ML20079F666
| ML20079F666 | |
| Person / Time | |
|---|---|
| Site: | Clinch River |
| Issue date: | 06/01/1982 |
| From: | Longenecker J ENERGY, DEPT. OF |
| To: | Check P Office of Nuclear Reactor Regulation |
| References | |
| HQ:S:82:037, HQ:S:82:37, NUDOCS 8206080129 | |
| Download: ML20079F666 (22) | |
Text
,
Department of Energy Washington, D.C. 20545 Docket No. 50-537 JUN 01 1992 HQ:S:82:037 Mr. Paul S. Check, Director CRBRP Program Office Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C.
20555
Dear Mr. Check:
RESPONSES TO REQUEST FOR ADDITIONAL INFORMATION
Reference:
Letters, P. S. Check to J. R. Longenecker, "CRBRP Request for Additional Information," dated February 26, March 11,15, 23, and April 9,1982 This letter formally responds to your request for additional information contained in the reference letters.
Enclosed are responses to Questions CS 210.5 and 9, CS 270.14, CS 410.5, 9, and 10 CS 281.6 (revised), and CS 421.2, 3,10,11,12,13,15, and 16.
These responses will also be incorporated into the PSAR in Amendment 69; scheduled for submittal later in June.
Sincerely, Oh N
v Jc n R. Longene r, Manager Licensing & Envi nmental Coordination Office of Nuclear Energy Enclosures c'c : Service List Standard Distribution Licensing Distribution 8206080129 820601 PDR ADOCK 05000537 4
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Ouestion CS210.5 Describe methods used to evaluate creep buckling for components under axial compression.
Specify acceptance criteria used and identify design margins.
Resoonse For the Reactor System, Reactor Enclosure System, and Primar/ Heat Transport System,no major high temperature component has been identified as being subjected to a sustained axial compression loading.
l l
However, where buckling is a potential f ailure mode for elevated temperature components, an evaluation wilI be performed per the procedures def ined in ASME Ade Case 1592-7 (para 3250 and Appendix T para T-1520).
QcS210.5-1
'Page 7 (82-0287)[8/22]i35 Question CS210.9 Specify strain or deformation limits used to ensure operability of active mechanical omnponents in elevated-temperature service.
Since sustained stresses at service level C and D may induce excessive def ormation due to j
creep mechanisms, specify the criteria used to ensure piping f unctional capabil ity.
Resoonse The heat transport system piping is evaluated for strain and def ormation resulting from the specified operating conditions (normal, upset, emergency, and faulted). This evaluation includes the ef fects of ratcheting, the interaction of creep and f atigue, and the possibility of buckling and structural Instabil ity.
The acceptability criteria f or the HTS piping applied as strain, creep-f atigue damage and buckling ilmits are contained in Appendix T of ASME Code Case 1592-7, as specified in the piping design specification. The limits on stress or total accumulated def ormation imposed by Appendix T ensure piping functional capability, in addition, Appendix T specifies f actors f or load-controlled and strain-controlled buckling f or dif ferent l oading conditions. These f actors provide a suf ficient margin of safety to preclude catastrophic buckling of the piping during plant operation. The predominant l
loading on the p! ping system is due to thermal expansion at operating temperature (strain controlled loading). Since the piping restraints constrain the piping system from excessive def ormations even f or seismic loadings, compliance with the strain f actors specified in the Code Case 1592-7 is suf ficient to preclude buckling of the piping and to ensure f unctional capabil ity, f
The limits used to ensure operability of active mechanical components in elevated temperature service are presented and discussed in WARD-D-0174.
Ref erence 12, PSAR Section 1.6.
QCS210.9-1 Amend. 69 C%M @@t
Page 5 (82-0287) L8/22Ji34 Question OCS270.14 The Nuclear Regulatory Commission is proposing to amend its regulations applicable to nuclear power plarits to achieve consistency, reliability, and reproducibility of test results, and to provide adequate control of equipment qualif ication testing.
To achieve these ends, the proposed rule requires that all qualification testing of equipment important to saf ety shall be perf ormed by testing organizations accredited by the Institute of Electrical and Electronic Engineers, Inc. (IEEE) to perf orm such tests. This requirement conf orms to an agreenent between the NRC of testing organizations.
How does the applicant plan to achieve consistency, reliability, and reproducibility of test results, and provide adequate control of equipment quellfication?
Resoonse CRBFP achieves consistency, reliability, reproduclbility of test results, and
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adequate control of Class 1E equipment quellfication through the ef fective progran management of the Project environmental quellfication requirements imposed by WARD-D-0165, all applicable Regulatory Guides and industry Standards, and the quality assurance requirements imposed by 10CFR50, Appendix B.
These requirenents are called out in the Class IE equipment speci f icati ons.
The equipment specif Ications require the equipment supplier to submit a Class 1E environmental quellfication program. Approval of an equipment supplier's qualification program is provided only if it satisf actorily fulfills the CRBRP requirements. This assures that all Class IE equipment quellf ication requirements are properly identif ied, the requirements are imposed on the equipment supplier, the supplier is Interpreting the requirements properly and conducting the qualification progran in accordance with 10CFR50, Appendix B.
Ultimately the qual:Ilcation documentation will contain suf ficient data to support the conclusion that the Class 1E equipment will maintain its f unctional operability under all servJce conditions postulated to occur during the Installed lif e f or the time It is required to operate.
CRBRP requires that quellf ication testing f or Class 1E equipment be performed in compliance with all applicable quality assurance requirements defined in 10CFR50, Appendix B.
This assures that the testing f acility has a program which provides Independent verification that the test results were obtained in a consistent, reliable, accurate and controlled manner, independent s
verification by the testing organization, and CRBRP Quality Assurance, provides assurance that the tests are being conducted properly.
s QCS270.14-1 Amend. 69 May 1982
Page - 6 82-0287 [8,223 #37 Question CS410.5 (9.6.1)
The control room HVAC system can be operated in normal, outside air filtration and total recirculation modes of operation. Provide a tabulation of valve / damper position and equipment status f or the control rom HVAC system for each of the possible modes of operation.
Response
Refer to attached Table CS410.5-1.
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Poge - 7 82-0287 [8,22] #37 TABLE CS410.5-1 NOTES:
(1) One redundant group of equipment, A or B, OPERATING (f ans on, valves and dampers open); while other redundant group, B or A, in STANDBY (f ans of f, valves and dampers closed).
(2) The f ailure of the operating unit to perf orm or the operating damper f alling closed will result in either: (a) a low flow signal from the operating group (A or B) flow switch which will energize the standby group (B or A) and deenergized the group operating with low flow; or, (b) a low Control Room to atmosphere dif ferential pressure alarm which will signal the Control Room Operator to switch to the standby group and decnergize the operating -group.
(3) Both close on loss of of f-site power.
(4) Note (2) pertains during periods when filter f an 25ACA441 ( A or B) operation is required (e.g., when toxic gases or high radiation is present at an outside air intake).
(5) MOD 75AA or BA on the operating group (A or B) is open while the standby group damper is closed, as is the outside air supply patnway to the 25ACA410A, B; by the closure of A0V122A, B.
(6) Control Room Operator will deenargize no -safety class exhaust f ans 25ACA465, 467. Exhaust pathway blocked automatically by closure of A00168, 169.
(7) Control Room fliter f an 25ACA441 of the operating group (A or B) wil I start and its associated dampers MOD 75 (AB or BB) and MOD 76 (A or B) will open. The standby group 25ACA410B or A, 25ACA451B or A, and 25ACA441B or A, will start and its associated dampers will open. Af ter the Control Room Operator ascertains the perf ormance of both groups, the Operator will deenergize one group and put It into the standby condition.
(8) For toxic gas or smoke at Remote Air intake MOV104A and B are closed.
Radiation signal alarms at the Control Panel to alert Operator for decision to close MOV104A and B manually remote from the Panel.
(9)
Should the Containment isolation Signal be present f or an extended period of time (as would be the case for a radioactive release to the Reactor Containment Building) the Control Room HVAC system can be returned to the normal mode of operation by the use of administrately controlled key operated switches. The system would still respond as shown to smoke, toxic gas, or radiation signals.
QCS410.5-3 Amend. 69
{
May 1982
' Page - 9 62-0287 Le,22J #37 Question 410.9 (9.6.1)
One exhaust f an is provided fot each of the four battery rooms.
Verify that the exhaust duct from each battery room is provided with positive air flow Indication with annunciation in the control room for loss of flow and Instrument f ailure.
Describe the steps that will be taken to prevent hydrogen bulldup to a combustible con entration on loss of a battery room exhaust f an.
Resoonse As shown on Figure 9.6-2, each Class 1E powered battery exhaust f an is provided with a low flow switch in its exhaust duct.
Loss of flow will activate an alarm in the control room.
If the batteries in the affected cell are being charged, the charging operation will be shutdown. Temporary ventilation will be provided and the cell atmosphere will be tested to insure the hydrogen level limits are not exceeded while the fan is being repaired.
1 1
QCS410.9-1 Amend. 69 May 1982
P;ge - 12 82-0287 [8,22] #37 Ouestion CS410.10 (9.6.2)
Provide Indications on Figures 9.6-4, 9.6-5, 9.6-6 and 9.6-9 to show the location between safety class and non-saf ety class portions of the Reactor Containment Building, containment annulus. Radioactive Argon Processing Subsystem (RAPS) and Celi Atmosphere Processing Subsystem (CAPS) HVAC systems.
The means for Isolating the essential portions f rom the nonessential portions should be shown on the figures.
Eniponse 1.
As Indicated on Figure 9.6-4 (Note 7), all components shown are non-safety related with the exception of the widainment penetrations and Isolation valves which are Indicated as safety Class 2.
Valves A0V47B and C isolate the essential portion of exhaust line 24-WBK-25ARD-26 from the non-essential portion and valves A0V46B and U Isolate the essent!al portion of supply line 24-mBK-25ARD-25 from the non-essential portion.
2.
As Indicated c,n Figure 9.6-5 (Note 6), all components shown are saf ety Class 3, with the exception of the containment penetrations and Isolation valves which are Indicated as saf ety Class 2.
Since all portions of the HVAC system shown are essential, Isolation from non-essential portions is not applicable.
3.
As Indicated on Figure 9.6=6 (Note 7), all components shown are non-saf ety related with the exception of the El&C unit coolers (25 ARA 021, 022, 023)
I which are Indicated as safety Class 3.
The El&C cubicle unit coolers are 100% recirculating units located totally within the cubicle served, and provided with emergency chilled water and IE power.
4.
As indicated on Figure 9.6-9 (Note 7), all components shown are non-saf ety related.
Since all portions of the HVAC system shown are non-essential, I
Isolation from essential portions is not applicable.
l QCS410.10-1 Amend. 69 May 1982
Page 1 (82-0321) [8,22] 064 Question CS 281.6 In the CRBR primary and intermediate sodium piping system, Fe, Cr, and Ni are dissolved frm the high terrperature regions and deposited in the lower temperature regions because of super-saturation. Included in this process of mass transfer is the formation and decomposition of various transition metal and sodium double oxides. Deposition of these mass transfer and corrosion products may cause flow restrictions and loss of heat transport efficiency of heat exchangers. Describe the criteria and bases in your analyses of mass transfer and deposition of corrosion products in the CRBR primary and IHX sodium systers to assure necessary system flew and heat transfer. Include the instrumentation and detection system which will alarm when these limits are exceeded.
Response
Primary Sodium Piping System
% c IHX is designed with an effective tube length of 24.21 feet which is 5.12 ft or 33% greater than required for nominal operation.
Included in the 33% excess allowance is a 9% factor for heat transfer degradation & e to the deposition of mass transfer products on the primary side of the unit.
Corrosion pro & cts go into solution in the core region of the reactor either by direct dissolution or by the formation of soluble oxide cmplexes. As the coolant flows through the cooler regions of the system it becmes super-saturated with respect to these corrosion products and they precipitate out.
Precipitation is expected to occur in the IHX. W e deposits will result in sme degradation of the overall heat transfer co-efficient. his potential f
probl e was recognized in the early 70's and work was performed to determine the magnitude of this effect. A suntrary of this work is given below.
'Ivo tube-in-shell heat exchangers were available from an ongoing corrosion program. W e first had operated in an all stainless steel syst s with a hot leg temperature of 1325 F for 0.84 years. We second operated in a
'I304ss/Incoloy 800 system for 1.5 years with a hot leg terrperature of 1100 and a high, (28 ppn by amalgamation) oxygen level. Heat transfer measurements were made on these heat exchangers and cmpared with similar measurements on new heat exchangers of identical design. We percentage change in heat transfer coefficient was determined for each set of readings.
T s
QCS281.6-1 Amend. 69 e
May 1982
__________.A__
P2ge 1 (82-0321) [8,22] #64 The exposure conditions experienced by the heat exchangers were excessive in that the f Irst one operated with a high maximum hot leg temperature (13250F) and the second with a high oxygen level. Equivalent operating 0
at 5.3 years ani ST"4 yeahs.onditions (1100 F mean tit was judged that the Sep)osit thic were calculated times at react ratin c mx reached equll Ibrium.
Addi tional increases in thickness are prevented by fl ow induced shearing of the f riable deposits.
The average deposit heat transf er resistance was calculated es 8.4 x 10-5 h-ft2 oF/ BTU.
The overall heat transf er coef ficient of the IHX design was 1190 BTU /h-f t2 oF.
Adding the deposit resistance gives a calculated degradation of 9%. This 9% value was used f or the FFTF lHX design and in the CRBRP design.
The ef f ect of corrosion products on pressure drop and flow blockage is not addressed in Section 5.3 of the PSAR.
Flow blockage is addressed in Sect! on 15.4.1.3.
FIw blockage in the Core Assembly in 15.4.1.3.1.
' Prevention and Detection' C, (Corrosion Products).
Deposition Induced pressure drop in the IHX is not considered to be a f actor because of the large flw cross sections on the shell side.
The ef f ects of corrosion product deposition In the PHTS will result in a very gradual change, If any, In system perf ormance.
The PHTS perf ormance is continuously monitored and critical perf ormance paraneters are calculated by the plant computer f rom temperature, flow rate and pressure drop sensors in the FHTS system.
Monitering sensors In each Ioop inelude resistance temperature detectors at the Inlet and outlet of the IHX, pressure sensors in the hot and cold l egs, and a f low meter.
The perf crmance evaluation f or the PHTS is identified in the system procedures and Ineludes: 1) lHX Thermal Perf ormance - calcuiatlon of overali U value, 2) Total Loop Pressure Drop, 3) lHX Primary Side Pressure Drop, and 4) Reactor Vessel Pressure Drop.
Informedlate Sodlum Plofng Systen The Intermediate Heat Transport System is designed to operate with the oxygen content in the sodium controlled to 2.0 PPM maximum to enhance the leak detection capability In the system.
A leak in a water / steam tube of a plant evaporator or superheater results in an oxygen concentration increase In the Intermediate sodium; thus low oxygen background is maintained f or early detection should such a leak occur.
The 2.0 PPM oxygen level is an order or magnitude lower than that required f or controlling corrosion of the austenitic stainless steel in sodium. Theref ore, the mass transport of corrosion products and the subsequent potential degradation of heat transf er capacity is not a concern in this system with this sodium coolant chemistry.
Futhermor e, QCS281.6-2 Amend. 69 May 1982
~
P ge 2 (82-0321) [8,22] #64 In the stesn generators, which are in the cooler sodium region of the system and are, theref ore, most subject to mass transf er deposition, the limiting heat transf er conditions exist on the steam / water side of the tubes which would not be af fected by sodium chemistry.
The intermediate sodium oxygen level is monitored by an on-line pluggl'ng
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temperature Indicator and periodic sodium samples are also obtained (see PS AR Sect i on 9.8 ).
QCS281.6-3 Amend. 69 May 1982
P2g3 3 [8,223#25 OuestIon OCS421 DZ The staf f requires that the applicant document how the CRBR primary and secondary shutdown systems meet GDC 20 through 29. Also, provide documentation showing the separation maintained between the shutdown systems, the independency of the shutdown systems, common mode f ailures of the shutdown systems (i.e., sharing of Inverters, an overvoltage of over/under frequency In the Reactor Protection system power supply), the testability of the shutdown syste,s manual Initiation for both systems, diversity of the shutdown system electrical circuitry components, and how each of the shutdown systems independently meet IEEE-279.
Resconse 10CFR50 GDC 20 through 27 are redefined as CRBRP GDC18 through 25 and are addressed in PSAR Chapter 3.1.
GDC 28 and 29 are not included in the CRBRP GDC. However, the present design fully meets both criterion, Criterion 28 - The plant is designed to limit the consequences of o
postulated reactivity transients to acceptable levels.
Events considered Include the accidental withdrawal of single control rods or groups of control rods at conservative withdrawal rates. Also considered are rod drop accidents, accidents that lead to f uel rod motion due to the loss of hydraulle holddown forces, and accidents that cause cold sodlum
! r.,:r t ! c r.,
Criterion 29 - The Project has taken steps to assure an extremely high o
probability of accomplishing required safety functions in response to anticipated operational occurrences. Not only are applicable standards ut!!Ized, but an extensive progrm of qualitative and quantitative analysis and developmental testing is undt rway. The object of this Reliability Progre is twofold:
- 1) enhance system reliability in areas where analysis shows improvement is desirable, and 2) verify component rellability.
Section 7.1 of the PSAR discusses separation criteria utilized in the design of Tne protection system.
Section 7.2 includes a discussion of testability and manual Initiation as welI as of those design features which provide Independence and freedom from common mode f ailure. These provisions ensure that under normal operating conditions, and the great majority of abnormal conditions, each shutdown system separately, fut ly meets the requirements of IEEE 279.
1.
Extremely Unlikely Events The Prtrrary Shutdown System Is required t'o respond to all anticipated, unlikely, and extremely u..likely events. The Secondary Shutdown System is required to respond to all anticipated and unlikely events. The only extremely uniIkely event to which the Secondary Shutdown System must respond is the safe shutdown earthquake. The al lowable damage limits f or the Secondary System are one level higher than for the Primary Shutdown System.
QCS421.02-1 Amend. 69 May 1982
'P ge 4 [8,22]#25 l
2.
Control / Protection Interaction 1
Ref er to the response to Question 421.01.
Strict independence is maintained between the Primary and Secondary Shutdown Systems with three exceptions:
1.
Instrument Ground The plant has one instrument ground.
Both Shutdown Systems use this same f
ground.
However, physically separated cables are run to this ground from each Instrument channel in the Primary and Secondary Shutdown Systems.
2.
HTS Pump Trip Both Shutdown Systems trip each HTS pump breaker through separate trip coils. The separate trip coils and relay logic used to drive them provide sufficient independence to prevent failure propagation from the pump trip circuits to the control rod trip circuits of either Shutdown Sistem.
3.
Power Supplies The Primary and Secondary Shutdown Systems obtain power from the same three power divisions. Each power division utilizes separate IE un-Interruptible power supplies including batteries, rectifiers, inverters, etc.
Which are designed to prevent transient electrical power f l uctuati ons.
Nevertheless, mitigating features have been provided in the design including:
isolation of IE power division from potential transient sources.
o surge suppression by the Inverters.
o power supplies utilizing protective devices (i.e., circuit breakers, o
f uses, veri stors) filtering, Isolation transformers, overcurrent and overvoltage trips, and component derating.
l QCS421.02-2 Amend. 69 May 1982
PIge 5 L8,22]s25 Question OCS 421.03 Provide a lIst of saf ety grade trips and non-saf ety grade trips f or the Reactor Protection System.
Provide conf irmation that credit will be taken f or only the saf ety grade trips in the analysis of Chapter 15.
Resnonse All trips in the Reactor Shutdown System are saf ety grade and are listed in Tabl e 7.2-1.
Theref ore, any Resctor Shutdown System trips used in Chapter 15 are saf ety grade.
l I
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QCS421.03-1 Anend. 69 May 1982
ttge 10 L8,22Jr25 Ouestion 0CS421.10 identify any sensors or circuits used to provide Input signals to the protection system which are located or routed through non-seismically quel lf led structures. This should include sensors or circuits providing input f or reactor trip, emergency saf eguards equipment, and saf ety grade Interlocks.
l Verification should be provided that the sensors and circuits meet IEEE-279 l
and are seismically and environmentally qualifled.
Testing or analyses l
perf ormed to insure that f ailures of non-seismic structures, mountings, etc.,
will not cause f ailures which could interf ere with the operation of any other portion of the protection system should be discussed.
Resnonse All sensors and circuits that provide input signals to the protection system are located and routed through seismically quellfled structures, are designed to IEEE 279 cr iteria, and wll l be seismically and environmentally qualifled.
QCS421.10-1 Amend. 69 May 1982
Page 17 [8,22]i25
_0uestion OCS421.11 Verify that a f ailure modes and of fect analysis wilI be performed for each of the ESF systems I denti f led in Section 7.3.1.
Response
a Section 7.3.1 only discusses the Containment isolation System (ClS). The CIS must meet IEEE 379 which requires an analysis to show compliance with the single failure criteria.
Although the FMEA is an acceptable format for this analysis, it is not the only approved of method. Therefore FMEA's wilI not necessarily t;e performed, however, analyses of the CIS per the requirements of IEEE 379 will be made, f
QCS421.11-1 Amend. 69 May 1982
mge 18 LB,22Jr25 Question OCS421.12 The staf f has recently issued Revision 2 to Regulatcry Guide 1.97,
" Instrumentation f or Light-Water-Cooled duelear Power Plants to Assess Plant and Environs Conditions During and Folicwing an Accident." This revision reflects a number of major changes in post-accident instrumentation, and includes specific implenentation requirements f or plants in the license review stage. Discuss this Reg. Guide and how it is applicable to the Breeder.
Resoonse The requirements of Regulatory Guide 1.97 Revision 2 is applicable in pri nci pl e to CRBRP. Proper Provisions f or instrumentation to monitor plant variables and systems during and f ollowing an accident are necessary for RRP r
as they are for an LWR plant. However, the requirements delineated in Regulatory Guide 1.97, Rev. 2, are specifically for LWR plants and a dif ferent set Is required f or LMFBR's because of f undcmenteI technologicai dif f erences af f ecting their design and operation.
For a diset.ssion of the implementation of these requirements see Appendix H, item II.F.3, Instrumentation f or 2nitoring Conditions (Reg. Guide 1.97) of the PSAR.
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QCS421.12-1 Amend. 69
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May 1982
'Page 19 [8,22]i25 RESPONSE TO NRC OUESTIONS 421.13 Ouestion Provide and describe Inf ormation f or NSSS and BOP safety related setpoints that verlfles that environmental error allowances will be based on the highest value determined in qualification testing.
Resoonse Perf ormance requirements are part of the input in calculating safety setpoi nts. All NSS ar.d BOP saf ety related instrumentation and control equipment shall meet the perf ormance requirements indicated in the ordering data under any combination of environmental conditions as defined in the applicable equipment specifications. The environmental conditions include normal end accident environments.
If test data obtained during equipment quellfications are greater than the specified perf ormanca requirements the Instruments setpoints will be suitably adjusted to assure conservative margins.
Requirements covering instrumentation perf ormance under worst case environmental conditions have been specified for the Reactor Shutdown System (RSS) equipment. These perf ormance requirancnts are part of the Input used in calculating the RSS saf ety related setpoints.
If the test data obtained during equipment quellfication Indicate that the predicted allowance for environmental errors is exceeded, the RSS setpoints wil l be suitably adjusted so os io ensure the same conservative margins.
QCS421.13-t Amend. 68 May 1982
Page 20 [8,22]#25 Ouestion OCS421.15 Identify and document where microporcessors, multiplexers, or computer systms may be used in or interf ace with saf ety related systems.
Resconse Many microprocessors, multiplexers, and computers are used in CRBRP systems; how ev er, in general, they are used in non-1E applications.
Whenever a microprocessor, multiplexer or computer acquires a 1E signal, that signal is isolated by a quallfled IE isolator bef ore being utilized by a non-1E system.
The two systes which use microp ocessors, multiplexers or computers f or 1E applications are the Solid State Programable Logic System (SSPLS) and the Radiation Monitoring System.
Inf ormation about these systems is provided beIow. The Piant Data Handling and Display System (PDH&DS) is the Iargest computer system used in the plant.
Information about this system is also provided bel ow.
The Radiation Monitoring System has Renote Processor Stations which are microporcessor based, radiation monitoring electronic and communication assembl ies.
PSAR Paragraph 11.4.2.1 describes the Remote Process Stations.
The microprocessor receives raw count rate and process system data, and manupulates the data into the desired f orm.
Data exchange and monitcr control is via channel dedicated multiplexed signal paths.
Non non-1E equipment can exercise control over a 1E radiatt on monitor.
Any data extracted f rom the 1E monitcrs f or use in non-1E equipment is via 1E grade buf fers.
The Solid State Programable Logic Systs controls and actuates Saf ety Rel ated, Cl ass 1 E equi pment.- it contains the control logic, signal condi ti oners, Isolation devices, and auxillary circuits. The SSPLS can potentially use microprocessor based circuitry.
PSAR Paragraph 8.3.1.1.2 describes the SSPLS.
The CRBRP Plent Data Handling and Display Syste (PDH&DS) Is a non-saf ety related microprocessor based system that interf aces with saf ety related systems and non-saf ety related systems as well for the purpose of retrieving data f or operator inf ormation.
The systm provides f or information display and data handling, inoperable status mont tcring of saf ety systes and emergency response f acil ity data display.
In all cases, 1E grade buffers are used f or isolatt on between the PDH&DS and saf ety related systes.
The PDH&DS is described in PSAR paragraph 7.8.
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QCS421.15-1 Amend. 69 May 1982
P ge 21 [8,22]i25 Question 421.16 1 & E Bulletin 80-06 addressed concerns related to safety equipment not remaining in its emergency mode upon reset. The applicant should specify and justify any places in the design of CRBR saf ety system logic where safety equipment will not remain in its emergency mode upon reset of an engineered safeguards actuation signal.
Resoonse:
The electrical system design at CRBRP ensures that saf ety equipment wilI remain in its emergency mode upon reset of engineered safeguard actuation signal. The electrical systems are designed to ensure:
a) Circuit breakers will close on the presence of an emergency signal where driven equipment is powered through medium or low voltage switchgear. The breakers will renain closed even af ter actuating signel has been reset. Opening of the breakers is achieved through any of the following: manual operation, or electrical f ault, or absence of process interlocks which otherwise are necessary for continuous operation of the equipment.
b) Where operated equipment is powered through motor control centers or power distribution panels seal in circuitry is provided f or the momentary contacts. The circuit will remain energized even when the actuating signal resets, and can be de-energized only by manual operation, or electrical f ault, or by absence of process Interlocks which are otherwise necessary for continuous operation of the equipment.
Examples of the system designs follow:
Frimarv and Secodarv Reactor Shutdown System Once initiated, the Primary and Secondary reactor shutdown systans and the automatic Containment isolation System (ClS) remain in a tripped condition until manually reset by the operator. These systems do not automatically reset if the actuation signal resets.
Containmen, Isolation Svstem (CIS)
As part of the CIS design, automatic back pressure valves are used on the argon supply, nitrogen supply and service air supply lines which penetrate contai nment. These valves are backpressure regulated and close automatically if the supply side pressure drops below the preset limit. the valve actuation point has been chosen to guarantee flow into the containment building if the supply side pressure is above the preset limit.
Selection of the actuation point includes consideration of the maximum accident pressure within contai nment, in addition, remote manual control switches are available to the operating staf f in the control room which allow manual operation of these val ves.
l QCS421.16-1 Amend. 68 27/ MM
P ge 22 [8,22]i25 Reactor Heat Transoort instrumentation System (RHTIS)
The SGAHRS Initiation signals are developed by the PPS system. The PPS system sends two redundant primary and two redundant secondary signals to the RHTIS 1 out of 4 trip logic.
Once a trip signal is sensed by the RHTIS It " latches In" and the PHTIS trip logic will not reset automatically when the primary initiation signal developed by the PPS system resets. All SGAHRS components wilI continue to perf orm in the "SGAHRS Initiation mode" until the operator manually resets the three SGAHRS Initiation trip in the RHTIS.
Aeroso! Release Mitigation System ( ARMS)
The Aerosol Release Mitigating System ( ARMS) sends a signal to the steam generator ventilation system upon detecting aerosols in the steam generator bays. ARMS detector coincidence (2 of 3) circults sends a signal to the Nuclear Island HVAC System which is used to melt fusible IInk closing damper valves in the HVAC duct. The f usible link controller cannot be reset without a maintenance ef fort to replace the link.
In addition to the f usible link, the present ARMS detector circuit design does not allow resetting a tripped detector if the alarm condition persists.
Design will also include provision f or preventing a reset if either an alarm or a f ault condition exists. Thi s latter provision accommodates the situation where an alarm condition (sodium leak) results in destruction of the detector.
QCS421.16-2 Amend. 68 h mm