ML19250C444

Forwards Info for NRC Generic Rept on Effects of TMI Accident on Bwrs,Requested by NRC

ML19250C444
Person / Time
Site:	Big Rock Point File:Consumers Energy icon.png
Issue date:	11/16/1979
From:	Hoffman D CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:	Thomas C NRC - TMI-2 BULLETINS & ORDERS TASK FORCE
References
NUDOCS 7911260132
Download: ML19250C444 (17)

Text

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Povver Company

~~ g)ifirin General Off.ces: 212 West Michigan Avenue. Jackson, Michigan 49201

• Area Code 517788-0550 November 16, 1979 Director, Nuclear Reactor Regulation Att Mr C 0 Thomas Bulletins and Orders Task Force US Nuclear Regulatory Commission Washington, DC 20555 DOCKET 50-155 - LICENSE DPR BIG ROCK POINT PLANT - RESPONSE TO REQUEST FOR INFORMATION FOR NRC STAFF GENERIC REPORT ON EFFECT OF THREE MILE ISLAND ACCIDENT ON BWRs NRC letter dated July 17, 1979 requested information relative to Big Rock Point for use in preparing a generic report on the effects of the Three Mile Island (TMI) accident on BWRs. A portion of this information was included in General Electric Report NEDO-24708, " Additional Information Required for NRC Staff Generic Report on Boiling Water Reactors" submitted August 17, 1979.

Pursuant to agreements reached during a meeting between the NRC staff and the GE BWR Owners' Group on July 11 and 12, 1979, portions of the information requested related to plant systems were to be submitted by November 16, 1979. The GE BWR Owners' Group has concluded that the systems-related information required by November 16, 1979 is plant specific and should be submitted individually by each licensee. Accordingly, this information for Big Rock Point is submitted as an attachment to this letter. David P Hoffman (Signed) l}g David P Hoffman Nuclear Licensing Administrator CC DLZiemann, USNRC JGKeppler, USNRC 7011260 / 23 $1 l'

PLANT B1G ROCK POINT PLANT-SI'ECIFIC SYSTEM INF01GiAT10N General Water Sources Instrumentation and Control Frequency Safety Seismic Safety Seismic Safety Seismic of System and Sys t,em Classif. Category Classif. Catego_ry Classif. Category Component Test,s 1. RCIC 2. Isolation Condenser A,B I N 1 lE I 1 yr 3 IIPCS 4. IIPCI 5 LPCS A,B I B I IE I E 6. LPCI '/. ADS /RDS A I N/A N/A lE I 7, 30, 90, R, CS 8. SRV A 1 N/A N/A N/A N/A R 9 EllH a. shutdown cooling N N N N N N N b. steam condensing c. uuppression pool cooling B I B I IE 1 30, R d. cont,ainment spray modes B I B I lE I R 10. SSW N N N N N N N 11. RECCW N N N N N N N _ 12. CHDS A I N N lE I 7, 30, R, S U 13 CST N N N N N N N sa a 14. Main Feedwater B I N N 1E 1 N 15 Reci rcula tion N N N N N N R g lbmp/t/.otor Cooling y CD

• '/= 'l day, 30= 30 day, 90= 90 day, R= Refueling Outuge, S= Shutdown, CS = cold Shutdown The above safety classifications were determined by comparing the system function with the requirements of Reg Guide 1.26 and 1.29 This is not to imply that the syster..a

..~ clesigned and built to the standards required by the classification as noted in these guides. Dueing the cocrse of the Systematic Evaluation Progrtun detenninations will be made as to the seismic and gi alit: adequac;,- of these systems.

PLANT BIG ROCK POINT BYPASS CAPACITY Plant steam bypass capacity, % rated lq0. SYSTEMS AND COMPCNENTS SHARED BETWEEN UNITS Big Rock Point is a single unit Plant. 1394 276

PLANT BIG ROCK POINT PRIMARY CONTAINMENT ISOLATION SYSTEM DATA PAGE 1 CONTINUED ON PAGE 2 ~ Isolation Valves c a S Positions p c 8 U M Y Ug 8 3 3 M S* 5 E 8 E Oc $O 11 M M g c O$ "o 8% Rb8 3 8 k y 5 Ae a 83 9 2 8 R*3 .U .ht % "$j u 3" 2 e e uk 3 5 " fo y b "1 o S a j; t 8 N? g5 t

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g NK@ b 00k N N ONo 0 N N wA S SS E SN ?? m 11 - 7 21 " Ventilation Supply Y A A CV-4096 1&2 0 CK A0 A IN 6 DC D 0 0 C C 4 CV 1097 B 4 II-8 24" Ventilation Exhaust Y B A CV-40914 1&2 0 CK AO A RM 6 DC D 0 0 C C CV 1095 B 4 11 - 9 7 3/4" Ventilation Probe Y C A SV-9155 1&2 0 SV SO A RM DC N O O C C SV-9156 11-2 1 2" Clean Smnp Disch. Y D W CV-4102 3&4 O DCV AO A RM AC D 0 0 C C CV-4031 I 11 - 1 5 C" Dirty Sump Disch. Y E W CV-4103 3&4 0 DCV AO A RM AC D 0 0 C C CV-4025 3&4 I DCV AO A EM AC D O O C C VEC-305 I CK RF N C C C C 2 0 2" Instrument Air N F A VA-305 1 CK RF N O O O C 11 - 2 5 2" Service Air N G A VA-303-1 CK RF N O O O C H-18 2" Demin. Water Y ll W CV 1105 O DCV AO RM AC N O O C C 8 MK-253F I CK RF N C C C C N N 11 - 1 0 12" Main Steam Y I S MO-7050 3&4 I GT MO A RM 60 DC D 0 C C C 11 - 3 7 1[" Main Steam Drain Y J W MO-7065 1 GT M0 RM DC D C C C C 11 - 1 1 10" Reactor Feed Y K W VFW-9 I SCV RF 0C C C VFW-304

PLANT DIG ROCK POINT PRTIORY CONTAINhENT ISOLATION SYSTEM DATA PAGE 2 CONTINUED ON PAGE 3 Isolation Valves C C o a Positions 6 Y M %g 8 3 3 S jc E" 3 4 8 E c0 GE M g E!$ "o ^e c 8% a 3U8 3 8 8 na e 8h 9 y 5 e y;: a "s .ht B 9, .D s% "a.g a" S u

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a u$ D ! !: es u ase 8 s e - :: :a N K :iE b 005 $ Y ONo 0 h N c2 0 NO ru n R 80 E RE.? ? H-35 2" Control Rod Drive flyd. Y L W VRD-310 O CK RF 0 0 0 C Poppets I CK RF 0 0 0 C 11 - 3 6 6" Core Spray Y M U VPI-302 I CK RF 0 0 3/9 0 H-23 13" Resin Sluice Y N W CV-4091 3&4 I BL AO A RM AC D C C C C CV-4092 CV-4093 H-17 3" Treated Waste Y O W CV-4049 1 DCV AO 104 AC D C C C C VRW-313 1 CK RF C C C C 11 - 2 2 2" Reactor and Fuel Pit Y P W CV-4027 3&4 I r :V AO A RM AC D C C C C Drain CV-4117 0 11 - 2 7 4" Backup Core Spray Y Q W VPI-301 1 CK RF 0 0 O/C O w 4 m .N NW

PIANT BIG ROCK FOINT PRIMARY CONTAINMENT ISOLATION SYSTEM DATA PAGE 3 CONTINUED ON PAGE Final ABBREVIATIONS Engineered Safety Function Isolation Valve Type Isolation Signal Codes (Utility Supply) N = NO B = Butterfly Code Parameter (s) Sensed Set Y = YES BCK = Ball Check or Group for Isolation Point (units ) BL = Ball 1. All scrams Position Indication in Control Room CK = Check D = Direct DCV = Diaphragm 2. New fuel & spent 15 mr/nr I = Indirect Control Valve fuel storage high N = None GB = Globe radiation Others stated in Table GT = Gate RV = Relief 3. High containment 1.5 psig Fluid SCV = Stop Check pres sure A = Air SV = Solenoid S = Steam VB = Vacuum Breaker 4. Low reactor water 2 '-9" above W = Water XV = Explosive level core Others stated in Table Others stated in Tabl'e Isolation Valve location Isolation Valve Power Source I = Inside Containment A = Air O = Outside Containment AC = AC Others stated in Table DC = DC H = Hand Isolation Valve Actuation Mode P = Process Fluid A = Automatic Others stated in Table ~ OP = Overpressure b 16 = Rever:e Flow hM = Rewote Manual M = Manual Others stated in Table n. N sg) Isolation Valve Positions 1 solation Valve Actuator AI = As Is AO = Air C = Closed MO = Motor O = Open SO = Solenoid Others stated in Table Others stated in Table

PLANT BIG ROCK POINT DESIGN REQUIREMENTS FOR CCNTAINMENT ISCLATION BAR'.IERS II. Discuss the design requirements for the containment isolation barriers regarding: a. The extent to which the quality standards and seismic design classi-fication of the containment isolation provisions follow the recommend-ations of Regulatory Guides 1.26, " Quality Group Classifications and Standards for Water, Steam, and Radioactive-Water-Containing Com-ponents of Nuclear Power Plants," and 1.29, " Seismic Design Classi-fication". Response: Big Rock Point Plant began ecmmercial operation in December of 1962. This was prior to the advent of todays standards, codes and guides and in particular Regulatory Guides 1.26 and 1.29, which were first issued as safety guides in 1972. The containment isolation valves and piping system are considered as Quality Group B and Seismic Category I for purposes of classifying the system. This, however, does not mean these systems have been designed and constructed to the standards and codes referenced within the Regulatory Guides 1.26 and 1.29 The designation of the containment isolation system provides a basis for conducting inspections, tests and analysis of the systen. !$94 280

PLANT BIG ROCK POINT PROVISIONS FOR LEAK DETECTION III. Discuss the provisions for detecting leakage from a remote manually controlled system (such as an engineered safety feature system or essential line) for the purpose of determining when to icolate the affecc-ed system or system train. Specify the parameters sensed, their set point, and procedure for initiation of containment isolatien. Response: Leakage detection from the primary coolant system within cen-tainment or leakage to outside the containment building can be detected in several ways. There is no specific leak detection system associated with BRP, however, adequate means to detect leakage by level, flow, temperature, dew point, and radiation monitoring are available and are expanded en below. In addition a daily Technical Specification requirement for primary coolant system identified and unidentified leakage exists. Daily, while at power and ' y procedure, a calculation is made to determine the leak o rate by observing sump levels and containment moisture content. Additional observation of sump levels via an alam in the control room and sump pump running times in the control room are also indications of leakage in containment. The sumps discharge to radwaste wh,ere an abnomal increase in tank levels can be detected. Te=perature and dew point monitoring alarms in the control room also exist. Five areas in the containment building and cne in the condenser pipe tunnel alam from one recorder. The areas and set points are: Temeerature Dew Point 1. Personnel Lock Area 100 F 80"F 2. New Fuel Storage Area 100 F 80 F 3 Emergency Condenser Area 100 F 80 F h. Sphere Exhaust Air 105 F 80 F 5 Sphere Inlet Air 120 F 120 F 6. Condenser - Pipe Tunnel 120 F 120 F A pipe tunnel steam leak alam at 90 F dew point and the condenser - pipe tunnel alam frcm the above mentioned recorder are both indications of leakage outside the containment building. 1394 281

In addition to the alarm frcm the recorder another control room alarm indicates sphere vent system trouble. The response is to check other indications and the local panel in containment which indicates among other things, containment pipeway high temperature and pipeway steam leak. A dew cell recorder is also present on the local panel. Constant air monitors provide indication locally and in access control of containment and turbine building radiation levels. An exhaust air cam monitors radioactive iodine and particulate air leaving containment. Radiation area monitors are located throughout the plant and alarmed in the control room. An emergency condenser vent radiation monitor would detect primary system leakage into the emergency condenser via a tube leak. Liquid process monitoring of the reactor cooling water (RCW) shows any leakage into the system from the primary system or other more radioactive systems. Monitoring of the RCW Tank also indicates cross leakage from other systems. Liquid process monitoring of the service water will indicate contamination from the RCW system as the service water leaves containment. Stack gas monitoring which would detect airborne radiation from the containment or turbine building is an additional means of leak detection. System specific leak detection is generally not available except for the monitoring mentioned above. But the core sprays and enclosures sprays and reactor recirculation pump seals all have high flow alarms. They are 500,100, and 1 gpm respectively. These high flow alarms are all in the control room and all are indications of system leakage. The core and enclosure sprays, however, are to provide the operator with information to show a possible failure of the system piping which may result in in-adequate sprays to the reactor or containment. All the leak detection methods mentioned above provide the operator with information. The severity of the problem then can be assessed by this information and other variables and corrective action taken to correct the problem. 1394 282

PLANT SIG ROCK POINT PROVISIONS FOR TESTING IV. Discuss the design provisions for testing the operability of the isolation valves. Reference Documents: Letter CPCo to NRC dated, Septembez ., 1975 Letter CPCo to NRC dated, November 6,1975 Letter CPCo to NRC dated, February 13, 1976 The following lines, penetrating containment, are c.1osed systems inside containment for which no typ C tests are performed (2-13-76 letter). (Type C tests per 10CFR 50 Appendix J.) 1. Reference vol the pressure sensing line. 2. Instrument air line. 3 Service air line. 4. Service water line. 5 Reactor building heating steam system. 6. Air operating lines to CV-4040, CV-kll4 and Cv-ho29 7 Shutdown fDishing line. (This line ha: been capped since sub-mittal of the 2-13-76 letter). The core spray recirculation system (3 penetrations) is a closed system outside containment for which no type C test is run. Leakage to the atmos-phere is tested during the containment integrated leak rate test (ILRT). The primary and backup core spray lines are also not tabjected to type C tests. Capped sphere penetrations are type A tested, i.e., during the ILRT. The containment sphere pressure sensing lines are also type A tested. Since there is but one Main Steam Isolation valve with a drain line parallel to it, these valves are tested hydraulically during primary system hydro-static tests and during the ILRT, other than the above mentioned exceptions, the leak rate testing of containment isolation valves is conducted en the basis of 10CFR 50 Appendix J requirements. Where mentioned, valves to be tested as part of the ASMI B&PV code section XI subsection IWV will be incorporated in a testing program in the near future. A discussion of the test provisions follows. Refer also to diagrams for arrangement of isola-tien valves. Ventilation supply and exhaust valves are leak rate tested each refueling between the outboard check and inboard butterfly valves. Testing the butterfly valves in the direction of flow can also be accomplished if warranted. Testing is conducted semiannually per Appendix J, section XI. Testing will require the valves to be exercised cuarterly. 1394 283

The ventilation probe valves are air tested each refueling by disecnnecting the differential pressure controller sensing line to the valves and applying pressure. Containment clean and dirty sump discharge lines and the emergency condenser drain check valve are tested between the valves during each refueling. The control valves will be exercised quarterly per section XI. Leak rate testing of the demineralized water isolation valves is done each refueling. The check valve is tested by removing a valve upstream, install-ing a test flange and applying air pressure to the check valve. The control valve is tested in the same manner, but the checs valve internals are removed. The check valve is retested after its internals are re-installed. The control valve is also to be exercised quarterly per section XI. The main steam isolation is air tested during the ILRT and hydrostatically tested during each system hydro. Valve exercising per section XI will be done during each cold shutdown. The main steam drain isolation valve i.1 tested the same as the MSI7. It, however, is closed and electrically disabled and will not be exercised quarterly. Reactor feedwater stop check and check valves are air tested between the valves each refueling. Operability is also checked during the ILRT and system hydrostatic test. Air testing of the control rod drive pump poppet valves and inlet check valve is done each refueling per Appendix J and section XI. The resin sluice valves are air tested between the valves each year. The treated waste check and control valve are air tested in the direction of flow during each refueling. The reactor and fuel pit drain valves are air tested between the valves each refueling. The shell side of the emer-gency condenser is open to the atmosphere. Lastly, the ILRT penetrations are also air tested for leak tightness eacn refueling to determine their leak rate. These lines are normally capped when not in use. I394 284

PLANT BIG ECCK POINT CCDES, STANDARDS AND GUIDES V. Identify the codes, standards and guides applied in the design of the containment isolation system and system components. Response: Containment isolati.on was not designed as a system at SRP. Isolation valves are designed based on their particular system require-ments and either check flow out of containment or are normally closed valves or have a safety system signal to close on high containment pressure and/or low reactor water level. The containment ventilation valves have input signals which cause them to close on all scrams and high radiation in the new fuel and rpent fuel pool storage areas. In addition, the safety piping syster.2. comprising the containment isolation system were designed and constructed using the American Standard Code for Pressure Piping, ASA B31.1, 1955 ed; the applicible American Society for Testing Materials, ASIM, specifications; and the American Society of Mechanical Engineers Boiler and Pressure Vessel code, ASME B&PV. I394 285

PLANT BIG ROCK N INT NOPRAL OPERATING MODES AND ISOLATION BARRIERS VI. Discuss the Tor =al operating modes and containment isolation provision and procedures for lines that transfer potentially radioactive fluids out of the containment. Response: Ventilation supply and exhsust control valves are normally open valves that close automatically in 6 seconds on all scrams. Both sets of valves in addition fail closed on air and power failure and can be closed by a hand switch on the control room console. They also re-ceive a signal to open if the containment is under a vacuum. The exhaust valves normally carry air out of containment where it is mixed with air from the turbine building and then enters the stack. The clean and dirty containment su=p lines are both normally open and their isolation valves close on high containment pressure (CHP), low reactor water level (LRWL) and loss of offsite power (LOP). The valves can also be operated via hand switches in the control room. Air or power failure to the solenoid valve will also cause the valces to close. The main steam isolation valve is the only automatic isolation valve in the main steam line and closes on CHP, LRWL and LOP. The valve is a DC motor-operated valve which is designed to close in 60 seconds. The clean-up resin sluice line is normally closed, be;cuer, cperation of the system via procedure is permissible when the reactor is in power opera-tion. The valves close automatically on CHP, LRWL and LOP. Additionally, the valves fail closed on loss of air and power to the solenoid valve. The valves also have a high pressure interlock to clcse if greater than 50 psi exists in the line. Reactor and fuel pit drain line valves are normally closed. Operaticn of the valves is controlled by two procedures. The valves close on the con-tainment isolation signals (CHP, LWRL and LOP) and fail closed on loss of air and power. b l0N

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