ML20065J978
| ML20065J978 | |
| Person / Time | |
|---|---|
| Site: | Clinch River |
| Issue date: | 10/04/1982 |
| From: | Longenecker J ENERGY, DEPT. OF, CLINCH RIVER BREEDER REACTOR PLANT |
| To: | Check P Office of Nuclear Reactor Regulation |
| References | |
| HQ:S:82:098, HQ:S:82:98, NUDOCS 8210060249 | |
| Download: ML20065J978 (74) | |
Text
{{#Wiki_filter:@ Department of Energy Washington, D.C. 20545 Docket No. 50-537 HQ:S:82:098 GCT 0 4199g Mr. Paul S. Check, Director CRBR Program Office Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555
Dear Mr. Check:
Reference:
Longenecker to Check,
Subject:
Meeting Summary for Instrumentation (Chapter 7) Working Meeting, September 21 and 22,1982, dated September 24, 1982 INSTRUMENTATION (CHAPTER 7) WORKING MEETING, SEPTEMBER 21 and 22, 1982 - ADDITIONAL INFORMATION Enclosed is the additional information requested during the subject meeting for which response dates of October 4,1982, were projected. The enclosures include a list of the items from the reference meeting summary indicating their resolution followed by question responses and marked up PSAR pages keyed to the item numbers. The marked up PSAR pages will be incorporated into a future PSAR revision. Item 43 (QR 421.42) from the reference meeting summary will be submitted as soon as possible. Any questions regarding the information provided or further activities can be addressed to Mr. R. Rosecky (FTS 626-6149) or Mr. A. Meller (FTS 626-6355) of the Project Office Oak Ridge staff. 1 Sincerely, Y u .3
#'.(Longen R r cting Director, Office of the Clinch River Breeder Reactor Plant Project Office of Nuclear Energy Enclosure cc: Service List Standard Distribution Licensing Distribution g I B210060249 821004 i PDR ADOCK 05000537 J- PDR
Page - 1 (82-0962) [8,123 #29 T/6LE 12.1-1 PLANT RADI ATION ZONE (LASSIFICATION Zone Dose Rate Design Dose Specification Rate faren/hr) farea/hr) Twoe of Con g h Area Tyne Access Continuous - ** Uncontrolled
- Unrestricted Area Continuous 0.2 10.2 Aeninistrative Control 1 Restricted
! Area RoutIneIy Occupied 2.0 >0.2 to <5 Aeninistrative Control ll Restricted Continuous, Area Not Routinely Occupied Periodic 10*** >5 to 1100 Administrative Control ill Radiation Area Limited Access for Routine Tasks Unoccupied
- 100 >IDO to <5000 Special Work Permits, IV High Radiation Locked Doors, Signs, Area Limited Access Temporary BarrIcodes, for Non-routine
- Health Physics or inf requent Tasks Surveillarco 1
Unlimited >5000 Positive exclusion, l V Extremely Unoccupied
- Locked Doors, Special l High Radiation Work Permits, Continuous Area Health Physies Monitoring
- 10CFR20 criteria.
** Approaching background radiation.
*** 25 mram/hr within HAA 12.1-28 Amend. 72 Oct. 1982
i 4 Resolution of Items From the September 21 and 22, 1982 Instrumentation (Chapter 7) Working Meeting i 11A5 Comments Resolution
- 1. 7.1 Listing of Saf ety Define " Safety Related" in Table 7.1.1. Revised Table 7.1.1 attached.
Related Inst. Assure consistency with H. Denton Memo, Consistency with H. Denton and verify adequacy of Table 7.5-1. memo reviewed and Table 7.5-1 Dif f erentiation between NSSS and BOP is adequate. Supplier is not sequired.
- 2. IEC Design Criteria Add note to Table 7.2-3 to clarify what's Revised Table 7.2.3 attached.
- Inst. error included in the response times, i.e. sensor to output of signal conditioner, and note 200 maec additional time for rod release.
- 3. Review 15.1.4.1-4, 15.31.1.1.a and Table Table 1-3-3, paragraph 15.1.4.1.4, 1.3-3 for potential conflicts. Correct and paragraph 15.3.1.1.1 revised.
confusing discussions. See attached.
- 21. - RSS bypass capebility Delete existing two loop operation IEC. PSAR corrected. See PSAR page Correct PSAR. 7.2-la, attached.
- 22. Provide revised Figure 7.2- 2A. Revised Figure 7.2-2A attached.
Place in PSAR.
- 23. - Diversity - RSS Add additional discussion on diversity Discussion added in paragraph primary and secondary and figure on diveristy. 7.1.2.1.
logic Diversity Table 7.1-5 added. ' See attached. i
- 25. - Response Time See Section 7.1 Item 2 and 3. See Items 2 and 3.
- 31. Q 421.9 Verify that there are no cases where a See revised QCS 421.9-1, attached.
control system xmitter and a protection system xmitter are connected to a common sensor or input line - Revise Q 421.9 accordingly.
- 33. Correct Table 7.1-6 The error will be corrected. Check Table See corrected page 3.1-20 and corrected Error (page 3.1-20) of Contents also. List of Tables.
- 47. Discuss Design of IEC Confirm power supplies & I&C to pony See Revised Section 7.5.2.1.2 and f or PHTS, IHTS & SGS motors is safety related, new Section 7.4.3, attached.
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- 71. Pony Motor I&C 0421.34 will answer questions. See 0421.34 Response, attached.
- 80. Radiation Monitoring Information in Ch. 11 & 12 is being revised. Revised sections of Chapter 11 & 12 System Draf t copies provided. Update PSAR. are attached.
Radiation Assessment Branch will review the Radiation Monitoring System.
- 87. Fuel Handling and Review of auxiliary Systems Branch required See revised Section 7.6, attached.
Storage Safety also. Typographical error in introduction Interlock System of PSAR 7.6 will be corrected. This is not a safety related cystem.
- 88. Structural Concrete PSAR will be clarified regarding See revised Section 7.6, attached.
Cooling System requirements f or structural concrete cooling. O h 9
TABLE 7.1-1
~
SAFE.TY RELATED INSTRUMENTATION AND CONTROL SYSTEMS
- Reactor Shutdown Systems Includes all RSS sensors, signal conditioning calculation units, compara-tors, buffers, 2/3 logic, scram actuators, scram breakers, scram solenoid 57 valve power sources, control rods. HTS shutdown logic, coolant pump breakers, and mechanical mounting hardware (equipment racks).
Containment Isolation System Includes radiation monitoring sensors, signal conditioning, comparators, 2/3 logic, containment isolation valve actuators and valves. 57j Shutdown Heat Removal System Instrumentation and Control System Includes initiating sensors, signal conditioning, calculation units, compa-rators, logic, auxiliary feedwater pump actuators and controls including feedwater turbine pump, PACC DHX actuators and controls, steam relief valve actuators and valves; sensors, signal conditioning, logic and 26 5) actuators related to shutdown heat removal functions of DHRS including control of sodium and NaK pumps and air blast heat exchangers and sensors, signal conditioning, logic and actuators related to removal of heat from the EVST. Other Safety Related Instrumentation and Control l Includes Instrumentation and Controls for portions of the following functions to assure the plant is maintained in a safe shutdown condition:
. Emergency Chilled Water System
. Emergency Plant Service Water System 33 Instrumentation necessary to assure plant is maintained in safe 49 shutdown status (See Table 7.5-4)
. Heating, Ventilating, and Air Conditioning System 3
. Recirculating Gas Cooling System g,gg
*The Clinch River Breeder Reactor Plant (CRBRP)Tstructures, systems, and components 'm;:rt:nt t: ::f:t; are 4ee designed to remain functional @ in the event of a Safe Shutdown Earthquake (SSE). Th::: p?:nt f::t.r;; ;r:
;;;; d;;ign:t:d :: ::f;t3 r;i;t;d fc.;t.c;; in th: "". These include, but are not limited to, those structures, systems and components which are necessary:
To assure the integrity of the Reactor Coolant Boundary;
]ab To shut down the reactor and maintain it in a safe shutdown condition; 1 To prevent or mitigate the consequences of accidents which could result in potential off-site exposures comparable to the guideline exposures of 10CFR100.
/ NOTE: Class IE equipment loads a-e identified in Chapter 8.
1 I Amend. 59 7.1-7 Dec. 1980 l
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TABLE 7.2-3 ,,yggs ESSENTI AL PERFORMANE REQUIREENTS FOR(h INSTRUENTATION CSANMEL5 Accuracy Response Time Plant Paramater (f of scan) (msec) Neutron Flux Primary 1.1.0 <10 Secondary 11.0 <10 Reactor Inlet Plenum Pressure 12.0 <150 Sodium HTS Pump Speeds 12.0 <20 Sodl um HTS Fl ow 5.0 <500 Reactor Vessel Sodium Level 13.0 <500 Undervoltage Relay 11.0 <230 Steam Flow 12.0 <500 i Feedwater Flow 2.5 <500 Evaporator Outlet Sodlum Temperature 12.0 <5000 Steam Drum Level 11 .0 <1000 IHX Primary Outlet Temperature 12.0 <5000 Underf requency Relay i2.0 <200 If Note that these accuracy and response times relate to the performance of the instrumentation channels from the sensors up to the signal conditioning output. In addition, as noted in Section 7.2.1.2.3, the reactor shutdown system logic, actuators and rod unlatch features require a further response time delay of 200 msecs. I 7.2-23 Amend. 67 March 1982
PSAR CRBRP - 975 Nt FFTF - 400 f%t MONJU-714 h t Section ' Reactor Trip Circuits No. Circuits Monitored 24-Pri. System 23-Pri. System -- For Trip Actuation 16-Sec. System 19-Sec. System Basic Signal and Trip Pri.-2/3 Local Pri.-2/3 Local --
*P Output Signal Logic Coincidence Logic Coincidence Logic Sec.-2/3 General Sec.-1/4 2/3 Coincidence logic Hybrid General Local Coincidence Logic 3 3 No. External Flux Monitors RSS tesic.
Max. Re nse Time 0.20Q 0.200 -- if - RSS y (From time Tsenses condition requiring trip to time when rods are released.) (Sec.) i 10. Containment 3.8 Type / Shape Single steel vessel, Single steel vessel, Single steel vessel, < cylindrical shell with cylindrical shell cylindrical shell with flat bottom and hemi- with hemi-ellipsoidal hemi-spherical top and ellipsoidal top. Con- top and bottom heads. hemi-ellipsoidal bottom. crete shielding inside, Concrete shielding Concrete cylinder i below operating floor. below operating floor. surrounds entire con-Steel containment surr- tainment. g g- 8 ue ounded by concrete con-
- 5. finement building. An
[_, a g nulus space between containment and con-m oo finement maintained at negative pressure with respect to outside a bnosphere. 18
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Is an extremely untIkely event). As with the past analyses, the following 26 conservative assumptions were made:
- 1) All full power cases are for the reactor operating at thermal hydraulic design conditions with a power generation of 975 wt at three-loop operation. (Power uncertainties are discussed in Section 4.4.3.2.)
- 2) 3 nce the sesliest Doppler coefficient occurs at the beginning-of-equilibrium cycle, the transient reactor power calculation was made for this particular phase in core life. This results in the highest possible reactor power changes being calculated. For overpcwer transients, the smallest Doppler coef ficient of all core cycles is used (see Section 4.3.2.3) and this value is reduced 30% to account for 30' uncertainties.
- 3) The highest cladding and fuel temperature hot rod occurs at the beginning of the first cycle of operation (in F/A 52 and 101).
The conservative reactor power calculation from item 2 above was applied to this particular rod. With burnup, the power generation and steady-state temperatures decrease (flows are constant) In the hottest fuel assemblies, and consequently, the temperatures, due to the transients, would decrease after beginning of cycle.
- 4) As described in Chapter 7.0 and Section 4.2, the maximum allowable time delays for PPS logic and electrical / mechanical delays have been conservatively enveloped by using a 200 millisecond delay )
between the -ty:!::! =-M?'!:r er:M; ic-- 'c.;. , "' c P I and the star of control rod insertion. t katrumt64 cWd odgM W bepd he Me levt.l
- 5) Three sigma (36) hot channel f actors were used for all the analyses and the cladding temperatures shown are the inner surf ace of the hot pin cladding at the highest temperature positloc, both axially and circumferentisily on the f uel rods. (Position is under the wire wrap.)
- 6) The most rapid flow decay af ter de-energizing the primary ptz..ps was used. (See Figure 5.3-22.)
- 7) Maximum decay heats were used for the hot rods considering 3 o~
uncertainties. Results from FORE 2M analysis are given by Figures 15.1.4-1, 2 and 3 and Table 15.1.4-1 for a 60p step reactivity insertion occuring at the worst time during the SSE (see Section 15.2 .3.3.1 ) . Comparisons of the heterogeneous core results are made with data for a homogeneous core previously reported in this ! sect i on. This previously reported data updated earIIer data for the I homogeneous core analyzed in Section 15.2.3.3. The figures show the
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15.1-106 Amend. 66 March 1982
1 k 15.3.1 Anticipated Events 15.3.1.1 Loss of Off-Site Electrical Power 15.3.1.1.1 _ Identification of Causes and Accident Description The off-site power supply to the 13.8 KV buses is available fro'n the generating switchyards and the reserve switchyard both of which are powered by outside sources as described in Chapter 8.0. Hence, the postulated 10 0 of power would' result only from simultaneous, multiple failures. The loss of all off-site power trips all primary and intennediate sodium pumps, commencing a flow coastdown. It also initiates starting of the emergency diesel generators. Action of the Plant Prctection System (PPS) trips the control rods thus limiting core over temperatures from reduced flow.' 44 l Either emergency diesel provides power to the primary and intermediate sodium pump pony motors and SGAHRS Auxiliary Feedwater Pumps for decay heat removal. Additionally, a third power supply (250 VDC Diverse Battery and Inverter) 44 provides power to the third loop pony motors. To provide conservatism in the analysis, the most rapid core flow coastdown was assured by using the minimum pump rotating kinetic energy and the maximum core flow re.eistance specified ir. the design. The action of the Primary and Secondary Shutdown Systems (SDS) are as foH ows: g33 g ,$; , ,d
- a. Primary trip - Loss of electrical power trip occurrina in 0.5 seconds. The 0.5 se'cond dela includes mea'surement and trip function lags. There ta m etod t be vestagc de.av ed inerwea+ delay ho4.n.+ % gsc.w.I cod catHeL% dete$ s.
- b. Secondary trip - Flux-Total Flow trip occurring 2 seconds after loss of electrical pumping power. This lag includes time for the flow to coastdown as well as the measurement j 1ags.
15.3.1.1.2 Analysis of Effects' and Consequences The loss of off-site electrical power event was analyzed with the DEMO computer code. The overall results of the analysis are sunnarized in Figures 15.3.1.1-1 and 15.3.1.1-2. As shown, the Primary PPS loss of electrical power trip limits the maximum core hot spot temperature to 1410*F. 51l : 52 ", In the event the prima (y shutdown system does not operate, Figure 15.3.1.1-1 shows that the secondary shutdown system limits the worst case clad hot spot temperature to 1630*F. While the transient temperature exgeedsthedesignbasisemergencytransietenvelopetemperatureby 51 l 30 F, the time above the normal operating temperature is only 6 second as r
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52 Amend. 52 l 15.3-6 Oct. 1979
permissive condition goes out of the allowable range, the protective function is automatically reinstated. The trip function will remain reinstated until the pemissive conditions are again satisfied and the operator again manually initiates the bypass. Operator manual bypass control is not effective unless the bypass comparator indicates that permissive con-ditions are satisfied. A functional diagram of the Primary and Secondary bypass pemissive logic is shown in Figure 7.2-2AA. Two loop bypasses are established under administrative control by changing the hardware configuration within the locked comparator cabinets. These bypasses are also under pemissive control such that the plant must be shutdown to establish two loop operation and if the shutdown loop is 57 activated the bypass is automatically removed.g Bypass pemissives are part of the Plant Protection System (PPS), and are designed according to the PPS requirements detailed elsewhere in this section of the PSAR. Continuous local and remote indication of bypassed instrument channels will be provided in confomance with Regulatory Guide 1.47,
" Bypassed and Inoperable Status Indication for Nuclear Power Plant Safety 15 Systems".
I N *N" Bypass features included within the Primary and Secondary RSS hardware fa N for two loop operation will be deactivated during all three loop operating modes so that the three loop operating configuraticn can not in suf 44
,be affected by these bypass features either by operator action or by j two loop hardware failure, s 1
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Amend. 57 7.2-la Nov. 1980
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1 .- 17 ST 7.6 OTHER INSTRUMENTATION AND CONTROL SYSTEMS REQUIRED FOR SAFETY (f The additional instrumentation and control systems required for safety which have not been discussed earlier in Chapter 7 are identified as the Plant Service Water and Chilled Water Instrumentation and Control Sys-tems, the F20! "-d!'n; :nd St;r:ge S:fet3 Ir,terlock C.7:;tej, and the Direct Heat Removal Service Instrumentation and Control. The Radiation Monitoring System also contains safety related components which are discussed in Chap-33l ter 11. (TM Cee! %; ef Structer:! Cen: rete 1r =dd c:: d i- Ch:pter 3A? The Nonnal and Emergency Plant Service Water and Chilled Water Systems, Fuel 26 Handling, and DHRS I and C are discussed in this Section. 7.6.1 Plant Service Water and Chilled Water Instrumentation and Control Systems 7.6.1.1 Description Those portions of the Plant Service Water and Chilled Water Systems which are required for safety, include the Emergency Chilled Water System, and the Emergency Plant Service Water System (see Sections 9.7.2,9.9.2). 33l 15 Instrumentation and control for these systems will include the necessary re-dundant instrumentation, control and indicating circuits and devices required for operation of the system. 7.6.1.2 Analys.is
, As required by IEEE Standard 279-1971, redundant monitoring and control equipment will be provided to ensure that a single failure will not impair the capability of the Emergency Plant Service Water and Chilled Water Instrumentation and Control Systems to perfonn their intended safety functions.
The systems will be designed for fail safe operation and control equipment, were practical, will assume a failed position consistent with its intended safety function. To comply with CRBRP General Design Criterion 19, adequate instru-
- mentation and control in the control room and locally, will be provided for l those portions of these systems which are required for safety.
7.6.2 DELETED I P l 44l 44
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59 Amend. 59 7.6-1 Dec. 1980 !}}