6.31. The control scheme for system can be manually operated from the the valves is shown in Figure 7.3-1. On main control room.
receipt of an HPCF ' initiation signal,
.the test line valves close and remain' Motor-operated valves are provided with closed. limit switches to turn off the motor when the full open or closed positions The HPCF pump is interlocked with a are reached. Torque switches also corresponding bus undervoltage monitor. control valve motor forces while the i The pump motor circuit breaker will not valves are seating.
close unless the voltage on the bus supplying the motor is above the set The HPCF valves must be opened suffi-point of the undervoltage m ' *
- or, ciently to provide design flow rate within 36 seconds from receipt of the (d) Redundancy and Diversity initiation signal.
The HPCF is actuated by reactor vessel The HPCF pump discharge line is provided i low water level (Level 1.5) or drywell with an AC motor operated injection l high pressure. Both of these conditions valve. The control scheme for this may result from a design basis loss-of. valve is shown in Figure 7.3-1. The 1 coolant accident, valve opens on receipt of the HPCF l initiation signal. The pump injection The HPCF system logic requires any two valve closes automatically on receipt of
~
j of the four independent reactor vessel a reactor high water level (Level 8) water level measurements to concurrently signal.
indicate the high water level (Level 8) condition. When the high water level Two pressure transtr.itters and associated condition is reached following HPCF control room. interfaces are installed in operation, these two signals are used to each pump discharge pipeline to verify stop HPCF flow to the reactor vessel by that pumps are operating following an ;
closing the injection valve. However, initiation signal. The pressure signals the pump continues to run unless delibe- are used in the automatic depressuriza-rately stopped by the operator with the tion system to verify availability of pull to-lock switch. Should the low high pressure core cooling.
water level (Level 1.5) condition recur, the injection valve will reopen automa- f )! >./.i ( %
ff Separation 4 2 d , O P M ,,
,i i tically. This action will restore water level within the reactor unless the ope. Separation within the emergency core ;
l u rator has used the pull to lock stop of l
the pump motor due to HPCF loop failure -design cooling basissystem.,,ig%ghgat eve can prevent coreog ,s,ingp,,
l (i.e., ruptured injection line, etc). cooling when required. Control and ele-I In that event, adequate water level is ctrically driven equipment wiring is se- 3 assured with the redundant HPCF and RCIC gregated into three separate electrical
- l.3-3 Am:ndmer.t 2
ABM 23A61%AF Standard Plant REV.A sients, or physical events from impair- Specific Regulatory Requirements: ,
ing the ability of the system to respond '
correctly. The specific requirements applicable to the RPS instrumentation and control tre shown in (k) Earthquake ground motions, as amplified Table 7.1-2.
by building and supporting structures, shall themselves initiate reactor scram, (2) Nont.afety-Related Design Bases and shall not impair the ability of the RPS to otherwise initiate a reactor The RPS is designed with the added objective scram, with the exception of turbine of plant availability. The setpoints, power building trips which originate from a sources, and control and instrumentation non seismic building. These shall be shall be arranged in such a manner as to backed up by diverse variables such as preclude spurious scrams insofar as reactor pressure and power trips. practicable and safe.
(1) No single failure within the RPS shall 7.1.23 Engineered Safety Features (ESF) prevent proper reactor protection system action when required to satisfy Safety 7.1.23.1 Emergency Core Cooling Systems Design Bases as described by the first Instrumentation and Controls three bullets under 1(a) above.
(1) Safety Design Bases (m) Any one intentional bypass, maintenance operation, calibration operation, or General Functional Requirements:
test shalltonot75a verifg::perational availability
- e the ability of the The ECCS control and instrumentation shall spo,108 reactor protection system to respond be designed to meet the following correctly. requirements:
(n) The system shall be designed so that two (a) automatically initiate and control the or more sensors for any monitored emergency core cooling systems to variable exceeding the scram setpoint prevent fuel cladding temperatures from will initiate an automatic scram. reaching the limits of 10CFR50.46.
The following bases reduce the probabi- (b) respond to a need for emergency core lity that RPS operational reliability cooling regardless of the physical and precision will be degraded by location of the malfunction or break operator error: that causes the need; l (o) Access to trip settings, component cali- (c) limit dependence on operator judgement l bration controls, test points, and other in times of stress by:
l terminal points shall be under the con- g C- -
~
N trol of plant operations supervisory automatic response of the ECCS so that personnel, no ac required of plant operators "
(p) Manual bypass of instrumentation and with xcoola minutes after a loss of-ccident; d ,
control equipment components shall be ~ -
under the control of the control room indication of performance of the ECCS by operator. If the ability to trip some main control room instrurr.entation; and essential part of the system has been bypassed, this fact shall be continuous- provision for manual control of the ECCS ly annunciated in the main control room. in the main control room.
Amendment 2 7.1 -6
ABM 23A6100AF Standard Plant REV A (2) Manual relationship to each other. Indication ,-
for each instrument channel is available (
Manual action by the operator on displays associated with the SSLC.
(either by ADS system level The logic is tested congruously by l actuation, or by individual SRV automatic self test circuits. The STS, I operating switches); the sixth test, discussed in RPS testa-bility (7.1.2.1.6) is also applicable (3) Pressure Relief Action bere for ADS. The instrument channels are automatically verified every ten Pressure transmitter signals above minutes as explained in that section.
setpoints as a result of high Testing of ADS does not interfere with reactor pressure (see Paragraph automatic operation if required by an (4)); or initiation signal. The pilot solenoid valves can be tested when the reactor is (4) Safety / Relief Action not pressurized.
Mechanical actuation as a result of (h) Environmental Considerations high reactor pressure (higher than pressure in item 3). The signal cables, solenoid valves, safety / relief valve operators and (f) Separation accumulators, and RV low water level instrument lines are the only essential Separation of the ADS is in accordance control and instrumentation equipment with criteria stated in Section 7.1. for the ADS located inside the drywell.
ADS is Division I (ADS 1) and Division These items will operate in the most 11 (ADS 2) system except that only one severe environment resulting from a ~
set of relief valves in supplied. Each design basis loss-of coolant accident ,
ADS relief valve can be actuated by any (Section 3.11). Gamma and neutron one of three solenoid pilot valves radiation is also considered in the supplying nitrogen gas to the relief selection of these items. Equipment valve gas piston operators. One of the located outside the drywell (viz., the ADS solenoid pilot valves is operated by RV level and DW pressure transmitt:rs Division I logic and the other by and multiplex interfaces) will also lhvision 11 logic. The third solenoid operate in their normal and accident pilot is used for nonsafety relief valve environments.
operation. Control logic manual controls and instrumentation are mounted (i) OperationalConsiderations se that Division I and Division 11 separation is maintained. Separation The instrumentation and controls of the from Divisions 111 and IV is likewise ADS are not required for normal plant maintained. operations. When automatic depressuri-zation is required, it will be initiated (g) Testability automatically by the circuits described in this section. No _ op r action is i ADS has two complete contrc,1 logics, one required for at lea minutes in Division I and one in Division 11. following initiation o e system. l@ 0-0 %
Each control logic has two circuits, both of which must operate to initiate A temperature element is installed on ADS. One circuit contains time delay the safety / relief valve discharge piping logic to give HPCF an opportunity to several feet from the valve body. The start. The ADS instrunnent channels temperature element provides input to a signals are verified by cross comparison multipoint recorder and interfaces with l between the channels which bear a known the PMCS computer in the control room to Amendment 2 7.M
i
- 1 M 23A6100AF ,
y Standard Plant - nr_v g n i I' (0 OperationalConsiderations t!oc, each power operated isol2 tion velve !
is provided with a separate manual control The pumps, valves, piping, etc., used switch in the control room which is indepen. ;
for the LPFL are used for on er modes of dent of the automatic and manual the RHR. Initiation of the LPFL mode is system-level isolation logic. . l
-- automatic and perator_ action is re- .
quired for at I minutes. ' The op- . Subsection (3), below, provides a descrip-h.-20-08 erator may con e RHR pumps and in. tion of the various input variables and 1 jection valves manually after LPFL ini- sensing methods used to monitor the vari- l tiation to use RHR capabilities in other ables and provide the inputs to the LDS for '
modes if the core is being cooled by init!ation of the isolation function. Each other emergency core cooling systems. variable is recorded and/or indicated in the main control room.
Temperature, flow, pressure, ar.d valve- .
por' ion indications are available in (2) Supporting System (Power Supplies) j the sontrol room for the operator to *
-- l assess LPFL operation. Valves have in- Supporting systems for the LDS include the i dications full open and full closed pos- instrument logic and control power supplies l itions. Pumps have indications for pump acd vrive motive power sources. All LDS in-running and pump stopped. Alarm and strument and logic pwer is supplied by the j indication devices are shown in Figures respective divisional SSLC logic power sup- !
5.410 and 7.3 4. plies. See Section 8.3 for description of '
the SSLC logic /ower supplies that are-(j) Farts of System Not Required for Safety typical for LDS. l The nonsafety related portions of the The power for the main steam isolation !_
( LPFL include the annunciators and the valves (MSIVs) gas pilot solenoid valve con-computes Other instrumentation con- trol logic is supplied from all fou divi. ;
s'idered messafety-related are those in- sions of the SSLC bu'ses.' The MSIVs are i dicators which are provided for operator spring loaded, piston operated valves de-information, but are not essential to signed to fail closed on loss of electric ,
correct operator action. power or pressure to the' valve ~ operator.. '
l Main steam'line pressure is used as the ac-
. 7.3.1.1.2 Leak Detection and Isolation System tuation supply source and an auxiliary pneu-(LDS) Instrumentation and Controls matic N2 supply is provided to open the valves for test purposes or when steam pres-(1) System Identification sure is not available.
The instrumentation and control for the leak . Tht direct solenoid operated isolation detection and isolation system (LDS) con- valves in the RHR are. isolated by spring- ,
sists of temperature, pressure, radiation force with valve opening power supplied ;
and flow sensors with associated instruments- . from divisional power sources. RHR hboard l tion, power supplies, and logic used to de- valves are isolated by Division I logic for-l tect, indicate, and alarm leakage from the RHR A, by Division II logic for RHR B, and reator primary pressure boundary in certain by Division HI logic for RHR C. RHR out-cases to initiate closure of isolation board valves are isolated by Division II valves to shut off leakage external to the logic .for RHR A, by Division III logic for containment. .RHR B, and by Division I logic for RHR C.
Manual system level isolation control RCIC inboard valves are isolated by Divi-switches are provided to permit the operator sion I logic. RCIC outboard valves are iso-to manually initiate (at the system level) lated by Division II logic. -
isolation from the control room. In addi-Amendment 4 7.3 19 ,
V0xse e9mmMm wm e wuv u e i
~
c n invetwo o broad epectrum of risults. For exa w ls:
Standard Pbnt , gty , m Trcnsients whirs appropriate operstor cetions return the plant to normal operating conditions risults in little to no j retense of radioactive species. Releases from these Subsection 15.2.43.2.1. transients occur to controtted areas such as the suppression l l
j pool and result in nestigible plant contamination and no l
15.2AA Barrier Performance release to the envirorvnent. 5]
1 15.2.4A.1 Closure of All Main Steamlice (2) Transients resulting from or coincident with the failure of Isolation Valves major RCPB equipment requiring (mediate plant shutdown and f
the associated depressurization under controtted shutdown The nuclear system relief valves begin to ope directives may result in targer retesses with larger at approximately 2.2 seconds after the start c ,l radiological i @ act. However, even such releases normally isolation. The valves close sequentially as th l result in either no environmental release or releases '
stored heat is dissipated but continue t within normat operating dose limitations. -
discharge the decay heat intermittently. Pea I
pressure at the vessel bottom reaches 1242 psig to envelope the potential for offsite radiologitat i mact, a worst I below the pressure limits of the reactor coolar case htsed upon a transient such as example (2) above is oescribed pressure boundary. Peak pressure tu the mai 'beluw. Such transients cordervatively overpredict the actual steamline is 1208 psig. , radiological impact by factors greater than 100, u.4 4.s.* mpeurmiuuu-suutuuwu 15.2AA.2 Closure of One Main Steamline Evaluation Isolation Valve 15.2A.5.2.1 Fission Product Release from Fuel No significant effect is imposed on the RCPB, since, if closure of the valve occurs at an While no fuel rods are damaged as a consequence unacceptable high operating power level, a flux of this event, fission product activity associated or pressure scram may result. The main turbine with normal coolant activity levels as well as bypass system continues to regulate system that rel tsed from previously defective rods will _
pressure via the other three open steamlines. be released to the suppression pool as a i, ~,
consequence of SRV actuation and vessel 15.2A.5 Radiological Consequences depressurization. The release of activity from previously defective rods is based in part upon 15.2A.5.1 General Observations measurements obtained from operating BWR plants The radiological impact of transients involves (Reference 1),
consequences which do not lead to fuel rod damage Because each of those transients identified as a direct result of the event itself. Addi- previously (which cause SRV actuation) will result tionally, many events do not lead to the depres- in various vessel depressurization and steam surization of the primary system but only the blowdown rates, the transient evaluated in this venting of sensible heat and energy via fluids at section is that one which maximizes the coolant loop activity through relief valves to radiological consequences for all transients of the suppression pool. In the case of previously defective fuel rods, a depressurization transient this nature. This transient is the closure of all main steamline isolation valves. The activity will result in considerably more fission product carryover to the suppression pool than bot- airborne in the containment is based on the analysis presented in Reference 1. The results of standby transients. The time duration of the these analyses are presented in Table 15.210, transient varies from several minutes to more which was used in evaluating the radiological dose ,
than four hours, consequences in this section.
These observations lead to-the realir.ation 15.2A.5.2.2 Fission Product Release to that radiological aspects /an involve / broad Environn.ent spectrun/of results. Fpf example: ' /
/ / / Because this event does not result in the (1) Transients whe*/e appropriate' operator actions (aeeedst/I win b qugrerr (p% immediate need to purge the containment, it is -
/
A 15.2 10
s A
4 e 4
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l l
CHAPTER 8 RESPONSES 1
.I l
I l
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4
- .t
. QUESTION ,
L435.027-Section 8.3.1.2.2 states. chat the SSLC: redundancy.is based on the
, . capability of any two of the four' divisions:to provide the minimum safetyffunctions necessary to shut down the unit in case of an accident-
- and maintain it in the safe shutdown condition.: Why'can't the~ unit be
. shut down in case of an accident with only-one of.the four divisions, available? Identify the. systems or loads needed that require that two of the four, divisions be available. g
RESPONSE
435.027 Section 8.3.1.2.2 was~ incorrect and has been revised in accordance with; '
-attached mark.up. The-reactor can beLsafely shut;down from the control- )
room with 'any one of the three load groups available. .j
'l 1 QUESTION 435.032.Section 8.3.1.4.2.1 identifies the standards that are used.for the' separation of equipment for the systems referred'to in subsection' 7.1.1.3, 7.1.1.4, and 7.1.1.6 (safety-related contro1~and. ,) :
instrumentation systems). 'IEEE 384i 1974 however.is not listed. The "
d separation of equipment in these systems should comply with thet ;j requirements of this standard. Please< verify that this is the case'. -!
l In addition, the listed. standards and requirements are not identified as being applicable to subsection'7.1.1.5 (safety.related' display )
instrumentation). Please verify'that they are indeed i!pplicable to this i subsection.
1 RESPONSE , l 435.032 IEEE 384 is addressed in Tables 7.12 and 8.1 1, as endersed by ]
Regulatory Guide 1.75 Since the requirements'of'this guide. envelope ;
and endorse IEEE 384, it is not'necessary to addre8s IEEE 384 ]
separately.
, , I To be consistent with the Standard Re'iew1 v Plan format (SRP Tables 7-1, I 7-2 and 8 1), and to avoid unnecessary redundancy in the. text,<we have a not addressed the IEEE standerds separate from~the Regulatory Guides .
which endorse them. However, since IEEE 379 was inadvet:tently mentioned
.in addition to RG 1,53, we have modified and clarified thelparegraph per- .
the attached mark-up.
i
'i Also,-the separation' requirements do apply to the Safety Related-L Display. Therefore, a' reference to.Section 7.1'.l.5thas been added as' :
l marked. <
j K '
j' ...........................................................-................... -
i I
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4 1
=
'. MkN 23A6100AG
... * ' Standard Plant REV A .
. As indicated in Subsection 8.1.3.1.2.3, BTP (b) RG 1.32 - Criteria for Safety Related PSB 1 is an interface requirement for the Electric Power Systems for ,
applicant. Otherwise, the onsite AC power system Nuclear Power Plants ;
is designed consistent with these positions.
(c) RG 1.47 - Bypassed and Inoperable Sta- !
(4) Other SRP Criteria: tus Indication for Nuclear Power Plant Safety Systems (a) NUREG/CR 0660 - Enhancement of Onsite DieselGenerator Reliability (d) RG 1.75 - Physical Independence of Electric Systems As indicated in Subsection 8.1.3.1.2.4. the operating procedures and training of person- (e) RG L118- Periodic' Testing of Electric nel are outside of the Nuclear Island scope Power and protection Systems of supply. NUREG/CR 0660 is therefore im- 3 posed as an interface requirement for the Regarding Position C 1 of Regulatory guide I applicant. 1.75, see Section 8.1.3.1.2.2(6). Otherwise, the SSLC power system is designed in accordance 5.3.1.2.2 Safety. System Igle and Control with recommendations of this guide, and with the (S) LC) Power Supply other listed Regulatory Guides. ,
The SSLC power supply oneline diagram is There are four independent electrical divi-illustrated in Figure 8.3 6. The following sions, each with its own individual power supply analysis indicates compliance of the SSLC power as illustrated on Figure 8.3 6. The normal unio-supply to applicable NRC General Design Criteria terruptible power (UPS) to each of the four SSLC (GDC), NRC Regulatory Guides and other criteria divisions is provided by its divisional inverter consistent with the Standard Review Plan (SRP). powered by its divisional DC bus. An AC/DC con-verter powered by a 430 VAC bus providec the nor-Table 8.1-1 identifies the SSLC power supply mal DC power with a floadng battery as a back- -
and the associated codes and standards applied in up. The SSLC power supplies are not shared (
accord.ance with Table 8-1 of the SRP. Applicable among multiple reactor units since the ABWR is a criteria sie listed in order of the listing on single unit plant design.
the table, and the degree of conformance is discussed for each. Any exceptions or The $SLC redundancy is based on the capabi-clarifications are so noted. tity of an%of the four divisions to provide . -g '
(1) General Des.tp Critena (GDC):
[M" "I"I"d" I"'n" ease of a.n accident and main-I MMCN
, down the unit,gi I
tain it in the safe shutdown condition. .
(a) Criteria: GDCs 2,4,17, and 18. d, l G\ The SSLC power r.upply system is designed to (b) Conformance: The SSLC power supply is in part, or as a whole, as applicable.p equipment permit inspection and fenacres, and alland testing automatic andof nllimportan The GDCs are generically addressed in - manual switching functions.
Subsection 3.1.2 (3) Branch Technical Positions (B'l Ps):
l (2) Regulatory Guides (RGs):
l (a) BTP ICSB 21 - Guidance for Application l (a) RG 1.6 - Indepcndeace Between of Regulatory Guih 1.47 l Redundant Standby (Onsite)
Power Sources and Between (b) BTP PSB 1 - Adequacy of Station Electric Their Distribution Systems Distrihetion Syst.:m Voltages AmcMment 2 0.M0
s /XBMStandard Plant 2sastooxo REV A discharged by: equipment. The equipment is then designated i
" associated per Regulatory Guide 1.75. Cables I (1) identifying applicable criteria; . used to' connect such equipment are safety grade and qualified and routed a's " associated cir-(2) issuing working procedure to implement these . cuits" and marked as described in Subsection criteria; 8.3.1.3.
(3) modifying procedures to keep them current gJ.1.4.2 Independence of Redundant and workable; . Safety Related lastrumentation and Control Systems (4)- checking the manufacturer's drawings sad . .
specifications to ensure compliance with . This subsection defines independence criteria procedures; and applied to safety related electrical systems and ~
instrumentation and controf equipment. Safety-.
(5) controlling installation and procurement to related systems to which the criteria apply are 1 assure compliance with approved and issued those necessary to mitigate the effects of anti- .)
drawings and specifications. 'cipated and abeermal operational transients or j design basis accidents. This includes all those 1
The equipment nomenclature used on the ABWR systems and functions enumerated in Subsections l
- standard design is one df the primary mechanism 7;1.1.3, 7.1.1.4, 7.1.1.5, and 7.1.1.6. The. i for ensuring proper separation. Each equipment tern; ' systems
- includes the overall complex of.
and/or assembly of equipment carries a single ; actuated equipment, actuation devices (actua- 3 number, (e.g., the item numbers for motor drivers tors), logic, instrument channels, controls, and I are the same as the machinery drivers). Based on interconnecting cables which are required to per- I these identification numbers, each iteto can be form system safety functions The criteria out- 1 identified as essential or nonessential, and each lines the separation requirements necessary to j essentialitem can further be identified to its achieve independence of safety related functions 1 safety separation division. This is carried compatible with the redundant and/or diverse !
through and dictates appropriate treatment at the equipment provided and postulated events. I design level during preparation of the .
j manufacturer's drawings. g.3.1.4.2.1 General '
Non Class 1E equipment is separated where de- Separation of the equipment for the s!
sired to enhance power generation reliability, referred to in Subsection 7.1.1.3,7.1.1,ka 'd >
although such separation is not a safety ~ 7.1.1.6 is accomplished so that they are in consideration. compliance with the substance and intent ofIEEE.
279g M T10CFR50 Appendit A, General W 2 i
Once the safety related equipment has been Design Criteria 3,17,21 and 22,' and NRC identified with a Class 1E safety division, the Regulatory Guides 1.75 and 1.53A-L divisional assignment dictates a characteristic. .. -QtitYFT) (IUIM). . 'j l color (Subsection' 8.3.1.3) for positive visual . ~ Independence of mutually redundant and/or di- ,
l- identification. Likewise, the divisional iden- verse Class 1E equipment idevices, and cables i
L tification of all ancillary equipment, cable and is achieved by physical separation and/or elec-associated raceways match the divisional assign- trical isolation. Physical separation and/on' ment of the system it supports. electricalisolation is provided to maintain the independence of nuclear safety-related circuits l There ere certain exceptions to the above and equipment so that the protective function re- !
where non Class IE equipment h connected to quired during and following a design basis event .
' Class 1E power sources for functional design rea . including a single fire anywhere in the plant or 1
. sons (viz., the stat.dby AC lighting). This is ' 'a single failure in any circuit or equipment can. .l immediately apparent by the absence of essential be accomplished._ ;
>;> classification id.entification of the connected 1 L ,
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