ML19210E285
| ML19210E285 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 11/26/1979 |
| From: | Andognini G, Mcguire P BOSTON EDISON CO. |
| To: | Thomas C NRC - TMI-2 BULLETINS & ORDERS TASK FORCE |
| References | |
| 79-243, NUDOCS 7912040259 | |
| Download: ML19210E285 (75) | |
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4 . BOSTON EDISON COMPANY GENERAL OFFICES 800 SOVL570N STaEET GOSTON. MaseAcNusETTs 02199
- o. cA L ANDOaNiN. November 26, 1979 S U PE RI NT EN D E N T NUCLEAR OPERATIONS DEPARTMENT BECo. Ltr. #79-243 Mr. Cecil Thomas Bulletins & Orders Task Force U.S. Nuclear Regulatory Commission Washington, D. C. 20555 License No. DPR-35 Docket No. 50-293 Bulletins and Orders Task Force Long Term Systems Info.1 nation
Reference:
(a) NRC Letter (T.A. Ippolito) to Boston Edison Company (G.C. Andognini) dated July 13, 1979, titled " Add-itional Information Required for NRC Staff Generic Report on Boiling Water Reactors". (b) Boston Edison Company letter (G.C. Andognini) to NRC (D.F. Ross) dated August 20, 1979, titled "BWR Operating Plant Owners Group General Electric Report, NEDE-24708".
Dear Mr. Thomas:
Please find enclosed, per the request of the subject task force (ref. a) the specific and comprehensive information previously deferred until November 15, 1979 (ref. b). On the pages entitled " Primary Containment Isolation System Data", no entries have been listed in the " shutdown position" and " post accident posi-tion". Since valve poseions are dependent upon the type of accident and time phase of shutdown, we feel that more clarification on the specific requirements of these questions is necessary and request that you provide this clarification at your earliest convenience. If during your review of this material you should have any questions or concerns, please do not hesitate to contact us. Very truly yours,
/
cc: P.W. Marriott 60% General Electric Co. 175 Curtner Avenue Mail Code 864 .5 San Jose, California 95125 1470 131 M Attachment JDK/gs 7 7 912040 259 W
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ATTACHMENT A PLANT PILGRIM UNIT (S) 1 BYPASS CAPACITY PLANT STEAM BYPASS CAPACITY, % RATED The turbine bypass system capacity is based on 25% of the turbine design flow. The turbine bypass system consists of three automatically and sequentially operated regulating valves mounted on a valve manifold. The manifold is connected to the main steam lines upstream of the turbine main stop valves. Each bypass valve outlet is piped to the main condenser and a pressure reducing orifice is located at the condenser connection. 1470 132
PLANT PILGRiH I SYSTEM AND COMP 0NENTS SHARED BETWEEN UNITS PAGE_ CONTINUED PAGE NONE SINGLE-UNIT PLANT CHECK HERE IAND DO NOT COMPLETE 1470 133
6 PLANT - SPECIFIC SYSTEM INFORNATION Instrumentation System Water Sources and Control Frequency of Safety Seismic Safety Seismic Safety Seismic System and Systen Classification Catetory Name Classification Category Classification Category Component Tests i CS Non-Q 11
- 1. RCIC Q I Suppression Pool l Q I See A Q
- 2. HPCI Q l CST Suppression Pool Non-Q Q
11 l O ' S
- 3. Low Press. Core Spray (LPCS) Q I c[pressinPool q ,1 q g c
- 4. LPCI (RHR) Q I Suppression Pool Q l Q I D
- 5. i.DS Q I N/A N/A N/A Q l E
- 6. SRV Q I N/A N/A N/A N/A N/A F
- 7. RHR (including Shutdown, cooling / containment spray q g Suppression Pool q g q g g steam condensing suppression Reactor Coolant pool cooling.
- 8. Cape Cod Bay !
SSW Q I (intake struc.) Q I Q l H
- 9. RBCCW Q l I" jj ', I Non-Q ll Q l i
- 10. CRDS Q l CST Non-Q II Q l J l
- 11. Condensate Storage Tank (CST) Non-Q 11 Treated Wtr. Tks. Non-Q 11 Non-Q 11 K
- 12. Main Feedwater Non-Q I CST Non-Q II Non-Q 11 L
- 13. Recirculation Non-Q 11 ( Coo L M Q I II Non-Q M Pump / Motor Coolin9 RU cc to i
s M I C~) . 1 E>
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(A) Frequency of System and Component Tests RCIC System
- a. Simulated Automatic Once/ operating Actuation Test cycle
- b. Pump Operability Once/ month
- c. Motor Operated Valve Once/ month Operability
- d. Flow Rate at 1000 psig Once/3 months
- e. Flow Rate at 150 psig Once/ operating cycle (B) Frequency of System and Component Tests HPCI System
- a. Simulated Automatic Once/ operating Actuation Test cycle
- b. Pump Operability Once/ month
- c. Motor Operated Valve Once/ month Operability
- d. Flow Rate at 1000 psig Once/3 months
- e. Flow Rate at 150 psig Once/ operating cycle
'470 135
(C) Frequency of System and Component Test Core Spray System Once/ operating
- a. Simulated Automatic cycle Actuation Test Once/ month
- b. Pump Operability Motor Operated Valve Once/ month c.
Operability Once/3 months
- d. Pump Flow Rate Each pump shall deliver at least 3600 gpm against a system head corresponding to a reactor vessel pres-sure of 104 psig (D) Frequency of System and Component Tests Low Pressure Coolant Injection Once/ Operating
- a. Simulated Automatic cycle Actuation Test Pump Operability Once/ month b.
Motor Operated Valve Once/ month c. operability Pump Flow Rate Once/3 months d. 1 (E) Frequency of System and Comoonent Tests ADS System
- 1. During each operating cycle the following tests shall be performed on the ADS:
f
- a. A sin.alated automatic actuation test shall be perfomed prior .
to startup after each refueling outage.
- b. With the reactor at pressure, 9ach relief valve shall be manually opened until a corresponding change in reactor pressure or main turbine bypass valve positions indicate that steam is flowing from the valve.
1470 136
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- 2. When it is determined that one valve of the ADS is inoperable, the ADS subsystem actuation logic for the other ADS valves and the HPCI subsystem shall be demonstrated to be operable immediately and at least weekly thereafter until the valve is repaired.
(F) Frequency of System and Component Tests Safety Relief Valves
- 1. At least one safety valve and two relief / safety valves shall be checked or replaced with bench checked valves once per operating cycle. All valves will be tested every two cycles.
- 2. At least one of the relief / safety valves shall be disassembled and inspected each refueling outage.
(G) Frequency of System and Component Test Standby Liquid Control System The operability of the Standby Liquid Control System shall be - verified by the performance of the following tests:
- 1. At least once per month each pump loop shall be functionally tested by recirculating demineralized water to the test tank.
- 2. At least once during each operating cycle:
- a. Check that the system relief falves trip full open at pressures less than 1800 psig, and reseat on a falling pressure greater than 1275 psig.
- b. Manually initiate the system, except explosive valves. Pump boron solution through the recirculation path and back to the Standby Liquid Control Solution Tank. Minimum pump flow rate of gpm against a system head of 1275 psig shall be verified.
- c. Manually initiate one of the Standby Liquid Control System loops and pump demineralized water into the reactor vessel.
This test checks explosion of the charge associated with the tested loop, proper operation of the valves, and pump operability. , The replacement charges to be installed will be selected from the same manufactured batch as the tested charge.
- d. Both systems, including both explosive valves, shall be tested in the course of two operating cycles.
1470 i37
(H) Frequency of Syscem and Component Test Containment Cooling Subsystem Containment Cooling Sub:ystem Testing shall be as follows:
- a. Pump and Valve Once/3 months Operability
- b. Pump Capacity Test After pump maintenance Each RBCCW pump shall and every 3 months deliver 1700 gpm at 70 ft.
TDH. Each SSWS pump shall deliver 2700 gpm at 55 ft. TDH.
- c. Air test on drywell and Once/5 years torus headers and nozzles (I) Frequency of System and Component Test Reactor Building Closed Cooling Water Containment Cooling Subsys'2 . Testing shall be as follows:
- a. Pump and Valve Once/3 months Operability
- b. Pump Capacity Test After pump maintenar.ce Each RBCCW pump shall and every 3 months deliver 1700 gpm at 70 ft. TDH. Each SSWS pump shall deliser 2700 gpm at 55 ft. TDH.
- c. Air test on drywell and Once/5 years torus headers and nozzles (J) Frequency of System and Component Tests Control Rod Drive System
- 1. At 16 week intervals, 50% of the control rod drives shall be tested as in la. so that every 32 weeks all of the control rods shall have been tested. Whenever 50% of the control rod drives 4 have been scram tested, an evaluation shall be made to provide ~
reasonable assurance that proper control rod drive performance is being maintained. 1470 138
- a. Following each refueling outage, each operable control rod shall be subjected to scram time t>Gts fran the fully with-drawn position. If testing is not accomplished with the nuclear system pressure above 950 psig, the measured scram insertion time shall be extrapolated to reactor pressures above 950 psig using previously determined correlations.
Testing of all operable control rods shall be completed prior to exceeding 40% rated thermal power.
- 2. Control Rod Accumulators Once a shift, check the status of the pressure and level alarms for each accumulator.
(K) Condensate Storage System Test of system and compcaents are checked during plant startup and monitored during plant operation. (L) Main Feedwater System Test of system and components are checked during plant startup and monitored during plant operation. (M) Recirculation Pump /Mortor Cooling System Test of system and components are checked during plant startup and monitored during plant operation, 1470 139
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Frequency of Instrument Test Instrument test for subject systems are grouped due to their interaction. 1.0 Reactor Building Isolation and Control System and Standby Gas Treatment System Instrumentation shall be functionally tested, calibrated and checked as indicated in Table 4.2.0 2.0 Drywell Leak Detection Instrumentation shall be calibrated and checked as indicated in Table 4.2.E. 3.0 Surveillance Information Readouts Instrumentation shall be calibrated and checked as indicated in Table 4.2.F. 4.0 Control Rod Block Actuation Instrumentation shall be functionally tested, calibrated and checked as indicated in Table 4.2.C. System logic shall be functionally tested as indicated in Table 4.2.C. 5.0 Radiation Monitoring Systems - Isolation and Initiation Functions A. Steam Air Ejector Off-Gass System Instrumentation shall be functionally tested, calibrated and checked as indicated in Table 4.2.D. System logic shall be functionally tested as indicated in Table 4.2.0. 6.0 Reactor Protection System A. Applicability Applies to the surveillance of the instrumentation and associated devices which initiate reactor scram. B. Objective To specify the type and frequency of surveillance to be applied to the protection instrumentation. C. Specification
- 1) Instrumentation systems shall be functionally tested and calibrated as indicated in Tables 4.1.1 and 4.1.2 respectively.
1470 140
- 2) Daily during reactor power operation, the peak heat flux and peaking factor shall be checked.
7.0 Protective Instrumentation A. Primary Containment Isolation Functions Instrumentation shall be functionally tested and calibrated as indicated in Table 4.2.A fvstem logic shall be functionally tested as indicated in Table 4.2.A. B. Core and Containment Cooling Systems - Initiation and Control Instrumentation shall be functionally tested, calibrated and checked as indicated in Table 4.2.8 System logic shall be functionally tested as indicated in Table 4.2.B. 8.0 Inservice Inspection Reouirenents for High Eneray Lines Outside Containment Item No. High Energy Area Insoection Method Frequency
- 1. Main steam lines outside Visual Monthly when containment from contain- operating ment to turbine stop valves
- 2. HPCI steam line in torus Visual Monthly when area and in HPCI turbine operating area
- 3. RCIC steam line in valve Visual Mor.thly when compartment and pump operating compartment
- 4. RWCU line in pump, heat Visual Monthly when exchanger compartments and operating valve compartment
- 5. Feedwater lines outside Visual Monthly when !
containment to the reactor operating - feedwater pump check valves
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e ISOLATION SIGNAL CODES FOR TABLE fignal Description A* Reactor vessel low water level - scram and close isolation valves except main steam lines. B* Reactor vessel low low water level - initiate RCIC, HPCI and close main steam line isolation valves. C* High radiation - main staam line (also causes scram). D* Line break - main steam line (steam line high space temperature or high steam flow). 32 E Reactor low low level or high drywell pressure - select LPCI and close other loop valves. F* High drywell pressure - close RER/ shutdown cooling and head spray plus the RHR to radwaste valves. G Reactor vessel low water level and low pressure; or high drywell pressure - initiate Core Spray and KRR systems. 30 J* Line break in cleanup system - high space temperature, or high flow. K* Line break in RCIC system steam line to turbine (high steam line space temperature or high steam flow) or low steam pressure. L* Line break in HPCI system steam line to turbine (high steam line space temperature or high steam flow) or low steam line pressure. Line break in RHR shutdown and head cooling (high space temperature; (f) alarm only; no auto closure). P* Low main steam line pressure at inlet to main turbine (RUN mode only). S Low drywell pressure - close containment spray valves. T Low reactor pressure permissive to open core spray and RHR-LPCI valves. U High reactor vessel pressure - close KHR shutdown cooling valves and head cooling valves. W High temperature at outlet of cleanup system nonregenerative heat exchanger. 1470 150
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Y Standby liquid control system actuated. PM* Remote manual switch from control room. Q Reactor high water level - isolate main steam line (except in run mode). X' FCIC or HPCI steam supply valve (as applicable) not fully closed.
- These are the isolation functions of the primary containment and reactor vessel isolation control system; other fu s tions are given for information only.
1470 151 e e O e
- PLANT UNIT
- PRIMARY CONTAINMENT ISOLATION SYSTEM DATA PAGE CONTINUED ON PAGE FINAL ,
ABBREVIATIONS Isolation Valve Type Isolation Signal Codes (utility supply) Engineered Safety Function Code Darameter(s) Sensed Set B = Butterfly N = NO BCK = Ball check or Group ~ for Isolation Point (units) Y = VES BL = Ball - Position Indication in Control Room CK = Check . DCV = Olaphragm D = Direct Conkol Valve , I = Indirect GB = Globe , N = None
- Others stated in Table [y = fat el f SCV = Stop Check -
Fluid SV = Solenoid A = Air VB = Vacuum Breaker S = Steam XV = Explosive W = Water Others stated in Table Others 1.tated in Table FG ' PL M FL - FLOW chtG(g Isolation Valve Location - Isblation Valve Power Source I = Inside Containment A = Air
- 0 = Outside Containment AC = AC Others stated in Table DC = DC '
61 = Iland ' Isolation Valve Actuation Mode P = Process fluid Others stated in Table $ 13 F 3 E E.S$ A = Automatic N OP = Overpressure -' W SP* SPRiMGi o RF = Reverse Flow RM = Remote Manual , Others stated in Table
- M' do AMAL'T CLoc.E.o Nsolation Valve Positions Isolation Valve Actuator A1 = As Is A0 = Air .
C = Closed M0 = Motor 0 = Open 50 = solenoid Others stated in* Table Others stated in Table
~ . _
e NOTES FOR TABLE These notes ar I .4ed by number to correspond to numbers,in parentheses, in Table
- 1. Main steam isolation vehes require that both solenoid pilota be deenergized to close valves. Accumuistor air pressure plus spring act together to close valves when both pilots are deenergized. Voltage failure at only one pGot does not cause valve closure. The vahes are designed to fully close in less than 10 seconJs, but in no less than 3 seco3ds.
- 2. Containment spray and c:ppressioc cooling valves have interlocks that allow them to be manually reopened after automatic closure. This setup permita centainment spray, for high drywell pressure conditions, and/or suppression pool cooling. When automatic signals are not prxent these valves may be opened for test or operating convenience.
- 3. Testable check vahes are designed for remote opening with zero differential pressure across the valve oeat. The valves will close. on reverse low even though the test sw.tches3 may be positioned for open. The valves o-en when pump pressure exceeds reactor pressure even though the test switch may be for close.
- 4. Control rod hydraulic lines can be isolated by the solenoid vahes outside the primary containment. Lines that extend j outside the primary containment are small and terminate in a system that is designed to prevent out. leakage. Solenoid vahes normally are closed, but they open on rod movement and during reactor eersm.
- 5. A.c motor operated valves are powered from the a.c standby power busses. D.c isolation valves are powered from the station batteries.
- 6. All motoe operated isolation vehes remain in the last position upon failure of valve power. AD air operated valves close on moti e air failm or power failure at the sof id pilots.
- 7. "Stan ad" closure rates for autorustic isolation valves refer to usual industry practice and are adequate to meet isolation requiernents.
- 8. Not used.
- 9. Vahes identified by an asterisk in the " Normal Status" column can be opened or closed by remote manual switch for operating mnvenience during any mode of reactor operation except when automatic signal is present.
- 10. Not usal.
- 11. Coincident signals "G" and "T" open core spray and eclected LPCI vahes. Special interlocks permit testing these valves by manual switch except when automatic signals are prment.
- 12. Normal statua position of valve (open or closed) is the position during normal pcwer operation of the reactor (see
" Normal Status" column).
- 13. Not need ,
- 14. Not used
- 15. Manual rwi- .averride aH sutomatic signals on the smaller valves that bypass the suppression chamber and drywell exhaust vehe
- 16. Signal "A" or "F" causes automatic withdrawal of TIP probe. When probe is withdrawn, the valve automatt y closes by mechanical action. ,
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( Table (Continued) rnTES FOR TABLE
- 17. Isolation signal "A" is permitted to close LPCI valves but cmly when the RHR shutdown cooling supply valves are not fully closed or reactor pressure (signal U) is below 75 psig. Valve position indicating lights are not required at the isolation valve dis-play paael.
32 1470 154
PLANT PILGRIM 'JNIT I DESIGN REQUIREMENTS FOR CONTAINMENT ISOLATION BARRIERS Question: Discuss the extent to which the quality standards and seismic design classification of the containment isolation provisions follow the recomendations of Regulatory Guides 1.26. " Quality Group Classifications and Standards for Water , Steam , and Radioactive-Water-Containing Components of Nuclear Power Plants", and 1.29, " Seismic Design Classification". Response: The station structures and equipment have been classified with respect to systems which must remain functional during and following the most severe natural phenomena which can be postulated to occur at this site. For the purpose of categorizing the mechanical-structural strength designs for loading conditions due to environmental events, the following definitions have been established:
- 1. Class I This class includes those stru::tures, equipment, and components whose failure or mcifunction might cause or increase the severity of an accident which would endanger the public health ant .:afety. This category includes those structures, equipment, and components required for safe shutdown and isolation of the reactor.
- 2. Class II This clac includes these structures, equipment, and components "hich are important to reactor operation, but are not assential for preventing an accident which would endanger if.e public health and safety, and are not essential for the mitigation of the consequences of these accidents. A Class II designated item shall not degrade the integrity of any item designated Class I.
The only exception of these two definitions is that a system whose failure or malfunction might increase the severity of an accident is not designed to withstand the effects of a tornado if the failure of the system will not cause an accident. The reason for making this exception is that ! the probability of the occurrence of a design basis loss-of-coolant accident or a design basis tornado during the life of a plant is small, therefore, the probability of the simultaneous occurrence of these two independent events is vanishingly small. 1470 155 .
PLANT Pilgrim UNIT (S) I NORMAL OPERATING MODES AND IS01ATION MODES The secondary containment system is designed to withstand the maximum postulated seismic event and be capable of providing hold-up, treatment and an elevated release point for any fission products released to it. In addition, the reactor building is designed to provide protection for the engineered safeguards and nuclear safety systems located in the building frcm all postulated environmental events including tornadoes. Containment Isolation Valves The basic function of all primary containment isolation valves is to provide necessary isolation to the containment in the event of accidents or similar critical conditions when the free release of containment atmosphere cannot be permitted. The primary containment isolation valves are grouped into four basic classes. Variations to the basic isolation valve definitions are used in certain circumstances. These valves are generally located in instrument lines or in core standby cooling system lines which may be required to be operational when primary containment is required. Class A valves are on process lines that communicate directly with the reactor vessel and penetrate the primary containment. These lines require two valves in series, one inside the primary containment and one outside the primary containment. They are located as close to the primary containment boundary as practical. Except in the case of check valves, both valves shall close auto-matically on isolation signal. Both valves shall receive the isolation (closure) signal even if normally closed during reactor operation. Since check valves close on reverse process flow, they are used to isolate some incoming lines. Testable check valves are used on selected process inflow lines where floa is expected to be zero or on lines which have low flow with intermittent use during normal station operation. All Class A valves except check valves are capable of remote manual control from the control rocm. Class B valves are on process lines that do not directly com-municate with the reactor vessel, but penetrate the primary containment free space. These lines require two valves, in series, both of them located outside the primary containment ! and as close to the primary containment boundary as practical. - Except in the case of check valves, both valves close automati-cally on isolation signal. Both valves receive the isolation closure signal even if normally closed during reactor operation. All Class B valves except check valves are capable of remote manual control from the control room. Class C valves are on process lines that penetrate the primary containment but do not communicate directly with the reactor vessel, with the primary containment free space, or with the environs. Class C lines require only one valve which closes automatically by process action (i.e., reverse flow) or by remote manual operation from the control room. 1470 156
PLANT Pilgrim UNIT (S) I_ _ NORMAL OPERATING MODES AND ISOLATION MODES Motive power for the valves on process lines which require two valves shall be puysically independent saurces to pro-vide a high probability that no single accidental event could interrupt motive power to both closure devices. Automatic isolation valves, in the usual sense, are not used on the inlet lines of the reactor core and containment cooling systems, and reactor feedwater systems, since opera-tion of these systems is essential following a design basis loss-of-coolant accident. Since normal flow of water in these systems is inward to the reactor vessel or to the primory containment, check valves located in these lines will provide automatic isolation, if necessary. No automatic isolation valves are provided on the control rod drive system hydraulic lines. These lines are isolated by the normally closed hydraulic system control valves located in the reactor building, and by check valver com-prising a part of the drive mechanisms. TIP lines and small diameter instrument lines are not provided with automatic isolation valves. TIP system guide tubes are provided with an isolation valve which closes automatically upon receipt of proper signal and titer the TIP cable and fission chamber have been retracted. In series with this isolation valve, an additional or backup isola tion shear valve is included. Both valves are located outside the drywell. The function of the shear valve is to assure integrity of the containment in the unlikely event that the other isolation valve should fail to close or the chamber drive cable should fail to retract if it should be extended in the guide tube during the time that containment isolation is required. This valve is designed to shear the cable and seal the guide tube upon an actuation signal. Valve position (full open or full closed) of the automatic closing valves will be indicated in the control room. Each shear valve will be operated independently. The valve is an explosive type valve and each actuating circuit is monitored. In the event of a containment isolation signal, the TIP system recieves a command to retract the traveling probes. Upon full retraction, - the isolation valves are then closed automatically. If a - traveling probe were jammed in the tube run such that it could not be retracted, instruments would supply this informa-tion to the operator, who would in turn investigate to determine if the shear valve should be operated. Effluent lines such as main steam lines which connect to the reactor vessel or which are open to the primary contairment have air-powered valves. This arrangement provides a high reliability with respect to functional performance. These valves are closed automatically. 1470 157
The Primary Containment System (Drywell, Suppression Pool, Isolation Valves, and Containment Penetrations are Class I. Additionally the Secondary Containment System (Rx Bldg., Standby Gas Treatment, Main Stack and Reactor Bldg. Isolation Control System) is Class I. All isolation associat5d with the NSSS are Class I. The Seismic I Q-List includes electrical equipment and instrumentation required for a Class I system's operation to meet the Class I criteria. Cables, cable pulls, and associated raceways required for safeguard and isolation systems are Class I. The Boston Edison Quality Assurance Program for Operation of Nuclear Power Plants is based on the understanding that each item in the plant can be determined to belong in one (1) of three (3) categories to which different quality re-quirements apply. These quality categories are identified in Exhibit I. Systems, structures, and components designated as safety related are identified on a Q-List. Provisions have been established for maintaining and controlling the Q-List. The determination of safety related items was accomplished by use of the following definitions and PNPS FSAR Appendix G and FSAR Section 14.
- 1. Safety Related Function Any function that is necessary to assure (1) the integrit.' of the reactor coolant pressure boundary, (2) the capat.ility to shut down the reactor and maintain it in a sa4 shutdown condition, or (3) the capability to prev y ;r mitigate the consequences of accidents that could result in potential off-site exposures comparable to the guideline exposures of 10 CFR Part 100.
- 2. Safety Related Systems, Structures and Components Those systems, structures, and components that have safety related functions.
Additional information applicable to this question may be : found elsewhere in this response. . 1470 158
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Quali ty Ca tegory Description e. (Q) Safety related systems, striictures, and components and those other systems. --
.- structures, a id components requiring BECo selected QA program element application -
; to meet 10 CFR 50, Appendix B, requirements. Included in this category are Code items, regardless of Code class, fabricated or installed unde Section III or
> 4 maintained under Section XI of the ASitE Code. (R) Systems, structures, and components deemed essential for reliable electric power . generation to which BECo selected QA program elements have been applied,although not necessary to meet 10 CFR 50, Appendix B, requi'ements. Included in this category are Code items maintained under Section V II of the ASME Code.
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PLANT Pilgrim UNIT (S) I PROVISIONS FOR TESTING Question: Discuss the design provisions for testing the operability of the isolation valves. Response: All essential parts of the Primary Containment and Reactor vessel Isolation Control System are testable during the reactor operation. Isolation valves can be tested to assure that they are capable of closing by operating manual switches in the control room and observing the position lights and any associated process effects. Testable check valves are arranged to verify that the valve disk is free to open and close. The channel and trip system responses can be func-tionally tested by applying test signals to each channel and observing the trip system response. The Main Steam Isolation Valves a*e tested (test buttons provided) twice/ week as required by Tech Spec. 4.7.F.D.I.C. Each valve is exercised from the co.1 trol room to the " Closed to 90%" position. Once per quarter, with Reactor power <50%, each MSIV is individually tripped and closure time is verified. Testable check valves are to be tested (although there is no Technical Specification requirement) only during an outage (once per cycle) when access to the drywell is pennitted. RHR, core spray, and HPCI testable checks have test buttons and position indicators on panel 903. RCIC System testable check valves have similar capabilities on panel 904. Acceptance criteria is that each valve cycle on demand and no unexplain-ed descrepancies are found. Testable check valve bypass valves are tested to open on initiation of the te3t to equalize pres-sure across the check valve. Othe.' primary containment isolation valves are required to be tested once/3 months. Remote manual switches are provided in the control room. Each valve is cycled and time to change state is checked against maximum allowable closing time. Feedwater check valves, control rod hydraulic return check valves, standby liquid control system check valves are assumed operable unless known to be failed. 1470 160 LDR/fej
PLANT PILGRIM UNIT I_ CODES, STANDARDS, AND GUIDES qupstion: Identify the codes, standards, and guides applied in the design at the containment isolation system and system components. Response: The initiation logic for the automatic closure at the primary containment isolation valves meet the intent of the requirements of IEEE - 279. Ali other applicable codes, standards, and guides are found in PNPS I, FSAR, Amendment 30, Appendices A and H. The pressure suppression containment system's material, design, fabrication, inspection and testing are in accordance with ASME Boiler and Pressure Vessel Code, Section III, subsections B (1967 edition) with all applicable addenda published to June, 1967 and Code Case 1177-5 and 1330-1. 1470 161
PLANT Pilgrim UNIT (S) I NORMAL OPERATING MODES AND ISOLATION IODES Question: Discuss the normal operating modes and containment isolation provision and procedures for lines that transfer potentially radioactive fluids out of the containment. Response: General The containment systems of Pilgrim Nuclear Power Station utilize a "multibarrier" concept which consists of two systems. The primary containment system is a pressure suppression system which forms the first barrier. The secondary containment system is a system which minimizes the ground level release of airborne radioactive materials and forms the second barrier. The fuel, fuel cladding, and reactor primary system form additional barriers to the release of fission products. Primary Containment Systems The primary containment system houses the reactor vessel, the reactor coolant recirculation system and other branch connections of the reactor coolant system. The primary containment is a pressure suppression system consisting of a drywell, pressure suppression chamber which stores a large volume of water, a con-necting vent system between the drywell and water pool, isolat i on valves, vacuum relief system, containment cooling systems, anu other service equipment. The primary containment system is designed to withstand the forces from any size breach of the nuclear system primary barrier up to and including an instantaneous circumferential break of the reactor recirculation piping and provides a hold-up time for decay of any radioactive material released. The primary containment system also stores sufficient water to condense the steam released as c result of a breach in the nuclear system primary barrier and to supply the core standby cooling systems. Secondary Containment System The secondary containment system encloses the primary containment system, the refueling and reactor servicing areas, new and spent fuel storage facilities and other reactor auxiliary systems. The , secondary containment system serves as the only containment during , reactor refueling and maintenance operations, when the primary containment is open, and as an additional barrier when the primary containment system is functional. The secondary containment system consists of the reactor building, standby gas treatment system, main stack, reactor building isolation and control system, and other service equipment. 1470 162
PLANT Pilgrim UNIT (S) I NORMAL OPERATING MODES AND ISOLATION MODES Lines, such as those of the reactor building closed cooling water cystem which do not connect to the reactor primary system or open into the primary containment, are provided with at least one a-c powered valve on the effluent line and a check valve on the influent line. Instrumentation piping connecting to the reactor primary system which leaves the primary containment is dead-ended at instruments located in the reactor building. These lines are provided with flow limiting orifices, manual isolation valves and excess flow check valves. The control rod hydraulic system is provided with three valves which can be utilized for isolation purposes. The first is a ball check valve which comprises an internal portion of the control drive mechanism. The other valves are normally closed hydraulic system control valves located in the reactor building. d 1470 Ib3
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TABLE 4.1 1 - - REACTOR PROTECTION SYSTEM (SCRAM) INS JMENTATION FUNCTIONAL TESTS MINIMUM FUNCTIONAL TEST FREQUENCIES FOR SAFETY INSTR. AND CONTROL CIRCUITS Croup (2) Functional Test Minimum Frequency (3) Mode Switch in Shutdown A Place Mode Switch in Shutdown Each Refueling Outage Manual Scram A Trip Channel and Alarm Every 3 Mor.ths RPS Channel Test Switch (5) A Trip Channel and Alcrm Each Refueling Outage IRM liigh Flux C Trip Channel and Alarm (4) Once Per Week During Refueling and Before Each Startup Inoperative C Trip Channel and Alarm Once Per Week During Refueling and Before Each Startup APRM tilgh Flux Inoperative B Trip Output Relays (4) Once/Weet (7) Downscale B Trip Output Relays (4) Once/ Week Flow Bias B Trip Output Relays (4) Once/ Week liigh Flux (15%) B Calibiate Flow Blas Signal Once/ Month (1) B Trip Output Relaya (4) Once Per Week During Refueling and Before Each Startup High Reactor Pressure A Trip Channel and Alarm (1) liigh Drywell Pressure A Trip Channel and Alarm (1) Reactor Low Water Level (6) A Trip Channel and Alarm (1) liigh Water Level in Scram Discharge Tank A Trip Channel and Alarm >N Every 3 Months Turbine condenser Low Vacuum q A Trip Channel and Alarm (1) Main Steam Line liigh Radiation B Trip Channel and Alarm (4) Once/ Week
> x.
gg Main Steam Line Isolation Valve Closure A Trip Channel and Alarm
/)1 N
CD Turbine Control Valve Fast Closure A Trip Channel and Alarm (1) Turbine First Stage Pressure Permissive A Trip Channel and Alarm e w Turbine Stop Valve Closure A Trip Channel and Alarm Every 3 Months g (1) Reactor Pressure Permissive A Trip Channel and Alarm Every 3 Moni.h4
NOTES FOR TABLE 4.1.1 ( l. Initia11g;once 2.0 x 10 per ter, thereaf month until exposure according (M4.1.1 to Figure as defined with an on Figure.aot interval 4.1.1) is less than one month nor more than three months. The compilation of instrument failure rate data may include data obtained from other boiling water reactors for which the same design instrument operates in an environment similar to that of PNPS.
- 2. A description of the three groups is included in the Bases of this Specification.
- 3. Functional tests are not required when the systems are not required to be operable or are tripped.
If tests are missed, they shall be performed prior to returning the systems to an operable status.
- 4. This instrumentation is exempted from the instrument channel test definition.
This instrument char _nel functional test will consiet of injecting a simulated electrical signal into the measurement channels.
- 5. Test RPS channel after maintenance.
- 6. The water level in the reactor vessel will be perturbed and the corresponding level indicator changes will be monitored. This perturbation test will be performed every month after completion of the monthly functional test program.
- 7. This APRM testing vill be performed once per week when in the run mode. If the reactor is out of the run mode for more than one week, the testing will be performed as soon as practicable after returning to the run mode.
1470 196
TABLE 4.1.2 . REACTOR PRUIECTION SYSTEM (SCRAM .NSTRUMENT CALIBRATION ~ MINIMUM CALIBRATION FREQUENCIES FOR REACTOR PROTECTION INSTRUMENT CilANNELS Instrument Channel Group (1) Calibration Test (5) Minimum Frequency (2) _ IRM Iligh Flux C Comparison to APRM on Controlled Note (4) Shutdowns APRM liigh Flux Dutput Signal B lleat Balance Once every 3 Days Flow Bias Signal B Internal Power and Flow Test Each Refueling Outage LPRM Signal B TIP System Traverse Every 1000 Effective Full Power llours liigh Reactor Pressure A Standard Pressure Source Every 3 Months liigh Drywell Pressure A Standard Pressure Source Every 3 Months Reactor Low Water Level A Pressure Standard Every 3 Months liigh Water Level in Scram Discharge Volume A Note (6) Note (6) Turbine Condenser Low Vacuum A Standard Vacuum Source Every 3 Months Main Steam Line Isolation Valve Closure A Note (6) Note (6) Main Steam Line liigh Radiation B Standard Current Source (3) Every 3 Months Turbine First Stage Pressure Permissive A Standard Pressure Source Every 6 Months Turbine Control Valve Fast Closure A Standard Pressure Source Every 3 Months Turbine Stop Valve Closure A Note (6) Note (6) Reactor Pressure Permissive A Standard Pressure Source Every 6 Months N CD N
NOTES FOR TABLE 4.1.2
- 1. A description of three groups is included ir: the bases of this Specification.
- 2. Calibration tests are not required when the systems are not required to be operable or are tripped.
- 3. The current source provides an instrument channel alignment. Calibration using a radiation source shall be made each refueling outage.
- 4. Maximum frequency required is once per week.
- 5. Response time is not a part of the routine instrument channel test, but will be checked once per operating cycle.
- 6. Physical inspection and actuation of these position switches will be performed during the refueling outages.
Id ) l9 e
PNPS TABLE 4.2.A MININUM TEST AND CALIBRATION FREQUENCY FOR PCIS Instrument Channel (5) Instrument Functional Test Cajibration Frequency Instrument Check
- 1) Reactor liigh Press.ure (1) Once/3 months None
- 2) Reactor Low-Low Water Level (1) Once/3 months Once/ day
- 3) Reactor liigh Water Level (1) Once/3 months once/ day
- 4) Main Steam liigh Temp. (1) Once/3 months None
- 5) Hain Steam liigh Flow (1) Once/3 months None
- 6) Main Steam Low Pressure (1) Once/3 months None
- 7) Reactor Water Cleanup Iligh Flow (1) Once/3 months Once/ day
- 8) Reactor Water Cleanup Iligh Temp. (1) Once/3 months None Logic System Functional Test (4, (6) Frequency
- 1) Hain Steam Line Isolation Vvs. Once/6 months Main Steam Line Drain Vvs.
Reactor Water Sample Vvs.
- 2) RilR - Isolation Vv. Control Once/6 months Shutdown Cooling Vvs.
IIead Spray
' Discharge to Radwaste
- 3) Reactor Water Cleanup Isolation once/6 months
- 4) Drywell Isolation Vvs. Once/6 months TIP Withdrawal Atmospheric Control Vvs.
Sump Drain Valves __. 5) Standby Cas Treatment System Once/6 months 43, Reactor Building Isolation N CD 4-W -- --- 1.
PNPS TABLE 4.2.B MINIMUM TEST AND CALIBRATION FREQUENCY FOR CSCS Instrument Channel ~ Instrument Functional Test Calibration Frequency Instrument Check
- 1) Reactor Water Level (1) Once/3 months once/ day
- 2) Drywell Pressure (1) Once/3 months None
- 3) Reactor Pressure (1) Once/3 months None
- 4) Auto Sequencing Timers NA Once/ operating cycle None
- 5) ADS - LPCI or CS Pump Disch.
Pressure Interlock (1) Once/3 months None
- 6) Undervoltage Relays (1) Once/ operating cycle None
- 7) Trip System Bus Power Monitors Once/ operating cycle NA Once/ day
- 8) Recirculation System d/p (1) Once/3 months once/ day
- 9) Core Spray Sparger d/p NA Once/ operating cycle Once/ day
- 10) Steam Line liigh Flow (IIPCI & RCIC) (1) Once/3 months None
- 11) Steam Line liigh Temp. (llPCI & RCIC) (1) Once/3 months None
- 12) Safeguards Area liigh Temp. (1) Once/3 months None
- 13) IIPCI and RCIC Steam Line Low Pressure (1) Once/3 months None
__. 14) IIPCI Suction Tank Levels (1) Once/3 months None J>
- -J Q
N O O
PNPS TABLE 4.2.B HINIMUM TEST AND CALIBRATION FREQUENCY FOR CSCS Logic System Functional Test (4) (6) F,requency Remarks
- 1) Core Spray Subsystem Once/6 months
- 2) Low Press. Coolant Injection Subsystem Once/6 months
- 3) Containment Spray Subsystem once/6 months
- 4) IIPCI Subsystem Once/6 months
- 5) IIPCI Subsystem Auto Isolation Once/6 months
- 6) ADS Subsystem Once/6 months
- 7) RCIC Subsystem Auto Isolation Once/6 months
- 8) Diesel Generator Initiation once/6 months
- 9) Area Cooling for Safeguard System Once/6 months N
O N O e e
PNPS TABLE 4.2.C MINIMUM TEST AND CALIBRATION FREQUENCY FOR CONTROL ROD BLOCKS ACTUATION Instrument Channel Instrument Functional Calibration Instrument Check Test
- 1) APRM - Downscale (1) (3) Once/3 months Once/ day
- 2) APRM - Upscale (1) (3) Once/3 months Once/ day
- 3) IRM - Upscale (2) (3) Startup or Control Shutdown (2)
- 4) IRM - Downscale (2) (3) Startup or Control Shutdown (2)
- 5) RBM - Upscale (1) (3) Once/6 months once/ day
- 6) RBM - Downscale (1) (3) Once/6 months , once/ day
- 7) SRE: - Upscale (2) (3) Startup or control Shutdown (2)
- 8) SRM - Detector Not in Startup Position (2) (3) Startup or Control Shutdown (2)
- 9) IRM - Detector Not in Startup Position (2) (3) Startup or control Shutdown (2)
Logic System Functional Test (4) (6) Frequency (1) System Logic Check Once/6 months amb
- J CD N
O N
PNPS TABLE 4.2.D MINIMUM TEST AND CALIBRATION FREQUENCY FOR RADIATION MONITORING SYSTEMS Instrument Channels Instrument Functional Calibration Instrument Check (2) Test
- 1) Refuel Area Exhaust Monitors - Upscale (1) Once/3 months once/ day
- 2) Refuel Area Exhaust Monitors - Downscale (1) Once/3 months Once/ day
- 3) Of f-Gas Radiation Monitors (1) Once/3 months Once/ day Logic Sys[gp, Functional Test (4) (6) Frequency
- 1) Reactor Building Isolation Once/6 months
- 2) Standby Gas Treatment System Actuation Once/6 months
- 3) Steam Jet Air Ejector Of f-Gas Line Isolation Once/6 months
~J O N O w
PNPS TABLE 4.2.E HINIMUM TEST AND CALIBRATION FREQUENCY FOR DRYWELL LEAK DETECTION Instrument Channel Instrument Functional Test Calibration Frequency Instrument Cs.tek
- 1) Equipment Drain Sump Flow Integrator (1) Once/3 months Once/ day
- 2) Floor Drain Sump Flow Integrator (1) Once/3 months once/ day
- 3) Air Sampling Syster. (1) Once/3 months Once/ day e
g -M N CD N CD 4
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