Text

Forwards Responses to short-term TMI Lessons Learned Task Force Requirements Re Proposed or Completed Method of Implementation
	ML19260C721
Person / Time
Site:	Calvert Cliffs
Issue date:	01/04/1980
From:	Lundvall A; BALTIMORE GAS & ELECTRIC CO.
To:	Eisenhut D; Office of Nuclear Reactor Regulation
References
	RTR-NUREG-0578, RTR-NUREG-578, TASK-2.B.1, TASK-TM NUDOCS 8001080507
	Download: ML19260C721 (81)
	v • d • e

{{#Wiki_filter:. BALTIMORE GAS AN D ELECTRIC COMPANY GAS AND ELECTRIC DUILDING BALTl MOR E, M ARYLAN D 21203 January h, 1980 ARTHUR C. LUN DVALL,.JR. Vict Pas siot=f

sum, Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C.

20555 Attn: Mr. Darrell G. Eisenhut, Acting Director Division of Operating Reactors

Subject:

Calvert Cliffs Nuclear Power Plant Units Nos. 1 & 2, Dockets Nos. 50-317 & 50-318 Follow-up Actions Resulting from TMI-2 Incident (Lessons Learned)

References:

(1) Lundvall to Eisenhut letter of 10/19/79, follow-up Actions Reculting from TMI-2 Incident. (2) Lundvall to Reid letter of 11/9/79, same subject. (3) Lundvall to Eisenhut letter of 11/20/79, came subject. (k) Lundvall to Reid letter of 11/23/79, same subject. (5) Lundvall to Eisenhut letter of 12/5/79, same subject. (6) Lundvall to Eisenhut letter of 12/lk/79, same subject. (7) Lundvall to Reid letter of 6/5/79, IE Bulletin 79-063. Gentlemen: Many of the staff positions on Short Term Lessons Learned required that information be provided by 1/1/80 on proposed or completed method of implementation. The enclosure to this letter contains our responses to these requirements for information. Very truly yours, f

ZE L

+ & c-rti L-Enclosure ec: J. A. Biddison, Esquire G. F. Trowbridge, Esquire Mr. E. L. Conner, Jr. 1703 332 o 8001080 gt q'

. to A. E. Lundvall Letter of 1/h/80 Information Pertainine to Imulementation of Short Term Lessons Learned Pecuirements Contents 2.1.1 Emergency Power Supply Requirements 2.1.2 Safety / Relief Valve Testing 2.1.3a Direct Valve Position Indication 2.1.3b Inadequate Core Cooling 2.1.h Diverse Containnent Isolation 2.1.6a Integrity of Systems Outside Containment Containing Radioactive Materials 2.1.6b Review of Plant Shielding and Environmental Quslification of Equipment 2.1.7a Automatic Initiation of Auxiliary Feed 2.1.7b Auxiliary Feed Flov Indication 2.1.8a Improved Post-Accident Samoling Capability 2.1.8b Main Vent Effluents 2.1.8c In-Plant Iodine 2.1 9 Transient and Accident Analysis 2.2.la Shift Supervisor

2. 2.1b Shift Technical Advisor 2.2.le Shift Turnover 2.2.2a Control Room Access 2.2.2b Technical Support Center 2.2.2c Operational Support Center Appendix A Conceptual Design for a Reactor Vessel Level Monitoring System Appendix B Design Review of Plant Shielding and Envir. Qual. of Equip.

Appendix C Conceptual Design for Post-LOCA Containment Air Sampling Appendix D Conceptual Design for Post-LOCA Liquid Sampling 1703 333

Page 2 2.1.1 Emereenev Power Sunnly Pecuirements The method of implementation of requirements for pressurizer heater power supplies is described in reference (2). Installation, procedure development, and operator training vill be complete before 1/31/80. The motive components of the power operated relief valves (PORV's) are supplied from safety related 480 V motor control centers which have a diesel backup. The control components of the PORV's are supplied from safety related 125 VDC battery buses. The motive and control components of the PORV block valves are supplied from safety related h80 V motor control centers which have a diesel backup. In each case the motive and control power for the block valve is supplied from a power supply train different from that which supplies the associated PORV. The PORV and PORV block valves motive and control power interfacec with the emergency h80 VAC and 125 VDC tases are breakers and fuses which are an integral part of these buses, and as such were purchased and are maintained in accordance with safety-grade requirements. Two of the pressurizer level instruments for each unit are powered from the vital de buses and the third is powered from offsite AC power with diesel backup. 2.1.2 Safety / Relief Valve Testing By letter dated December 17, 1979, Mr. Willimn J. Cahill, Jr., Chairman of the EPRI Safety and Analysis Task Force, submitted " Program Plan for the Performance Verification of PkTR Safety / Relief Valves and Systems", December 13, 1979 This program is responsive to the requirements of NUREG-0578. The EPRI Program Plan provides for completion of the essential portions of the test program by July, 1981. We intend to participate in the EPRI pro-gram to provide program review and to supply plant specific data as required. 2.1.3a Direct Valve Position Indication The direct position indication of the two PORVs and the safety valves per unit vill be accomplished with an acoustic monitoring system. The system vill be safety grade without redundant sensars and vill be alarmed with indication in the control room. The major equipment for both units is on site, with other equipment being expedited. The final engineering package from our consultants is being reviewed. The system has been seismically tested and qualified to meet our response spectra. The charge converter located inside the pressurizer housing has been qualified for our LOCA temperature profile and for 2.5 x 10 Rads. It is presently being qualified for an integrated dose of 100 Rads. It will be housed in a NEMA hX housing to provide further protection from the containment atmosphere. The accelerometer is qualified for continuous operation at 7500F, 1703 334

Page 3 0 and an integrated dose of 10 Rads. It is also hermetically sealed. The interconnecting cable between accelerometer and charge.unverter in qualified for our temperature profile and is radiation resistant. The connectors are temperature qualified, radiation resistant, and sealed from the containment atmosphere. 7he cable from the charge converter to the penetration is qualified safety related cable. Documented radiation qualification of the vendor-supplied cable and connectors vill be pursued with the vendor. 2.1.3b Inadequate core cooline Existine A detailed description of the procedures used with currently available instrumentation to detect and cope with inadequate core cooling may be found in item 2 of reference (7). Subcooled Marcin Monitor Existing reactor coolant system temuerature and pressure signals are narrow range. Rochester Instrument Systems can provide a single temperature trans-mitter with two continuous outputs covering different ranges each isolated from the other with an accuracy of 0.1% of span. A similar transmitter is available to provide two continuous analog outputs covering different ranges of pressurizer pressure. These devices are on order. The narrow range signals vill be used to feed existing instrumentation including the TM/LP trip calculat:* of the reactor protective system. The vide range signals vill be used to feed the subcooled margin monitors. These signals vill 0 cover 212-705 F and 15-3208 psia. See Table 1 (following page) more infor-mation. 1703 335

Page b INFORMATION REQUIRED ON THE SUBC00 LING METER i Display P Information Displayed (T-Tsat, Tsat, Press, etc.) T Marrin or Mnrein Display Type (Analog, Digital, CRT) Dicital Continuous or on Demand continunne Single or Redundant Display Redundant Location of Display ~'1005, 2005 Alarms (include setpoints) 500F Subcooled Overall uncertainty (*F, PSI) 3.1 F Range of Display 0-1000F Subcooled Qualifications (seismic, environmental, IEEE323) IEEE 3h4, 323 Calculator ~ Type (process computer, dedicated digital or analog calc.) Dediented nfeital If process computer is useo specify availability. (% of time) n/A Single or redundant calculators Redundant Selection Logic (hignest T., lowest press) Highest T & Iovest P Qualifications (seismic, environmental, IEEE323) IEEE 3hh, 323 Calculational Technique (Steam Tables, Functional Fit, ranges) Steam Tables Input Temperature -(RTD's or T/C's) RTD's Temperature (number of sensors and locations) 2 Cold Leg & 2 Hot Leg Temps / Meter 0 0 Range of temperature sensors 212 -705 F 1703 336 b

Page 5 +0.hoF in 515-6150F range Uncertainty

of temperature sensors (*F at 1 )

+0.80F # 212oF Elements: Seismic, LOCA Pressure Qualifications (seismic, environmental, IEEE323) Transmitters: seisnie. Control Rm Envrmt Pressure (specify ~nstrument used) F&P 50EP1041B-NS Pressure (number of sensors and locations) 2 Pressurizer Pressure / Meter Range of Pressure sensors 15-3208 usia Uncertainty

of pressure sensors (PSI at 1 )

+ 8 usi Qualifications (seismic, environmental, IEEE323) ~ Seismic, LOCA P&L ?'. Backup Capability All h channels of Tc, Th, Ppzr Availability of Temp & Press available to control rm operator Saturation curve visible to Availability of Steam Taoles etc. Control room onerator Trained to use availabel data to Training of operators determine level of subcooling Modilled to rerlect importance o' Procedures maintaining system in subcooled state

Uncertainties r.ust address conditions of forced flow and natural circulation
I 1703 337"'

Page 6 Additional Instrumentation The instrument response of a reference C-E plant for events which have the potential for inadequate core cooling is documented in CEN-ll7, " Inadequate Core Cooling -- A Response to NRC IE Bulletin 79-06C, Item 5, for Combustion Engineering Nuclear Steam Supply Systems". The conclusion reached in CEU-117 in that there current 1v is sufficient instrumentation in the nInnt to be used to detect inndeounte core cooline. Emergency procedures and operator training are being reviewed to ensure they adequately consider the information in CEN-ll7: any additional procedure revisions and training vill be completed by 1/31/80. The functional requirements for and a conceptual design description of a reactor vessel water level measurement system are provided as Appendix A to this enclosure. This design was prepared by the CE Owners Group for discussion with NRC in a generic resolution review. It is included here because the staff specifically required that reactor vessel vater level be evaluated, even though our analysis concluded that additional instrumentation is not necessary. 2.1.h Diverse Containment Isolation As described in reference (1), the containment isolation system is being modified so that the non-essential systems are automatically isolated on diverse parameters. The physical mod'.fication consists of reviring these containment isolation actuation circuits to actuation relays initiated by the safety injection actuation signal (SIAS). SIAS is initiated on either high containment pressure or low pressurizer pressure. The results of our re-evaluation of the definition of essential and non-essential systems, the identification of essential and non-essential systems, and a description of the basis for selection of each essential system is described in reference (3). This re-evaluation did not result in any changes to containment isolation designr. The design of the containment isolation valves control circuits is being modified to preclude the need for operator action to prevent inad7ertant opening of containment isolation valves after resetting the isolation signal. This is being accomplished by viring the valves' control switches so as to form a reset permissive, i.e., resetting vill only be accomplished if all the isolation valves' handsvitches are in the isolation position. This is being done for all the isolation valves, except the oxygen and reactor coolant system sample valves. Physical constraints require that this reset design objective be accomplished using additional lock-in relays that vill interrupt the power to the valves and fail them closed once an isolation signal is received; the circuit vill keep the valves closed until the operator repositions the valve control switch to the open position. 1703 338

Page 7 2.1.6a Intecrity of Systems Outside Containment Containine Radioactive Materials A. Systen Integrity Testinc and Lenkare Reduction A surveillance test program has been developed and implemented to determine leakages of systems that may carry highly contaminated fluids in the event of an accident. This surveillance has a refueling outage frequency. The surveillance has been completed on both units with the exception of some retesting after the repair of leakages. All shutdown cooling, high pressure safety injection, low pressure safety injection, containment spray, containment sump recirculation, containment atmosphere sampling and reactor coolant sampling systems were included in the surveillance. "As Found" leakages for these systems are shown on Table 2 (following page). Drainage from leaks are directed by way of gravity or sump pumped to liquid vaste receiver tanks. Dnergency core cooling exhaust systems or vaste processing exhaust systems ventilate areas where leakage could occur. Exhaust duct vork connects directly to each compartment thereby minimizing the spread of airborne radioactive contamination to adjacent compartments. Before leakage inspection was performed all boric acid residues were removed from piping and components. Systems which were insulated vere required to be at operating pressure for four hours before inspecting. Uninsulated systems were inspected after ten minutes. Valves, pump seals, flow orifices, flange connections, instrument tubing, vents, drains and piping vere all included in the inspections. All leakages were measured and a rate was recorded. Maintenance requests were generated to repair leakages and after repair leakage points are reinspected to verify reduction or elimination of the 1.eak. All leakage points over 5 drops per minute vill be repaired before or during the next cold shutdown. All other leakages vill be repaired before or during the next refueling outage. With regard to the NRC letter of 10/17/79 corcerning the North Anna Unit 1 release incident, our systems review, identification of modifi-cations, and schedule vill be complete by 1/31/80. 1703 si9

Table 2 Initial Leak Tent Procram Results Page 8 ENGINEERED SAFEGUARDS PUMP SUCTION PIPING INCLUDING SHUTDOWN COOLING RETURN PIPING Unit 1 No Leakage Found Unit 2 No Leakage Found CONTAINMENT SPRAY SYSTEM Unit 1

12 CTMT Spray Pump Seal 1 Qt.

/ min. Unit 2

21 CTMT Spray Pump Seal 30 Drops / min.

2-SI-315 Valve Packing 20 Drops / min. 2-SI-318 Valve Packing 3 Drops / min. HIGH PRESSURE SAFETY INJECTION SYSTEM Unit 1

11 HPSI Pump Casing Drain

< 1 Drop / min.

13 HPSI Pump Casing Drain 1 Drop / min.

1-SI-636-MOV Valve Packing 6 Drops / min. 1-SI-637-MOV Valve Packing 30 Drops / min. 1-SI-647-M0V Valve Packing 10 Drops / min. 1-SI-646-MOV Valve Packing 1 Drop / min. 1-SI-654-MOV Valve Packing 1 Drop / min. 1-SI-655-MOV Valve Packing < 1 Drop / min. Unit 2

23 HPSI Pump Casing Drain 1 Gal. / min.

2-SI-655-MOV Valve Packing 2 Drops / min. 2-SI-616-MOV Valve Packing 5 Drops / min. 2-SI-626-M0V Valve Packing 5 Drops / min. 2-SI-636-MOV Valve Packing 5 Drops / min. 2-SI-646-MOV Valve Packing 5 Drops / min. 2-SI-617-MOV Valve Packing 5 Drops / min. 2-SI-627-MOV Valve Packing 5 Drops / min. 2-SI-637-MOV Valve Packing 30 Drops / min. 2-SI-647-M0V Valve Packing 5 Drops / min. LOW PRESSURE SAFETY INJECTION SYSTEM Unit 1 1-SI-615-MOV Valve Packing < 1 Drop / min. 1-SI-625-MOV Valve Packing <; 1 Drop / min. 1-SI-635-MOV Valve Packing < 1 Drop 7 min. 1-SI-645-MOV Valve Packing < 1 Drop / min. Unit 2 No Leakage Found HYDROGEN ANALYZER / CONTAINMENT ATMOSPHERE SAMPLING Unit 1 Total System Leakage at 10 psig 5.7 secm. Unit 2 Totc) System Leakage at 10 psig 6.9 sccm. REACTOR COOLANT SAMPLING SYSTEM Unit i No Leakage Found Unit 2 No Leakage Found ( 1703g40

Page 9 2.1.6b Review of Plant Shielding and Environmental Qualification of Equipment Assuming a post-accident release of radioactivity equivaler.t to that described in Regulatory Guides 1.3 and 1.4, a radiation and shielding design review was performed of the spaces around systems that may, as a result of an accident, contain highly radioactive materials. The design review identifies the location of vital areas in which personnel occupancy may be unduly limited and the location of safety equipment which may be unduly degraded by the radiation fields during post-accident operation. The design review, included as Appendix B to this enclosure, determined the corrective actions needed for vital areas throughout the plant. 2.1.7a Automatic Initiation of Auxiliary Feed Refer to references (2) and (h) for method of implementation. In addition, we are ir. the process of preparing a response to R. W. Reid's letter of 12/21/79 2.1.7b Auxiliary Feed Flow Indication New cabling is being installed between the flow transmitters and square root extractors. Safety-related power is being provided to the instrument loop. Power supply, transmitters, and square root extractors are all safety-related. 2.1.8a Innroved Post-Accident Samuline Canability A design and operational review of the reactor coolant and containment atmosphere sampling systems was performed to determine the capability of personnel to promptly obtain a sample under accident conditions without incurring a radiation exposure to any individual in excess of 3 and 18 3/4 rems to the whole body or extremeties, respectively. Appendix C to this enclosure describes the conceptual design for post-accident condition air sampling and also includes the interim measures taken. Appendix D to this enclosure is the conceptual design for post-accident reactor coolant sampling. Appendix D also includes the interim measures which are being taken. 2.1.8b Main Vent Effluents In the event that the main vent monitors go off scale, grab samples of the main vent shall be obtained and analyzed. In the event that a sample is unable to be taken, a sample shall be taken at the site boundary and the release rate shall be determined using meteorological diffusion factors described in our Site Emergency Plan Implementation Procedures. Noble gas interference on the charcoal cartridge shall be reduced by purging the sample with non-radioactive air and by heating the cartridge at 80 C for 15 minutes. Details concerning this sampling have been described in RCP-1-503. 1703 34I

Page 10 2.1.8e In-Plant Iodine Whenever possible, iodine analysis shall be determined by utilizing a multichannel analyzer with a Ge (Li) detector. For this purpose, a multi-channel analyzer and its Ge(Li) detector shall be dedicated for the purpose of analyzing iodine sampics for respirator protection. In remote areas where it would not be feasible to have the sample analyzed at this location due to distance, a single channel analyzer and a Na(I) detector shall be utilized since it is predicted that the dose rates and noble gas interfer-ence vill be minimal, enabling this type of detector to be more useful. These samples will be re-analyzed using a multichannel analyzer and a Ge(Li) detector. 2.1.9 Transient and Accident Analysis Reactor Coolant System Vent It is our intention that vents be installed from each reactor vessel and pressurizer for the purpose of venting non-condensible gases from the reactor coolant system. It is further our intention to have these vents installed and operational prior to January 1, 1981. The design of these vents is described herein and an illustration is provided as Figure 1 (following page). The reactor vessel vent originates from an existing 3/4" vent line which also serves as part of the piping and tubing associated with the core differential pressure transmitter (1-PDT-12h for Unit No. 1). A 3/h" branch line vill be installed with two solenoid operated valves placed in ceries downstream of the connection. These solenoid operated valves shall be rated at 2500 PSIA, 800 F, and shall be nuclear class I and seismic class I, as vill the associated piping. These valves shall be normally closed and failed closed. Downstream of the valves shall be a test connection with two normally closed and locked valves. Further downstream, the 3/4" schedule 80 piping shall be reduced to 1/2" stainless steel tubing. This tubing vill join tubing from the pressurizer vent to form a common header which will connect into a 10" line immediately upstream of the quench tank. All tubing and piping downstream of the two solenoid operated valves shall be nuclear class II, and seismic class I. The pressurizer vent line vill originate from an existing 3/h" line used to obtain samples from the pressurizer vanor space. A 3/4" branch line shall be' installed with the valves and test connection as described for the reactor vessel vent. This line shall also reduce to 1/2" tubing and connect to the reactor vessel vent. The common header shall be equipped with a surface sensor for temperature indication. The existing quench tank vent valve vill be replaced with a three-way control valve enabling the operator to vent the quench tank to the vaste gas system or the containment as required. 1703 342

Page 11 The line size has been selected to ensure that the vent discharge from the reactor vessel, assuming saturated water at 2500 psia, would not exceed the make-up capacity of one charging pump. This results in no additional LOCA analysis being required for this installation. Calculations have also been performed to verify that the designed venting system has the capability to vent at least one half of the reactor coolant system (RCS) in one hour if hydrogen was assumed to occupy one-half of the RCS. The resultant energy loss from venting fluid from the reactor coolant system shall not exceed that pressurizer heater capacity which is povered from emergency buses. Handsvitches with indicator lights, as well as temperature indication vill be located at panels IC06 and 2006 in the control room. All handsvitches vill be installed with locks. Each valve shall be placed on a separate dedicated circuit breaker which could be removed during operation rendering the valve inoperable. Temperature indication vill serve to notify the operators of venting performance and to alert opcrators to possible valve leakage. There vill be an audible and visible temperature alarm to alert the operator to fluid in the vent line header. All solenoid operated valves shall be povered from emergency buses. Valves on the reactor vessel vent shall be povered from a different emergency bus than valves on the pressurizer vent. This vent system can, therefore, be used to vent either the pressurizer or reactor vessel to the quench tank, containment or vaste gas system remotely from the control room, with valve position indication and temper-ature indication to monitor venting effectiveness. Guidelines by the NSSS vendor concerning the plant characteristics to be nonitored to determine when to initiate and suspend venting vill be finalized prior to January 1, 1981. This constitutes our formal submittal of the reactor coolant vent system lesign. In order to maintain our schedule for installation, ve request that this design be reviewed and approved prior to March 1, 1980. 1703 343

ag cc-9 6 con) , w g c.:.-3 pips o c. ll X D Cp#.ll 0 > To cop.E otrr. 3A"ss 34'55 E/'. ' [h = . ' '4 TutN' TUBI V . I-Sy A 101 0b 0o F] V l-SM M 7 h \\b2 y V M y4 ] TEST O C ff Rc5 sr.MPt.c gY V ' Hooo j U ___] Test Y "ss TOBitG V .A l-RV l-Sy l-S V 242 to3 104 A %" Pl0E yg yg' (VA M 6 D NPE. IvttuG > WASTE. %% sysicM '$$c e' x x x x 0 ' ~ " V <w a >coNTMT tcMos. a4 cc-se /,S fl acww wx*u y a E VCV y c

y a

e ~ a ~ N > R.C.DRAtti TAMK o V u 2 REACTOR C00UdlT VENT SYSTEvi FIGURE 1

Page 13 / 2.2.la , Shift Sunervisor 1. A " management directive" has been issued by the Vice Precident-Supply that emphasizes the primary responsibility of the Shift Supervisor for safe operation of the plant. To insure the appropriate dissemination of this information, the directive has been issued as a supplement to Quality Assurance Procedure 25 (QAP-25), Plant Onerations. 2. The pertinent Quality Assurance Procedure, QAP-25, Plant Onerations, has been reviewed to assure that the duties, responsibilities, and authority of the Shift Supervisor and other NRC licensed operators are properly defined. QAP-25, as supplemented by the " management directive" discussed in item 1 above, adequately establishes a definite line of command and a clear delineation of command authority. 3. The incumbent Shift Supervisors have been trained regarding the " management directive" discussed in 1 above. h. In addition to past reviews of the administrative duties of the Shift Supervisort, a recent review of the entire structure of the Operations Unit was conducted by our Corporate Staff Services Department (at the direction of the Vice President-Supply). A portion of this recent review included an analysis of the adequacy of the chift and total complement of the operations uait to allow for the various administra-tive functions to be performed by persons other than the Shift Supervisor. The previous reviews had already increased the comnlement of the Unit to decrease the administrative burden on the Shift Supervisor. Specifically, complement changes were made to support the following functions. A. The establishment of a nernanent day-tino Tacring Authority Early in the commercial service of Calvert Cliffs, it became apparent that the burden presented by the administration of the safety tagging system detracted substantially from the Shift Supervisor's ability to monitor and administrate plant operations. Ey far the largest portion of the burden presented by tagging occurs on day-shift during the normal work veek, the hours during which the various shops perform their work. To alleviate that burden, a permanent position of Tagging Authority was initiated to work day-work on weekdays and administer the safety tagging system. The position is filled by a Control Room Operator; personnel are rotated on a monthly basis. The Tegging Authority interfaces with the shops, defines the scope of the tagging, makes out the tags and applies them. The Shift Supervisor reviews the proposed tagging and authorizes the application and removal of the tags. The Shift Supervisor must retain the authorization responsibility since the removal of equipment and systems from service may have an effect on the safe operation of the plant. 1703 345

Page lh B. Assignment of an Onerations Refuelinc Outare Coordinator It became apparent during the first two Unit 1 refueling outages that the increased activities during outage conditions created a commensurately increased workload for the Shift Supervisor, who also retained the responsibility for the safe operation of the in-service unit. Since that time, it has become the practice of the Operations Unit to rearrange the shift schedule to totally remove one of the Shift Supervisors from shift rotation and assign him the responsibility of outage coordination for the Operations Unit. This responsibility includes interfacing with the shops, outside repair organizations, and the Production Maintenance outage coordin-ator. The Operations outage coordinator also coordinates post-maintenance and modification testing and the pre-startup valve lineups, and operations surveillance testing for the outage unit. This organization has proven to be a practical means of relieving the Shift Supervisor of a distracting administrative burden during refueling outages; thus the practice vill be continued in the future. The most recent review of the organization of the Operations Unit resulted in another complement increase. One aspect of the complement increase was to allow the assignment of an experienced operator to directly assist the Shift Supervisor in several facets of routine plant operations. Commencing on December 31, 1979, an ernerienced operator was assigned on day shift, Monday through Friday, as the Shift Supervisor's assistant. It is expected that the Shift Supervisor vill use this person to relieve him of some of the routine t~.sks which occasionally may detract from the Shift Supervisor's primary respon-sibility for the safe operation of the units. Specifically, the assignment to day-shift allows the Shift Supervisor's assistant to provide coordination with the shops for system trouble-shooting, a function which is time-consuming and has, by default, become a burden to the Shift Supervisor. Other tasks to be performed by this assistant vill be the coordination of routine testing, as necessary3 and the coordination of non-routine operational evolutions, such as resin transfers. 2.2.lb Shift Technical Advisor The method of implementing this position is described in detail in references (1), (3) and (6). 2.2.lc Shift Turnover The plant procedure governing shift relief and turnover (QAP-25, Plant Operations) has been reviewed. In order to meet the requirements stated by the NUREG position several shift turnover checklists have been instituted 1703 346

Page 15 by means of the Huclear Plant Engineer-Operations Standing Instructions. The effectiveness of the turnover procedure vill be evaluated by the use of surveillance test procedures, independent checklists performed by the Shift Supervisor, Senior Control Room Operator and Control Room Operators, and by routine Quality Assurance audits. 2.2.2a Control Room Access An existing company procedure, QAP-25, Plant Onerations, establishes the authority of personnel in charge of the Control Room to limit access. The same procedure, as augmented by the " management directive" discussed under section 2.2.la above, establishes a clear line of authority for command in the Control Room. 2.2.2b Technical Sunnort Center The interim upgrading of the Technical Support Center is discussed in reference (3). The Control Room itself may be used as an alternate location. A new Site Emergency Plan Implementation Procedure has been issued to describe the activation and staffing of the Technical Support Center. The long range plan for the Technical Surnort Center is to install a data acquisition system in the same location as that described for the 1/1/80 implementation plan (reference (3)). This system vill meet our interpretation of the requirements of NUREG 0578 section 2.2.2b, the NRC clarification letter of October 30, 1979, the NRR Regional Meeting of September 24, 1979, and the NRC Topical Meeting of October 10, 1979 2.2.2c Onerational Support Center A portion of the plant (north) service building has been designated as the Operational Support Center; communications to the Control Room are available via the telephone and plant page system. A new Site Emergency Plan Implem-entation Procedure has been issued to describe the activiation and staffing of the Operational Support Center. 1703 347

Appendix A to Enclosure 1 C-E POST TliI EVALUATION TASK 2 COllCEPTUAL DESIGil FOR A REACTOR VESSEL LEVEL MONITORiliG SYSTEM DECE!4BER 21, 1979 1703 348

ABSTRACT In March 1979 an incident occurred at the Three Mile Island Unit 2 Nuclear Power Plant which resulted in considerable damage to the reactor core. As a result of investigations into the cause of this incident, areas where improvements in nuclear power plant administration, design and operation are necessary have been identified. With respect to nuclear power plant design, the Nuclear Regulatory Commission (NRC) has requested licensees to evaluate the use of a reactor vessel level measurement system as a means of providing additional useful information to power plant operators or as a means for directly determining the adequacy of core cooling. Subsequently, the C-E Owners' Group authorized Combustion Engineering to develop the functional requirements and a conceptual design for a system to monitor reactor vessel level. This report provides the result of the C-E investigation of reactor vessel level monitoring system designs. The report includes a description of the design basis and functional requirements for a Reactor Vessel Level Monitoring System (RVLMS); a description of the designs and design configuration recommended by C-E; and a brief discussion of other designs that were evaluated. The recommendec design will not require major plant structural changes and has a high probability of being installed by January 1, 1981. In developing the functional requirements and the conceptual design recommended in this report, it was assumed that the RVLMS is technically required to provide a valid indication of reactor vessel level or the adequacy of core cooling. This is an assumption upon which the design is based - analyses have not been performed to substantiate such an assumption. 1703 349 Page 1 of 43

TABLE OF C0flTEriTS Section Title Fage fio. ABSTRACT 1 TABLE OF CONTEflTS 2 LIST OF TABLES AtlD FIGURES 3 1.0 PURPOSE 4 2.0 SCOPE 4 3.0 REFEREfCES 4 4.0 BACKGROUf1D 5

5.0 DESCRIPTION

OF ASSUMPTI0flS, DESIGfl BASIS AND 7 FUNCTIO llAL REQUIREf1Ef1TS 5.1 ASSUMPTI0f1S 7 5.2 DESIGft BASIS 7 5.3 FUNCTIONAL DESIGil 8 5.4 FUNCTI0f!AL REQUIREMErlTS 9 6.0 RECOMMENDATI0f1 16 6.1 RECOMf1 ELIDED DESIGN 16 6.2 RECOMMENDED SYSTEM CONFIGURATION 18 7.0 OTHER DESIGNS CONSIDERED 20 1703 350 Page 2 of 43

LIST OF TABLES At!D FIGURES Table No. Title of Table Page No. I RVLMS DESIGN BASIS EVENTS 29 Figure No. Title of Figure Page No. 1 RVLMS FUNCTIONAL BLOCK DIAGRAM 30 2 HEATLD JUNCTION THERM 0 COUPLE INSTALLATION 31 3 HEATED JUNCTION THERMOCOUPLE INSTALLATION 32 4 SYSTEti CONFIGURATION FOR A RVLiiS 33 5 R-F PROBE I!!STALLATION 34 6 FLOATING SOURCE INSTALLATION 35 7 FLOATING SOURCE INSTALLATION 36 8 FIXED NEUTRON SOURCE AND DETECTOR 37 9 FIXED NEUTRON SOURCE AND DETECTOR 38 10 FLOATING DIPSTICK 39 11 FLOATING SPHERES 40 12 ULTRASONIC PROBE 41 13 BUOYANT FORCE INDICATION PROBE 42 14 EXTERNAL STANDPIPE WITH FLOAT SENSOR OR DP CELL 43 1703 351 Page 3 of 43

1.0 PURPOSE The purpose of this report is to document C-E's research and to recommend a Reactor Vessel Level Monitoring System Design. The design effort was requested by the C-E Owners Group in response to a fluclear Regulatory Commission Lessons Learned Task Force requirement. 2.0 SCOPE This report provides C-E's conceptual design for a Reactor Vessel Level Monitoring System. The system functional requirements, description of the designs considered, a design recommendation, and the recommended system configuration comprise the report. 3.0 REFEREf1CES 3.1 TMI-2 Lessons Learned Task Force Status Report And Short Term Recommendations 3.2 United States fluclea; Regulatory Commission, " Follow-up Actions Resulting From The IIRC Staff Reviews Regarding The Three Mile Island Unit 2 Accident", To All Operating fluclear Power Plants, Dated September 13, 1979. 1703 352 Page 4 of 43

4.0 BACKGROUND

As a result of reviewing the Tf11 incident, the NRC Lesson Learned Task Force issued a status report and short term recommendations document (Reference 3.1). This document consisted of recommendations to be implemented to further insure the public health and safety. One of these recommendations was associated with instrumentation for detection of inadequate core cooling in PWR's and BWR's. In its assessment of instrumentation adequacy, the task force concludes that sufficient instrumentation existed at TMI-2 to provide adequate information to indicate reduced reactor vessel coolant level, core voiding and deteriorated core thermal conditions. However, in order to preclude failure to recognize such conditions in the future, the task force proposed to address this problem in two stages. The first stage further evaluated the use of installed instrumentation. The second stage was to study and develop system modifications that would not require major structural changes to the plant and that could be implemented in a relatively rapid manner to provide more direct indication of certain plant parameters than that available with present instrumentation. This second stage includes the study of PWR vessel level detectors. In stating its position with respect to this recommendation, the task force required licensees to provide a descrip-tion of any additional instrumentation or controls (primary or backup) proposed for the plant to supplement the currently installed instrumentation including a description of the functional design requirements. In reference 3.2 the NRC imposed the Lessons Learned Task Force's recommendations upon all Operating Nuclear Power Plants and provided a final schedule for implementation. The Owners Group subsequently authorized C-E to develop the functional requirements and conceptual design for a system to monitor reactor vessel level. This repcrt documents the results of that study. In developing the functional requirements and the conceptual design Page 5 of 43

recommended in this report, it was assumed that the RVLMS is techni-cally required to provide a valid indication of reactor vessel level or the adequacy of core cooling. This is an assumption upon which the design is based - analyses have not been performed to substantiate such an assumption. In addition to clearly understanding what the results of this task are, it is important to understand what is not provided as part of this task. 1. The described system does not monitor reactor vessel level below the bottom of the hot leg, i.e. into the active core region. Various options appear to be open in this area. It is believed that already installed instrumentation--the incore thermocouples and self powered neutron detectors--can be used to provide an indication of level in the core. Further analyses are required to confirm this. Alternatively, the heated junction thermocouples described in this report could be adapted to extend down into the core region. 2. No analyses or studies have been undertaken as part of this task to correlate the output of the described RVLMS to an " unambiguous indication of inadequate core cooling". The system monitors and displays reactor vessel level above the core, only. Further analyses would be required to show how the indication of reactor vessel level relates to core cooling adequacy. 3. No procedures or guidelines have been developed as part of this task to show how the RVLMS indication should be used. 4. No analyses have been done as part of this task to determine if a system to directly measure reactor vessel water level is required. 1703 354 Page 6 of 43

5.0 DESCRIPTI0ft 0F ASSUMPTI0flS, DESIGft BASIS AffD FUNCTI0ftAL RE0VIREl1Ef1TS 5.1 ASSUMPTI0fts The following assumptions were made in the conceptual design effort of the RVLMS. 1. The reactor vessel inventory between the bottom of the hot leg to the top of the reactor vessel head is to be monitored. This range was specified at the beginning of the task and represents that area of the reacter vessel where no instrumentation exists which can be used to measure vessel inventory. In addition, the response of existing in-core instrumentation may be useful in measuring core inventory. 2. The RVLMS will be useful to the plant operator, therefore, it must be reliable throughout the operating cycle. 5.2 DESIGli BASIS A Reactor Vessel Level Monitoring System (RVLMS) will supplement the existing f1SSS instrumentation that gives an indication of the adequacy of core cooling. The llSSS design addresses the need to supply water to the reactor coolant system during an unlikely event that affects reactor coolant system inventory. However, under these conditions, the RVLMS will inferm the plant operator via its alarm that the reactor vessel level inventory is affected, specifically, and additional effort to assure the adequacy of core cooling may be necessary. The continuous recording of reactor vessel level will provide both a permanent record of this parameter and a means of evaluating the results of evolutions which may affect reactor vessel level by indicating trends. In addi-tion, the recording of reactor vessel inventory, in conjunction with other plant instrumentation, will provide the plant operator with a means of determining core cooling adequacy. The reactor vessel level alarm setpoint will be sucn as to alert the plant operator in time to allow for evaluation of overall flSSS response to the cause for the alarm initiation. The alarm setpoint will also be such as to preclude spurious alarms during plant operation. Page 7 of 43

In the event the system should suffer a loss of electrical power or a major failure, a unique indication shall be provided to inform the plant operator that the system is not monitoring reactor vessel level and it should not be used to make decisions which cculd affect core cooling adequacy. The RVLliS Design Basis is as follows: 1. Sense reactor vessel level and continuously record the parameter in the control room. 2. Provide an alarm and annunciation in the control room when reactor vessel water level decreases to a preset value to aiert the plant operator prior to further degradation of reactor vessel inventory. 3. Monitor and record reactor vessel water level during the following plant operating modes: a. Power, operation b. Start-up c. Hot standby d. Hot shutdown e. Cold shutdown 4. Monitor and record reactor vessel water level during post accident conditions where reactor vessel inventory may be lost. The Design Basis Events under which the system must meet its design require-ments are listed in Table I. 5.3 FUNCTIONAL DESIGN 5.3.1 General Figure 1 provides a functional block diagram of the Reactor Vessel Level Monitoring System (RVLMS) and its interfaces. The RVLMS monitors changes in the reactor vessel bulk water level. This input is utilized to perform two basic functions by the RVLMS; Display and Alarm. g()3 356 rav 8 of 43

The first function is to display the reactor vessel water level to the plant operator by employing a strip chart recorder. The second function is to provide an alarm to the Annunciation System when reactor vessel water level reaches a low preset level. 5.3.2 Display Function The reactor vessel water level is the input to the RVLMS. The RVLMS continuously displays the reactor vessel water level on a strip chart recorder in the control room. 5.3.3 Alarm Function The RVLMS provides its output signal to the plant Annunciator System. The plant Annunciator System will initiate an alarm when the sensed reactor vessel water level reaches a low preset value. 5.4 FUNCTION REQUIREMENTS This section specifies the requirements to be met in the design of the RVLMS. 5.4.1 Design Criteria The following Regulatory Guides, IEEE Standards and 10 CFR 50 require-ments will be met in the design of the RVLMS. 5.4.1.1 Regulatory Guides 1. 1.97 Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident - Revision 1, August 1977. 2. 1,29 Seismic Design Classification - Revision 3, September 1978. 3. 1.100 Seismic Qualification of Electric Equipment for Nuclear Power Plants - Revision 1, August 1977. 1703 357 p,g, g,, 43

4. 1.89 Qualification of Class IE Equipment for Nuclear Power Plants - Original, November 1974. 5. 1.75 Physical Independence of Electric Systems - Revision 2, September 1978. 6. 1.53 Application of the Single-Failure Criterion to Nuclear Power Plant Protection Systems - Original, June 1973. 7. 1.118 Periodic Testing of Electric Power and Protection Systems - Revision 2, June 1978. 8. 1.70 Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants - Revision 3, November 1978.

9.

Draft Revision 2 to 1.97 Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant Environs Conditions During and Following an Accident. This includes the draft standard ANS-4.5 Functional. Requirements for Post Accident ;1onitoring Capability for the Control Rcom Operator of a Nuclear Power Generating Station. 5.4.1.2 IEEE Standards 1. 323 Qualifying Class 1E Equipment for Nuclear Power Generating Stations. 2. 338 Criteria for the Periodic Testing of Nuclear Power Generating Station Safety Systems. s 3. 344 Seismic Qualification of Class 1E Equipment for Nuclear Power Generating Stations. 4. 384 Standard Criteria for Independence of Class 1E Equipment and Circuits. 5. 379 Standard Application of the Single-Failure Criterion to Nuclear Power Generating Station Class 1E Systems.

These items are included to represent those requirements which should t'e considered in the design but are not presently required.

1703 358 eage 10 of 43

5,4.1.3 10 CFR 50 Appendix A 1. CRITERI0t1 1 - Quality Standards and Records 2. CRITERI0ft 2 - Design Bases for Protection Against Natural Phenomenon 3. CRITERI0ft 3 - Fire Protection 4. CRITERI0tl 13 - Instrumentation and Control 5.4.1.4 10 CFR 50 Appendix B 5.4.2 Interface Reauirements The following interface requirements shall be accommodated: 5.4.2.1 Input Requirements Electrical Power Source The RVLMS power sources shall be from two Class 1E power buses. One power source shall supply electrical power to one of the RVLMS channels. The other, separate, power source shall supply electrical power to the other, separate, RVLMS channel. 5.4.2.2 Output Requirements Annunciation and Alarm The RVLMS shall provide output signals to the Plant Annunciator system when the sensed reactor vessel water level reaches a low preset value. A separate output signal shall be provided for each separate RVLMS channel. These signals shall alarm separate indications on the Plant Annunciator system. The Plant Annunciator shall provide an audible and visible indication to the plant operators of the RVLMS low reactor vessel water lovel condition. Page 11 of 43

~ 5.4.3 Sensor Requirements Reactor Vessel Level The RVLMS shall provide signals via reactor vessel level transmitters which represent the reactor vessel bulk water level. The range of the signals shall be at least from the top of the reactor vessel head to the bottom of the reactor vessel outlet nozzle. Sensor Requirements Locations The RVLMS sensors may be located interrial to or external to the reactor vessel. Sensor Requirements Range The RVLMS sensors shall monitor the range from the bottom of the reactor vessel outlet nozzle to the top of the reactor vessel head, as a minimum. Number of Channels The RVLMS shall have at least two redundant channels for sensing reactor vessel level. 5.4.4 System Accuracy and Response Time Requirements System Accuracy The RVLMS system accuracy shall be better than 10% of the required range discussed above. This accuracy applies to both the alarm set-point and to the indication to the plant operator as read on the RVLMS recorders. 1703 360 Page 12 of 43

8 TEST TARGET (MT-3) ' DE D3A d 1.0 y @ I!!!al I.I ['HM la 1.25 1.4 1.6 9 %? O /% 4

If $/I

'S4A,,4 y,,, a

A>>. /% s f '4 im e ev <e 1,. TEST TARG.T (MT-3) 1.0 l# 22 BE y y ll!!E Il j,mHe in 1.25 1.4 1.6 6"

4

++4\\ O f4{g[fk

%8; 43 I

.~_.~~~~'5

4 4 M+!)* k.k TEST TARGET (MT-3) 1.0 '" EE HM f @ EE u j," M l.8 1.25 1.4 1.6 6-

4
+}g 4%

%fh/L thN 5,,

I ,_..~.-U -d e

4

+%s%> $<#4 s><

'+ me... <e 1,. TEST TARGET (MT-3) 1.0 l!EM BM -y5 EE I.I [m EM l 1.8 i 1.25 1.4 1.6 = 6" b e 9 h# Oh

ff>$ b hth%g$

s _1 a

System Response Time The RVLMS system response time shall be at most 6.0 seconds from the time a change in reactor vessel level is sensed until the change is indicated at the recorder. It shall also be at mosc 6.0 seconds from the time the recctor vessel water level reaches the low preset setpoint until the alarm signal is present at the output of the RVLMS. 5.4.5 Display Requirements Recording The RVLMS shall continuously monitor and simultaneously record the signal from each reactor vessel level sensor. Separate recorders shall be provided for each separate RVLMS channel. The recorders shall be a permanently mounted component located in the control room and uniquely identified as post accident monitoring equipment. 5.4.6 Alarm and Annunciation Requirements The RVLMS shall provide alarm signals to the Plant Annunciator System upon determining that reactor vessel level ha. reached a low preset value. A separate alarm signal shall be provided for each RVLMS channel. This setpoint will be determined later during the setpoint effort. 1704 001 Page 13 of 43

5.4.7 Testing Requirements Instrument Channel Checks The RVLMS channel outputs shall be displayed such that channel operability may be verified by comparing readings between channels. Functior.a1 Tests The RVLMS channels shall have the capability to be tested by injecting a test signal as close to the sensor as is practical to assure that each channel is capable of performing its design function. This shall include having the capability of testing the accuracy of the setpoint used for the alarm function as well as testing the accuracy of the display available to the operator. The ability to test the system response time shall also be provided. Sensor Calibration Verification Tests The RVLMS sensors shall have the capability to be tested such that with a known precision input the instrument gives an output signal of the required accuracy and response time. 5.4.8 Operational Requirements The RVLMS is required to function within the required accuracy and time response limits during the modes of plant operation specified in the Design Basis section. 1704 002 Page 14 of 43

5.4.9 Post Accident Monitoring Requirements The RVLMS is required to perform its required functions with the required accuracy and time response limits during post accident monitoring conditions where reactor vessel level may be affected. 5.4.10 Qualification Requirements The RVLMS shall be qualified to perform its required functions with the required accuracy and time response before, during and after the most adverse environmental conditions associated with the Design Basis Events listed in Table I. With respect to seismic qualification, the RVLMS shall continue to perform its required fur.ctions with required accuracy and time response following the seismic event. 5.4.11 Indication of Loss of Power or System Failure Reouirements An indication of loss of electrical power or a system failure to each RVLMS channel shall be provider!. 5.4.12 Accessibility & Mountina Reauirements The RVLMS recorders shall be permanently mounted in the control room and the recorders shall be clearly displayed for operator use. 1704 003 Page 15 of 43

~ 6.0 RECOMMENDATION This section of the report provides Combustion Engineering's recommenda-tion of the best means by which to measure reactor vessel level and the reason for the recommendation. 6.1 RECOMMENDED DESIGN 6.1.1 Heated Junction Thermocouple The heated junction thermocouple sensing means measures the change in thermocouple output voltage as a result of the difference in the thenral conductive properties between steam and water. A series of heated junction thermocouples will be located at different axial positions above the core. Figures 2 and 3 depict heated junction thermocouple installations in the reactor vessel. Each discrete location would have two thermocouples connected in series but in electrical opposition to each other. One of these thermocouples will be heated, thereby establishing a reference differential voltage output which is a function of the heater input. When the water sr.rrounding the heated junction thermocouple is replaced by steam or air the voltage generated by the heated thermocouple will change because the heat generated will no innger be removed from the area by the water. Thus the temperature of the heated thermocouple will increase relative to the unheated thermocouple. This change will be used to determine the steam water interface in the reactor vessel. 6.1.1.1 Evaluation of the Heated Junction Thermocouple The heated junction thermocouples are adaptable to the operating reactors by insertion through an In Core Instrument (ICI) nozzle port or a Part length CEA location. This method of determining vessel level is presently in use in a research facility in Idaho. The resolutina of the system is adjustable over the range of 1704 004 measurement based upon the distance between thermocouples. The accuracy of the system is expected to exceed the + 10% of span Page 16 of 43

presently required. The heated junction thermocouples may be compatible with existing core designs for measuring level in the core. However, this application has not been thoroughly evaluated. Since the application of heated junction thermocouples for monitoring vessel level has been demonstrated, there is a high probability of system installation by the January 1981 deadline imposed by the Nuclear Regulatory Commission. Three disadvantages associated with the use of heated junction thermo-couples are as follows: 1. Use of the ICI nozzle port for thennoccuple insertion will require removal of in-core instrumentation. With respect to the Technical Specifications, the plant will then be operating witn a certain percentage of unavailable in-core-instruments based upon the in-core-instrument uncertainty analysis. 2. Use of a Part length CEA location will require removal of a Part Length CEA in order to accommodate the heated junction thermocouples. 3. The resolution of the system is restricted by the number of electrical leads which can physically be accommodated by the ICI nozzle port and/or spare containment electrical penetrations. 1704 005 Page 17 of 43

6.2 RECOMMENDED SYSTEM CONFIGURATION 6.2.1 Description

he Recommended System Figure 4 shows the recommended system configuration for the Reactor Vessel Level Monitoring System as discussed below.

The recommended system configuration will require that four Part Length CEh's be removed to provide locations for the sensing devices. Two Heated Junction Thermocouple level sensor groups will comprise one channel of vessel levei indication. Each g*oup of Heated Junction Thermocouples will be inserted into a vacated Part-Length CEA location. The output signal will be sent to a recorder located in the control room where each output signal can be continuously recorded. A selector switch may be provided to allow the average of the two signals or any one or both of the two level sensor outputs to be recorded at a time. The remaining two Part-Length CEA locations will also contain a Heated Junction Thermocouple level sensor in each location. This second channel will be identical to the first channel. The second Heated Junction Thermocouple channel will record its sensor output on a separate recorder. 6.2.2 Reasons for the Recommended System The advantage of the specific type of sensor recommended has previcusly been discussed. Following is a discussion of the reasons for the recommended system configuration. It must be emphasized that in developino the functional requirements and conceptual design recommended in this eport, it was assumed that the RVLMS is technically required to provide a valid indication of reactor vessel level or the adequacy of core cooling. TM assumption was required to be made in order that the task could proceed. No analyses Page 18 of 43

have been performed as part of this task to substantiate such an assumption. Making the assumption that the RVLMS is technically required, it is believed that the recommended design is the optimum method of per-forming its required functions. A system of lesser complexity (i.e. no redundancy) may be acceptable from a licensing standpoint. However, if a RVLMS is not required for a technical reason, it is recommended that no system be installed. During the course of developing a conceptual design for a Reactor Vessel Level Monitoring System, certain basic criteria emerged as being necessary or very desirable for such a system. Meeting these criteria will go a long ways toward ensuring that the system will not only indicate level to the required degree of accuracy but will also have sufficient reliability and clarity so as to be genuinely useful to a plant operator. The criteria and how the recommended system configuration meets the criteria is discussed below. Redundancy Redundancy is a desirable design feature to incorporate because it is a means by which the availability of the system can be assured throughout the operating cycles. If one has a system of two separate channels, one has redundancy. If one sensor fails, the redunoant sensor is still operable. However, it is very desirable to have more than two degrees of redundancy in any system that is to be relied upon. If there are two channels ir,dicating different levels, which one is correct? It is thus desirable to have at least three degrees of redundancy to alleviate this problem. If there are four degrees of redundancy, the system can continue operating reliably with one redundant indication system failed. This still leaves the capability to identify a failure that may develop as a result of the event the system is needed for. The recommended system has four degrees of redundancy. There are two redundant channels, each of which has two separate sensors. This is Page 19 of 43 qg g

d the optimum system from a redundancy sta dooint and is easily imple-mented because the designs require removal of part length CEAs and they are arranged in symmetric groups of four. Non-Hydraul ic During the events that the RVLMS is required to operate, the hydraulic forces in the area of the reactor vessel head can be extremely dynamic and unpredictable. Although this concern may not be insurmountable, much analytical and design work would be required to ensure proper indication of level using a sensor system that worked on a hydraulic principle. The recommended design does not utilize this principle to determine level. No Movina Parts A level system that relies on physically moving components should be avoided if possible. Historically, moving components tend to fail by virtue of their movement. The recommended design uses no moving parts to sense vessel level. 7.0 OTHER DESIGNS CONSIDERED This section provides a brief description of other designs considered for use in measuring Reactor Vessel Level along with the advantages and disadvantages associated with each design. 7.1 Radio Frequency (RF) Probe The R-F probe design (Figure 5) consists of a Time Domain Reflectometry unit which sends a voltage step along the length of the R_-F Probe. The pulse is reflected back to the Time Domain Reflectometry unit when it encounters a discontinuity along the probe such as water. The time between pulse initiation and reflection is measured by the circuitry and translated into a level signal. 1704 008 Pagc 20 of 43

7.1.1 Advantages 1. It is highly feasible since the design was used successfully in the C-E pot boiler blowdown test conducted in 1976. 2. Test Report, TR-ESE-164, "R-F Probe for Two Phase Liquid Level Measurements Description and Calibration", dated September 29, 1977, indicated this system is suitable for measurement of reactor vessel l evel. Accuracy 1.5"; Response Time 60 msecs. 3. It is adaptable to operating operators. 4. There are no moving parts. 5. Corrcctions to problems are documented in the test report. Therefore, development time is reduced making the January 1981 installation date realistic. 7.1.2 Disadvantages 1. Reauires removal of part length CEA. 2. May impact existing refueling procedures (probe must be withdrawn). 3. Probe and accessories must be manufactured. 4. The effect of radiation, coolar.t temperature changes and coolant chemical concentration changes on system stability and reliability must be further evaluated. 1704 009 Page 21 of 43

7.2 FLOATING SOURCE The floating source concept is shown in two variations on Figure 6 and 7. It consists of a low strength Eerylium-Plutonium neutron source double encapsulated within a thin wall stainless steel sphere. The sphere is pressurized with helium to minimize the shell thickness and for detecting leaks. In one variation (Figure 6) the sphere is 7" in diameter and is contained within a CEA shroud. The only addition required to the upper guide structure (UGS) is the removal of CEA guide brackets and the addition of a " top hat" which is mechanically joined to the top of the CEA shroud. This arrangement of course means that a part length CEA must be removed. The second variation (Figure 7) utilizes a 3" dia sphere contained within a perforated sleeve. The sleeve is installed within the CEA shroud and mechanically locked in place with no machining or cutting required. During normal operc ion, the sphere is pinned against the top of the sleeve by the buoyant force of the coolant. When the water level drops, the sphere follows the water level. A neutron detector mounted on the R.V. closure head directly above the source sleeve measures the position of the sphere as a function of neutron reception. 7.2.1 Advantaces 1. It is feasible as determined by review of existing technology. 2. It is expected to operate under normal and accident conditions. 3. It is adaptable to operating reactors by modifying vessel components. 4. It has minimum impact on existing refueling procedures. 5. Its accuracy is expected to meet the functional requirements. 1704 010 Page 22 of 43

7.2.2 Disadvantages 1. Requires removal of part length CEA and modification to CEA guide tube. 2. Design must ensure that ball and/or source remain in CEA guide tube. 3. Floating ball movement against the CEA guide tube may wear through ball surface. 4. Problems associated with design development may prevent system installation by January 1981. 5. Source, ball and CEA guide tube modifications must be manufactured. 6. Vessel Head Removal is required to correct any mechanical problems. 7.3 FIXED f1EUTRON SOURCE AND DETECTOR The fixed neutron source and detector concepts are shown on Figures 8 and 9 and utilize the ICI routing conduit. The two schemes under consideration include a fixed separate source conduit and two assemblies of small neutron detectors axially located within an adjacent conduit (Figure C), and an alternate scheme where the source and detectors are located in the same conduit (Figure 9). The detectors measure the neutron flow from adjacent sources and, therefore, accurately predict water level and steam quality above the water layer. 7.3.1 Advantages 1. It is feasible as determined by review of existing technology. 2. It is expected to operate under normal and accident conditions. 3. It is adaptable to operating reacters by inserting detector string through ICI port. 4. Its accuracy is expected to exceed the functional requirements. 5. There are no moving components. 6. Detectors exist. 170401k m Page 23 of 43

~ 7.3.2 Disadvantages 1. Requires use of ICI Nozzle Port. If there are no spare ports, ICI removal is necessary. 2. There is a history of failure associated with the detector. 3. Problems associated with design development may prevent system installation by January 1981. 4 Source must be manufactured. 7.4 FLOATING DIP STICK - FLOATING ShCRES The floating dip stick concept is shown in Figure 10 and consists of large diameter thin walled stainless steel spheres connected to a shaft. The top of the shaft contains a magnet which can be located by a reed switch assembly in the same manner as CEA position is determined. The large diameter spheres serve as a floatation device for the stick. This concept requires that the density of tha coolant be constantly monitored by means of temperature recording equipment. An alternate scheme to the floating dip stick concept is shown in Figure 11 and consists of a long column of 1-1/2" diameter thin walled spheres. The column of spheres is supported by the buoyant force acting on about the 15 lower spheres. The entire column is enclosed within a perforated sleeve. The upper sphere contains a magnet similar to that used in the floating dip stick scheme. This scheme is much simpler than the floating dip stick when it comes to special refueling procedures because the upper containment sleeve and spheres are handled along with the closure heed, i.e., the column of spheres are held in place by battery operated electromagnetic means. 7.4.1 Advantages 1. It is feasible as determined by review of existing technology. 2. It is expected to operate under normal and accident conditions. 3. It is adaptable to operating reactors by removal of a part length CEA. 4. Its accuracy is expected to meet the functional requirements. 5. The existing reed switch position transmitter may be used. 1704 Oli nye 24 of e

7.4.2 Disadvantages 1. Requires removal of part length CEA and modification to CEA guide tube. 2. Accuracy is dependent upon water density. 3. More than one floating ball is required. 4. Movement of the floating ball against the CEA guide tube may wear through the surface of the ball. 5. Floating balls, dipstick and CEA guide tube trodfications must be manufactured. 6. Vessel Head Removal is required to correct any mechanical problems. 7. Problems associated with design development may prevent system installation by Janue y 1981. 8. It may impact existing refueling procedures (dipstick must be removed). 7.5 ULTRASONIC PROBE The ultrasonic probe design (Figure 12) consists of a. pulse transmitter which sends a pulse down a stepped shaft wave guide which is located within a Part Length CEA guide tube. The wave guide is reduced in circumference (stepped) at regular intervals to provide a distinc-interface upon which the signal will act. The pulse will reflect back when it encounters the discontinuity in the wave guide. The amplitude of the back reflection will be larger if a step is covered with steam than with water. The difference in the amplitude of reflected signals is analyzed by the circuitry for water level determination. 7.5.1 Advantages 1. It is feasible as determined by review of existing technology. 2. It is expected to operate under normal and accident conditions. 3. It is adaptable to operating reactor by removal of part length CEA. 4. There are no moving parts. 5. Head removal not necessary for repair. 1704 012 Page 25 of 43

7.5.2 Disadvantages 1. Requires removal of part length CEA and modification to CEA guide tube. 2. Problems associated with design development may prevent system installation by January 1981. 3. May impact existing refueling procedures (wave guide must be withdrawn). 4. Wave guide and CEA guide tube modification must be manufactured. 5. Steam bubbles may collect on shaft steps thus causing false level indication. 7.6 BUOYAt1T FORCE TRAf1SMITTER The buoyant force transmitter is shown in Figure 13 and consists of a 15 feet long - 2" diameter light weight hermetically sealed stainless steel tube. The shaft is connected to a linear voltage differential transmitter (LVDT) cell which monitors change in weight of the tube as a function of the buoyant force. The shaft is positioned within a CEA shroud. 7.6.1 Advantages 1. It is feasible as determined by review of existing technology. 2. It is expected to operate under normal and accident conditions. 3. It is adaptable to operating reactors by removing part length CEA. 4. Its accuracy is expected to meet the functional requirements. 7.6.2 Disadvantages 1. Requires removal of part length CEA. 2. Output must be compensated for fluid density changes to meet accuracy functional requirements. 3. Problems associated with design development may prevent system installation by January 1981. 4. Prob id accessories must be manufactured. 5. Vessel removal is required to correct any mechanical problems. 1704 013 eage 2e of 4,

6. Receiving an accurate indication of the buoyant force on the LVCT cell , complicated by the system pressure, the dynamic forces on the tube and the frictional resistances. 7.7 EXTERNAL STAT 4DPIPE WITH FLOAT SENSOR OR DP CELL The external standpipe approach (Figure 14) would make use of existing connections on the primary system. The standpipe would be connected between the head vent and a flange connection on the bottom of tne hotleg piping. The standpipe, which is filled with reactor coolant, would contain a float mechanism which contains a source or a magnet. A suitable detector would be located alongside the standpipe to detect the position of the float. The position of the float would reflect the reactor vessel water level. A DP cell (Figure 14) may be added to or substituted for the standpipe. 7.7.1 Advantages 1. It is feasible as determined by existing technology. 2. It is expected to operate under normal and accident conditions. 3. It is adaptable to operating reactors by connecting to the head vent and flange on the hotleg. 4. Installation will not require modification of. actor internals or removal of the vessel head. 5. DP cells are used to measure level and inaccuracies as.cociated with level measurement are understood. 6. Accuracy associated with float in the standpipe is expected to meet the functional requirements.

i. System is accessible.

1704 014 Page 27 of 43

7.7.2 Disadvantages 1. External standpipe mus'. be re- "ed and stored during refueling. 2. DP cell must account for f'... .i the reactor vessel, changes in reacto" coolant temperature,. .fifference between reference leg temperature and reactor cool:.r.c temperature. 3. DP cell reference leg flashing effects on accuracy must be considered. 4. Stanc* pipe layout problems are plant specific. 5. Increases primary sys;em piping which can experience leakage. 6. Leakage at flanged conr..tions will affect system accuracy. 7. Problems associated with standpipe design and installation may prevent system installation by January 1981. 1704 015 Page 28 of 43

TABLE I RVLMS DESIGN BASIS EVENTS The RVLMS shall be designed to function during the following Design Basis Events (including the environmental effects associated with these events). I. Loss of Coolant Accident (LOCA) up to and including a double-ended rupture of the largest pipe of the Reactor Coolant Pressure Boundary (RLpB) as defined in 10 CFR 50.2. II. Steam Systen pipe rupture up to and including a double-ended rupture of the largest pipe in the system. III. Feedwater System (main, auxiliary or emergency, and steam generator blowdown system) pipe rupture up to and including a double-ended rupture of the largest pipe in the system. IV. Depressurization due to inadvertent actuation of a pressurizer or secondary safety valve. 1704 016 Page 29 of 42

RV LM S FUNCTION AL BLOCK DIMH AM l e ~ I l t l l I O I N l $$'rc. I l ANNUNCIATOR SY3 TIM T 8 y scuso n, LA3. A R M p l -] U u l T OtSPLAY T l l -q

fWER, g(RFAR bER(Onygg(ggcM) g Su PPCf i

1 u-PDwCR l I Su pr:.Y O y I I u 4(REcoRDERcournot RonM) u I 7 _ g,,, p P .LEyFL Scus , ALARM ___ _j R T N I T CD DISPLAY g ANNhNCIATOR SY3TEf1 4 l y i e I i l y-l l

y FI6 Lt RE 1 a

CD W, O O e W

yo D

D N]Q' IM a. S.

inl a

* " - - ~

HEATFD 70NCT'lON TilEPsMQC_QOPLES IN EY.l5 TING C.r GUttE PATRG p l@RE 2. ~~ 7 t _J o 4 e. ..___._-....6,... ~; OUTER SHEATH ' l- , (SottD en FLEydEL@, i .. :..= i g 3 W 4 Ehe... y .,g, 1-f n p /s i 4 MTECTOF?lg:)%' g L_\\ o. -;= as scer.

-..:... t p

uEnrw. p LTHEP.MO 3 O O' O O x couptas .s g.tg4 G .j N \\ r3 N \\ ww /g, / 1CI PL ATE. /.9 r ', l l [ ~ ,/ g~~ t. n j = 7 J h ) p nor i.ca J WlGTING <,=cz nes ~ y / + ~. f 1704 018 . m, s s s s..xx s. s / Page 31 of 43 t

HE ATED TURCTION THERMo covPLE. PRcEC-IN CE A SrtKCOD Fis uac 3. PR.08E. LE ADS d J . o

GRAYLOCK CONNIECT'OR.

(g - y\\ F 6 /ORPTEt'. FOR GRANtoCK s 1 - VACAMT PLCE DM f McTDR. HoustNC, ~ p...__.. ' 00TER SHEATH (50CtD ot FLExteLE) A < gl a ,a e4 ~ ~- \\ /\\ g/ '~ I ,~ 7HERMO 7\\ COOF ES \\ HEATER \\ monu ces secTioN g7 gg g 3 O y 0 o x \\ l \\ l o

NO MDDLFtCkitCt4 IJ y seRGUD

] f J /HE ATEb JU N C. TION ~ f $y,< o,- =33 ~ THERM o COVPLE PROBE 'h.. ~ / / / s p.110T LEG I i BOTTOM END Q ..I GUID ANCE LUG e. f/' Stton.T CER PLUG g C To FOSil!ON FROC.E m- " uuot nr 1704 019 de o 8 g ), I Page 32 of 43

i

) 9 )e c (Y6h.%Wgh 9 ?' l I l l l .l E R R E h I _D R k D M R n R B n o O c t. A A r E. C 3 t S R L 'A L E f J y_ l E N M N N A r A t H L P u P C V 2 c 2 R A t i ~ e i R t. r a n ~ e t e v s w r. an A@v a. m + aF' O v s c o F o r s a e E s s u A at s Y R N Y L: 1 B-L u L 0 F P P P c u I u m' y T S E F S A L R sR E sC R Aw U CP L o 6 e T t F P n N / o O c / o C m M D-N n C C e A T T E E J J nr H H T H-9 a S L e C Y F 5 s t_. r S o5 e WV-e E N x R s EI C_ oT a V e P A r N a E O C C e T R-T I n T I I I H I c r p l l 1 s

l g

l n f u No ON

1 U o a

D 0D U D on U radio rinQ UCtJCY [n.r-D t Roar:. g'/ad FJ 6u R E 5~ y A ,y it,CMOTG Cunn blGC.GMNECT i, CRAY Locn cowsccTo.t o ~~ ~ D AD APTED TD Git AYL.OCK ) ~ - VAC Atrr Pt.ctDM i RE Movr2 vffEP. PRESEDRG nous u j. l J .1.. q wmg .(f3iul W s N,. l \\ 7.;

h. T~Asx.

<f l t / s ..^ r\\ pn .y%)- 3 s.\\. s . d.. n r, s c. . ~. g t, . o e,l _o._.o \\y DI I I \\ J1 \\ \\. i . :.'el N E'h.'.s, IJO MCDLFt:t.Tict45 t\\ s )W/M i l TO 6H;tD P.D. / Val'l ut g.l. I I:~. .I_.-d 4 ._=. ' ;==.; C.R.EPaoc.t 2- .r \\ e w m . M.! ' 1 .,/. y \\ ' \\l i! p}j; t r E A_. -l, } ~ 1 ') l l'~ W 1-M 3 \\ -7~[~ i s l GOTTOM E ND ' {, I r[- w[f GuiDNK5 wG .,.7..? 1..\\ ll 4 {! ' l' ' I,1( li. t !.!! j-.!!, e I 1 L,.v0"UET CGA PLOG-s 70 l'OSITION IT;c! 7 !)2r.2.jJ. y, _J.s ] lr

;-(GLtDE FIT) 5 s r. &.c..,v 1704 021 Page 34 of 43 i

['LOATit4.G. S.PH. E'R.E wtTH SOURCE MAT't- ~ ......:. '. E,tW RT /a... .s 's W.CAMT pl. cG A ... POSIT IO N ..s_.. g/ ..~ g DETECTOR MCONWD ~ y t.N PLAC.C OF CED M e =,, - - 4, ,,-.g.-............,.......... ~. : -.s.. s. q .1,

,_ o.,.,

'\\ \\ t \\ e 3 f d 1 FLDAilt4G -(

gtQ cnts$tt:4, ;t;

/\\ soce.cE covra neo uows 3 .k AND PROVIDG Go lp AMct-ebb b lq h b [.,~.~~ ~ ~ PROTECTIC:0icR ~7' a .n; \\ ru m m seuecc. 0 ,v O o s - A T a u r a n w.av.sst.c 3 ^ >l l. .... _ :. o.-... y. hg .. ~,.. '. guf. .,.r., +6 45 w

5..

~ ~ l q.y h ! =,- T- :- '2 ja v t e. <. ass Ante s.ctssutivsc. 6. ~ / 3 ... v R,l / ,C,

retssuu, g

t WIT H H. g ..g / ,v, .,s. ~ <...'/ i L / x fs* G *;u:cc 1 .5 1:11 / 5 / s \\ I w - - y. ' s ~. / 7 'o. o f.ccRCEb j v[o.Mct 19 $PelCI C / o.w rua.p .i,-

x.....,..

n . Q.. s t /, \\ y ~' *~ s s. ~ .A.L %. _ . d.wa.n p.c ..... o c.c.. 1704 022 page 3s y a

@,M All. SPAER\\C AL SodR.C6 v .i'; . F"I b u R E- 'l .s. YAClutT PLCEA Y-FOSLTton orretc2 v l y ,..g......'. =. a .a. )4 gy n Q, Q Ns' l.' /\\ ( h \\ neam ..,Nx sw.u e e r om.1 L ,s 5t s &8 y O (' ) O o \\ e, T .....~......../..- .\\ \\' A . (L '. - k. . r.w ~~ !!.. /{f,

~~

i 1 yji ,,..s PE ftrC<tti'ED 60:.'RCE C01DE WE! I dL .f 'l;'.,.,. _- =-!--.! --d-tdLTll tinCuS\\0M IN j T-HEAD FW.E A.. top Imr I /

l.*I t

15 P.EHoVAGLE. (MiucR ~ ~ '[. car.nc,ES To W:Gt.::inicb / 1 -( mTEntut.5)- '/ l y /\\ / + g i 0 .fp110TLEC 4 t p ,i t .e i ) . f / 0 () cer, r>uic c t a t , r'..a g',,f'/t rxNtrt. c,omt iu.9 / rc..: t. n t n. en teoutu cotut.vydn ~2 JIOr.r.I5.$--3'XI'.,,'l :lVI.\\ZT T'. d,'7"'[ l 1704 023 Page 36 of 43

D**n *M MTUAL lD l' S D ee 140t.vtPt.E va ccTcR sie.iNcs SOuti,CE toc /sTt.T) to ccn sillioon

Fs 6ta R E' 6 t

+ ~ -.,i r -- < G o to cA stif.rr 15 RE'I tI4 .f'*'.., t Y ,~T

)-.

.c - ) (}i, ya._. l SHROOD T SMP.OOD 6, I EYTTt45. TON tatTI) ~\\,] g,.]I g I S00RCE 140LDER / /l.g, Soup.CE IU /\\ [ ! \\, g-[NOY' ' ELE- .. ~ i DUTECTOR crniscvs n .i l8-.. -\\ y s 8'l } BETE CTof'.S IN s \\ i l top GTRit4 G. g N \\ ) 'I,' { ~ \\j; ~O o. .i \\ \\ 'T-O ~ ., ). .\\ 5 - 1 g 'gs3?Y' - 1,i,N.u.c ). I'\\f c;.M. ~ ~ ~ ': ~ / _8.s. I r 4<,f f s "c. ., =.

l. -

k.;Z / ~.i l ' Arn /.) f re-- w w rcI etsre l, '..,.. lt9..'.. 52. l l s l ,l.

== r.. t/ \\s ./,. l,.* LOCAL. SCORCES Il r in CE n CIWJ.'.23 l i[ toc ATuu t;truitN ..d , 's l '.I. - ( bi / TROCounot.t'oDS '1 l /. s, t- _.6 2 .a ay! ) l ocur. rotes in 'i ..s 0 07T014.S'i~ntitM g x .t jl ., l. l, ~ I d g,110T LEt~a [ t g ') *

{l

. l:n ,]a: 3-(c J _r.'~. ly 'hDS - t 14Ec11M1lChl l.i -). . 'c:auntm

1. 7~ -

p'~ J j)1 C ' -._. ,j l j gl y exis.Tns Icr TOGF. t !It ! '..' ! ii ~~ .g......-(..-)..r,.g..j-..i),3..,...,........$ s. t 1704 024 Page 37 of 43

SitutE S rmt4G v/irH &LT.RRt4 AT.1,t3 6,,,50,u,RcE'S .c g o oerscront, .. p., g gc q t. 4 g ~,, ~.. _.,. *. ,g s. [. \\- l -. j n . h.,x ,3 C I i. .i-c s m te /\\ l_\\ \\ r q, ,'I s .,. ~. \\/S j g q l} // { t \\ fv .a 6 a

c. a lO, x.

- 5 ~ Al-l.0)l I /\\ E:tUfd.LT 0 0 \\N srAc m L~ r.- - sow.cti s Y I I l I N . =$ (, W( sFTccTcP.s y,Q. p { L . / 4, i_ a 'i. _ 1 m #mm mee,are ..u. E-.-~; '=.Ev'2!=E=2] i. scmu l ~ \\1 ry,!: = ./ '--y' T -) so0Rct" ctarcTDP. PlatWD

p. p. IloT LEQ circor s

t e -**;""~~l = P. A } /g' / . li; ./,/

l

..r .g j* l- !! Ji! .) - ~. 4 { s!!I N* =*.8 8.l .....("...g _g .-s....g._ ",g(... ., 4 y.,,y..,, n -, -a i 1704 023 Page 38 of 43

F LO ATING D LP STICW/ REED StJI TC+4 $YSTEt1 um u t'. v,r 190Ni Rt1:1) t,tastc n I~I O U 'i E 'O PMtTn00 I!40tc/tt10r4 i y / l SYGTGl4 j i '"[ f H8(IICT y i, I VAC/ HT Pl.CE D M l / t uc.uT uti citT l MAGNCT 00MCUT POLE-p [ -7)4 8,,,- NE.W E XTE11 Sit,M sH A FT guide UnTil Eonct4 EX1TN90N To 4 l T COV=R ' FLCh1 HOLE C AN D I' ROV IDE. \\ /g[\\ - };- t\\\\ %i C-d!C/ N CE AND PROTCCTION N N ijM. \\ [. \\pl FOR noATS. TOP. HAT IS /jp.y "" "^** ,x \\ Jy h.,/ g PRE PRESS 1. Ri?.'Ed. 's IN o '7:B c.R TITAN t' :', J ~ STJ6RCS g- ~. \\ 5 88 -3 O. O ,O _o 3 ai. j L' r 1 u.- g 5 ~ \\ \\, p. .. ~......".... g-I;_C ' -~

4.}

.....T..'* ,/ %'Il1 .. 'a - ;+ t . I, n [.m '. - -j,, s .~- ,i g.) s ~, l

M_

s. .,/: / s w N.(Y { _/ ~- .i s ~ ->libT'LUCi (\\ g

}i M,

.s ; L rtetsys ti sotou ]* i 7-F0 E,rt~t O N 7.-,< (,7 H n. M ..i hi . !,r an

i,... <

t. r..y-i.

s +,, m..... s. I j a2-17 1704 026 Page 39 of 43

D

f90
]D' T ]

A 'o.L d oN,X h k, N ,.;dlTIPLE FtoATillS SPHERE 'g A . E.tj$.t.:V to S. td t TCH S15 T C11 c i FeetRc 11 </- St\\tCl:C LOtitt Hft,ItCT* q.n tQ G i) 4 e., 3 ~t'QECD Wi rCil POSLT'(Cl3 tubicA11ola (t/STQt w]' = 3 t== f. 'v LIGilT IJE.t0LO' SPr.cGR SPBCiMG n ,' I.y - VACAth* Pl.CEDM, r; .2 y,0 Tot REl40VED . -j ,.j.; .~ --,.7 t e 's sR '2 D.-

- ~ '

s /\\ l?4k l._\\ N \\- ..,J g d ' rtor riot. L_\\ t: cosaces \\' 2 ~ ' '. N i X \\ t .s ~. M- { d. p. /:\\ \\ j O ,6 o x l. l \\... ~ s. .,....u... y..g

g., -

.1..n,.. 6s@,1.mw y.y. s vem,ww / 9 -ud=_ -o ~ c e a ro a. .a... n m i i a x x /, gum tca l Tf n o g$ c a l

v. s,- m a v t % m w "..

1704 OU page 40 or 43

gji - - = -. _... -.it. AS_O.N. I C, P RO D G. Utf FibuRE I2. e 8. -p Ut1 R A SOMIC Tilhl4 5 0VC.C C.S / l -- RCIAOff!, ClutCK D\\sCcMNCC T 1,,. C t*Nf Lf Cn; CCn:4 G r.TCQ, f ',g- {, - ADATT( t1 FP.ct4 Moitt*. 415Cg TO GRuttoctC 1 ,y --- ultuS'ID PLCED M ' l f ret 40vt CITr0. 0P.E55uP.E HOUstsc p A~ A ND CCDM Moir.,a 7 l. 8 ) e -r .c ,-.\\ls s.' \\.%} 7 s s ~j \\\\ s 'a k %j g f \\ I ~(. g g s .5

\\

b[ } '# ft '~f-l ~ O! o O .~ 'q p ~ ~ .g ' I ,y N, y, ..L .'.n .- = v. ~ \\nq - g e,p - p "noe. nim arwas ,i TO.6DROO.D 's', i, e '.j' '. __ J. _..J.,.j d..... .,i,,.. .-.~--- j9 7 r. ifi A;m y. 4 "% DtW.A50.ntC tJAv::coibE, ./ p _3 STEP,iED Oit OERT.rsTCD. f CAR (>)li11 InFP8,0X. IT STE.i's. " e.. E'QY tKg l m, = ~ i .j; t { t. \\.,,,,,. - -7 ~ ~~) / //

l sl i

,j .s 0 \\ b. .t * : /

::t 3.a u I' 0 j.

,U! .tJ succ (t/.nt86 To (.,.rn : f.. s-:-.- y., r. Fg, i oitrios.,r.u. f. py-g ---. J.,,.p g /J -? -i 7 i. 1704 028 Page 41 of 43

'O D

D W }A n

u1 3 a Stech 9I_fpR.c...E..I.ND..IC.A...T. IOld PROSE $VU l' I~. Cec Ct*ll. y / 7 ccm.,e c.cc o.:.c,. war ganycur c...ae.v.a .g-G. ip / ,nonerte (wotsy 1. q&fl.ce -a fr man i , - ususco PLc.c om / Remon o ut paa. Pms..M flo,4 0s.Jf, ( CCOM MOToC -] g f ; a oo ssr ~ 15;r. iq. \\. j q.- J-s 1.y / ,s t. j /\\ \\ [_! . G. 2 1 i. \\v ' 21 b. ~eP/rocsr .\\\\ s e,..oc.-ec ,7 ~ n 'pr. s .I .) y> a (a \\ cl c. \\m _l ~ s o s i-s 3 i I I I 'y'g K*\\ \\ g .[. m.ot C&a..r,..,s. t' \\ \\ .rs ', % c... o 1.. ~ s. 9,rq 5'y. : + /,./..,. e 4. 3. / }Zj l . m...,.,.. % 3.-- m. i ,/ t. I f / ^ I._/ /h.s j% dit . I),o I J w t.ca ~ n- .~ \\ // /, i. e. I li il i 1 ,1 c g !.....-,..j..:. s.. g.i:: Is _... l., .. q... 3..... 3... s. ......;... r, g,yggc,,, g 1704 029 Page 42 of 43

E XTcRu AL STAnopsyc wers A Vt.OhT SENSOR OR DP CCLL Fuc,ues 14 M' R Conocn:A rc.co 7- , y (.1TRILTt.4b DRsnec, E N T rt A u s t, F U .% CAT wird M aler.*gr-/ ,S .~ .DETCCToR% s I = / I \\ / / HOT Lt'b \\ za 7 Fr:rascrissa e.n<,. DP TrArt:n,r:cg - + n.. _.-_ m-i 1704 030 Page 43 of 43

Mn s APPENDIX B TO ENCLOSURE 1 BALTIMORE GAS AND ELECTRIC COMPAhT CALVERT CLIFFS NUCLEAR POWER PLANT UNITS 1 AND 2 DESIGN REVIEV 0F PLANT SHIELDING AND ENVIRONMENTAL QUALIFICATION OF EQUIPMEh7 FOR SPACES / SYSTEMS WHICH MAY BE USED IN POST ACCIDEhT OPERATIONS (2.1.6.b) NRR Lessons Learned Task Force Short-Term Recot:xnendations NUREG-0578 1704 031

DESIGN REVIEW OF PLANT SHIELDING TABLE OF CONTENTS 1.0 IhTRODUCTION 2.0

SUMMARY

3.0 SCOPE 3.1 SYSTEM SELECTION 3.2 SYSTEM IDCATION 4.O CALCULATION OF RADIATION FIELDS 4.1 QUANTIFICATION OF RADIOACTIVE SOURCE RELEASE FRACTIONS 4.2 SOURCE TERM MODELS 4.3 ANALYTICAL SHIELDING TECHNIQUES 5.0 ENVIRONMENTAL QUAIIFICATION OF EQUIPMENT 6.0 PERSONNEL EXP1LJRE LIMITS 7.0 LOCATION OF VITAL AREAS i 1704 032

PALTIMORE GAS AND ELECTRIC CG4PANY CALVERT CLIFFS NUCLEAR POWER PLANT UNITS 1 AND 2 DESIGN REVIEW OF PLANT SHIELDING AND HIVIRGIMBITAL QUALIFICATIGI 0F EQUIPMENT FOR SPACES / SYSTEMS IN ACCORDANCE WITH HURm-0578 (2.1.6.b) 1.0 INTRODUCTIGT A radiation and shielding design review has been conducted for the spaces around syste=s that may, as a result of an accident, contain highly radioactive materials in accordance with the guidance contained in Section 2.1.6.b. Systems outside the contain=ent most likely to con-tain highly radioactive fluids during a serious transient or accident were selected for evaluation. Large radiation sources were assumed to be present in these ;;ystems. The rooms containing the piping and equipment, of these systems were identified. In addition, actions outside the control roo= required in recovery from an accident were identified, together with the time required to accomplish the action and when it would most likely be required. Roccs containing equipment that may require operator action were evaluated together with access routes. The development of special post accident procedures, installation of additional permanent or temporary shielding, modification of existing co=ponents or syste=s, or installation of remote instrumentation and control was addressed. Drawings showing the floor plans of the varicus elevations of the auxiliary building were prepared showing the general run of the selected systems piping. The locations of the required post-accident operator action were added. Post-accident radiation levels were assigned to selected roo=s. The radiation levels are not the result of detailed shielding calculations but represent good engineering estimates. The accuracy of the dose rates reported is within one order of magnitude. Due to the time allowed, the design review has, of necessity, been limited in scope to the post-accident recirculation systems. In addition, only an accident in Unit 2 has been considered. No consideration is given in this report for high level sources in the volume control tank and waste gas systems. The letdown line which automatically isolates on SIAS, was not considered. Core damaga, and therefore the release of radioactivity, would not occur before conditions activated safety injection. The study will centinue, expanding to cover both Units. The design review will also include the analysis of the effectiveness of the containment wau s as radiation snielding. The examination of safety equipment that may be unduly degraded by the radiation fields during post-accident operation of these systems will continue. The radiation levels in the various rooms will be refined as edditional piping is considered, and will be revised taking into account the installation of additional shielding. The final drawin6s will be issued to aid the operators to assess potential radiation levels during accident conditions. 1704 033 1

2.0 SUNMARY The,.results of the design review as it has proceeded to this date indicate that problems exist which require further study and evaluation, and this work will continue. Problem areas identified and potentir.1 solutions are shown in Table I. 3.0 SCOPE OF DESIGN REVIEW 3.1 Systems Selection Systems outside the containment which may contain highly radio-active materials as a result of an accident were selected in accordance with the guidance of NUREG-0578. 3.1.1 Recirculation Systems The emergency core cooling systems were designed to mitigate the consequences of a loss of coolant accident and prevent extensive core damage. However, these systems were assumed to contain significant additional sources of radioactivity above the original plant design basis for the shielding analysis to ensure that operation of the systems will not adversely impact operator or equipment functions. The systems selected are shown in red on Figures 1 and 2 and are: Engineered Safeguard, Pump. Suction Containment Spray System High Pressure Safety Injection Low Pressure Safety Injection Core Flush System 3.1.2 Extensions of Containment Atmosphere There are other systems or portions of systems which could contain radioactivity by virture of their connection to tne containment atmosphere following an accident. Such systems are provided with double automatic isolation valves or are equipped with normally locked closed isolation valves. 1704 034 2

TABl.E I TABULATION OF POTENTIAL PROBIEH AREAS REVEALED AS THE RESULT OF 111E SHIELDING DESIGN REVIEW Description of Problem Modification Being Considered 1. Core Flush Install motor operators on valve SI-400 and LA CVC-269. A O 2. Contair. ment Air Sampling Install new sampling station. O 3 Reactor Coolant Liquid Sample Install new post-IDCA reactor coolant and g containment sump sampling system. 4. ECCS Pump Rooms Install additional 8" block shield at doorways as shown on Figure 3. ta 5 Component Cooling Pump Rooms, Rad. Exhaust Install additional 8" block shields as shown on Vent Equipment Roams, Decontamination Room, Figure 4. Passage 202 and 212 6. Containment Purge Air Discharge Chase lui Install additional shielding as shown on Figure 5. Switchgear Room 311. 1 7 Streaming from Pipeway Adjacent to Control Install lead shielding in control room to prevent Room local hot spot or install platform shield at a lower elevation. See Figure 6. 8. Streaming from Containment Personnel Airlock Install 8" block shield at doorway between Room 527 and 526. See Figure 7. 9 Stair No. AB-2 Install improved shielding at elevation 5'-0" and 27'-0".

Consideration of systems in this categroy is therefore limited to the following: C(Itainment Air Sampling System Reactor Coolant Sa=pling System 32 System Location The selected system's piping and equipment were located on the Piping Plans - Elevation Drawings, and the size and quantity determined fer each room for purposes of calculating dose rates. The general run of piping is shown in the shaded areas of Figures 3 through 7 and includes the following rooms: Figure 3 - El. (-) 10'-0" Room No. 101 Safety Injection Pu=ps and Spray Room 102 Safety Injection Pumps and Spray Rocm 120 Contairment Recirculating Pipe Tunnel Figure h - El. 5'-0" Room No. 201 Component Cooling Pump Room 203 Piping Area 206 Penetration Room 211 Penetration Room 1704 036

Figure 5 - El. 27'-0" Room No. 321 Piping Penetration Room t f.gure 6 - El. 45'-0" Room No. 413 Sampling Room 4.0 CALCULATION OF RADIATION FIELDS 4.1 Quantification of Radioactive Source Release Fractions The following release fractions were used as a basis for determining the concentrations for the shielding review: Source A: Containment Atmosphere: 1007. noble gases, 25% halogens Source B: Reactor Coolant: 1007. noble gases, 50% halogens, 1% solids Source C: Containment Stemp Liquid: 50% halogens,1% solids The above release fractions were applied to the total curies available for the particular chemical species (i.e., noble gas, ' halogen or solid) for an equilibrium fission product inventory for an LWR core. The release fractions for Cs and Rb were assumed to be 1% for the purposes of this shielding review., Further evaluations of the TMI radioactivity releases may conclude that higher release fractions are appropriate. Houever, the overall effects of higher release fractions on radiation levels or integrated exposures are not expected to be significant. Therefore, the Regulatory Guide 1.7 solids release fraction, 1% was used in this review. Sicilarly, no noble gases were included in the containment sump liquid (Source C) because Regulatory Guide 1.7 has also set this precedent in modeling liquids in the containment sump. Furthermore, cursory analyses have indicated that the iodines dominate al'. shielding requiremente and that contributions to the total done rates from noble gases are negligible ~ for the purposes of a shie'. ding design review. 4.2 Source Term Models Section 4.1 above outlines the assumptions used for reltase fractions for the shielding design review. These releaks fractions are the first step in modeling the source terms for the activity concentrations in the systems under review. Decay time and dilution volume also affect any shielding analysis. 1704 037 3

4.2.1 Decay Time For the first stage of the shielding design review, no decay time credit was used with the releases in developing the post-accident radiation dose drawings (Figures 3 through 7). The decay curves shown in Appeadix B can be used along with the time-zero doses shown on Figures 3 through 7. The coefficients taken from these curves es.n be directly applied to the indicated dose rate to quantitatively assess a room's status quickly for any given time for a post-accident operator action outside the control room. 4.2.2 Dilution Volume The volume used for dilution is important, affecting the calcu-lations of dose rate in a linear fashion. The following dilution volumes were used with the above fractions to arrive at the final source terms for the shielding review: Source A: Containment Free Volume. The volume occupied by the ECCS vater was neglected. Source B: Reactor Coolant System Volume Based on Reactor Coolant Density at the Operating Temperature and Pressure. Source C: The Volume of Water Present at the Time of Recircu-lation (Reactor Coolant System + Refueling Water Storage Tank + Safety Injection Tanks). 4.2.3, Sources for systems Source A was used for the containment air sample while Source B was used for the reactor coolant ss=ple. Source C was used for the engineered safeguard pump suction piping, the containment spray, safety injection, core flush and c~ontainment sump samplee 4.3 Analytical Shielding Techniques After locating the affected systems and formulating the sources for those systems, the next step in the review process was to use those sources along with standard point kernel shielding analytical techniques to estimate dose rates from those selected systems. For rooms containing the systems under review, estimates were made for a general area dose rate rather than superimpose the ms.ximum dose rate at contact with the surfaces of all individual components of that system in the room. For corridors outside rooms, reviews were done to check the dose rate transmitted into the corridor through the walls of adjacent rooms. Checks were also made for any piping or equipment that could directly contribute to corridor dose rates, i.e., piping that may be running directly in the corridor or equipment / piping in a compartment that could shine directly into corridors with no attenuation through compart-ment walls. The final dose rates indicated are accurate to within one order of magnitude. 1704 038 6

The dose rates tranr=itted through walls and floors to the other rooms and elevations were added and are shown in the respective room on Figures 3 through 7. 5.0 ENVIRON! ENTAL OUALIFICATION OF EQUIPMENT Most organic materials used for insulation, seals, etc. are generically unaffected by integrated gn=ma doses below 100,000 rads (carbon). For the purpose of estimating integrated doses to safety related equipment near the system components analyzed in the shielding design review, the - following criterion is being used: Those rooms showing post-accident radiation dose rates between 500 R/hr and 500,000 R/hr are identified as potential problem areas. The rationale for this is that the total integrated exposure for sources listed in Section 4.1 could exceed 100,000 rads integrated dose in about one or two days. This was ascertained by developing a corresponding set of integral energy release curves as a function of time for the same sources for which decay time curves were developed. Section 6.2 of the FSAR states that all components of the engineered safety features systems and associated critical instrumentation are designed to operate in the most severe environment to which they could be exposed in the event of a loss-of-coolant incident. This equipment is designed to operate in the containment des 273F,100percentrelativehumidityand10ggnatmosphereof50psig, rads in the year followinS the incident. The problem areas will undergo further review to quantify the potential severity of any ra'diation damage to safety related equipment in that area. Should the results of this detailed review indicate that it is necessary, plant modifications such as additional shielding or component / equipment replacement with qualified equipmeng will be initiated with every effort to have such modifications ce=pleted by January 1,1981. 6.0 PERSONNEL EXPOSURE LIMITS 6.1 Personnel Radiation Exposure Guidelines The general basis for personnel radiation exposure guidelines was 10 CFR 50, Appendix A, GDC 19. The following additional radiation limit guidelines were used to evaluate occupancy and accessibility of plant vital areas. General area dose rates were used rather than maximum surface dose rates. Contributions from all sources were considered. Further detail evaluations are necessary to further quantify the problem areas. Vital areas requiring continuous occupancy such at the control room are limited to a direct dose rate less than 15 mr/hr. For areas requiring infrequent access or corridors to vital areas, the dose rate must be less than 5 R/hr. For dose rates greater than 100 mr/hr, a man-rem calculation including time and motion analysis will be performed to insure that the integrated exposure for an operator action shall not exceed 5 rim as given in GDC 19. For dose rates less than 100 mr/hr, a man-rem calculation is not required. 1704 039 7

7.0 LOCATIO:! 0F VITAL AREAS 7.1 The following actions outside the control room are required in recovery from an accident. The locations are indicated on the attached illustrations, and the rooms inve'.ved are designated as vital areas. 7.1.1 Core Flush If the accident consists of a reactor coolant cold leg break, an increase in reactor boron concentration may make a core flush necessary 8 to 11 hours after the accident. Core Flush will be performed as described below: On Figure 1, (Coord. G k), line 2"GC5-2008 is provided to allow L.P. safety injection flow to the shutdown cooling line 14"GC5-2003, and in the reverse direction to the reactor, by opening manual locked close valve SI-400. (L.P. safety injection pump suction valves SI kk0 and SI kh1 must be in their nomally closea position). Referring to Figure k, Valve SI kOO is located in Room 203 - Piping Area on El. 5'-0". Access from the centrol room is by Stair AB-2 and Passage 212, and it is estimated to take 20 minutes. As this is a high radiation zone, the valve will be modified so as to be motor operated from a remote area. 7.1.2 Containment Air Sampling Within one hour after the accident, a containment air sa=ple will be required. The present station is located on El. 5'-0" in Roo= 203 - Piping Area - as shown on Figure 4. Access would be by Stair AB-2 and Passage 212 to Room 203 As this is high radiation area, a new air sa=pling station will be engineered. 1704 040

7.1.3 Sample Systems Sa=pling will be required after an accident. On El. 45'-0", Figure 6, Sampling Room 413 contains the equipment for reactor coolant and steam generator sa=pling while all radioactive mis-cellaneous waste sa=pling is located in the Unit 1 Sampling Room 444. The reactor coolant sampling system is enclosed by a hood ventilated through a high-efficiency filter. Interlocking high density concrete block shielding separates the hood from the rest of the room. At the present, the reactor coolant remotely operated sampling valves are closed by the containment isolation signal, and the area should remain accessible. In order to meet the requirements of NUREG-0578, Section 2.1.8, a separate post LOCA sampling system is being designed with the existing system remaining to be used for normal planc sampling. The new sa=pling system will be designed to obtain a sample of reactor coolant from the primary loop and from the containment sump, and will be shi.elded so that the maximum dose to personnel taking the sample will cot be exceeded. 7.1.h Hot Laboratory The Hot Laboratory 522 and Counting Room 519 at El. 69'-0" are required after the accident for the analyzing which may be required for evaluating radiological problems. Analyses of reactor coolant and radioactive miscellaneous waste are performed in this laboratory. Access from the control room is by Stairs AB-1 or AB-3 and Corridors 521 and 522. 7.1. 5. Control Room The control room and proposed'onsite technical support center will req ire continuous occupancy with'a dose rate -criteria of 615 mr/hr. As shown on Figure 6, the control room has a dose rate of 5 mr/hr., and che proposed technical support center has 1 mr/hr. However, the review has revealed the pipeway in the southwest corner of the control room is open on the bottom, and there is the possibility of approximately 50 mr/hr. shine in this area of the room. It is proposed to install lead shielding as required on the wall in the control room, or alternately to install a platform shield for the pipe chase at a lower elevation in order to ensure an acceptable dose rate throughout the control room. These proposed modifications are shown on Figures 5 and 6. 1704 041 10

1 I 2 3 1 4 l 5 l 6 %. y....,w r e =- e,._ m.......,,,,.v2..g. + 3.m. m-q .;= m.. w " p M C' n1m-. o ..o c J _f. R en ._acp%y& c.$@fW ( i \\ r_ e i VCNQ A o-4 9 9 9 6 e o..r

c. g m.~

o g"L ny w: a n A m.-

e%

- 44 E _g;;o:p,..o~. , 4 a.. A ' = > $- cm3 'i'"- ' = > p

44 g._.

n,

g e-

.,... m. -.s .k. *)_,. >% = -a - g g t ;); _M d. %.- A.g'ib' Wrg#( 4;. t L% % % @ n? ~M y y 4 ' F T. e g *a.. d W99 9 1 i * @Nh ' F_ ras g Ag. %,s r i+o'0: &... %,g;. d }9 7. = 4.,.e_ 4 tE @_PFg

g y

<_..s o p (c=> e m, y... f,>xw. w. 4.g e n (M .. ?".d ~...W g. " eb..'.W h. M,Y n,8 e m,_ ,,,.J 4 i ~. .. 2, /m; ~.qp. qD:e:.ss m.. .m. -w .y 2 =o p v e . s, y,.-- sw -= h n n ->." ' ~ ' v (7": x T..* x. p' Q *9 [- . 4,, b.g 5e...rw i jy y g jg =eh L .6 ?.@ -t o

g. 4

-... m. 4.. g. y;ep..~.m.. 4 .2 y.. a ...e. r _- - y w ,. s y%,s. r..,p. 4 1

. m '-

...j, e. ri, n

e..

=nms 9 C< x.am.-m 7't - $= 4.rA.+ 1 = r,,,g g..g.u.:x+.4..-=>. -.,., n...-n,.,w v ... e.o -, m. p g w.e r. s e 9 m q ge._.., .cg_ ,. gp8 ,. ; g., ... e a. a q ,u p- -m %h5hi?g,,9 4...g4,.h y_a_ - 1 e _ 09 .Oa t g' o .. m,[ o 4 2$'f@'wl m 55?e" g.? kN rer? M., cyg Te -m ,.~=ix@ f. C,.;s Q. m...m. f... e".~T+ 4 g.,,W.. 2-.- 96h_ i4 il m +-~J<m, .2 o ~;'.".. g.cym.-m,..:., % W-g ='.,_ g n 4 c. s.v..y ee.os a { ,p a o a ,1 m. .~ ~ ~ ' 9.=

%[""U,

, 8' cD E.fr,sL a <.s, w ec,shoc. F -- - 1 8 gm g y .n m g ... gm..p_,,,,y_ _.1 - , g 4 _ m _ _,e._. m '~ ..w.- v w u - M-im, r.. 2 i.a. w_ L,r,,= mgw_ p

...Q3g5.......

.t t =m 4-( f 6 @ ' ~'~~ " ~' ' ~i @ @n~4-hgh-@_.i c p ~.-,.-.......~. -. = _7,p re-c 1704 042 <. ~. - og., g i 2 i 3 i i 3 i

7 I 8 I 9 1 10 1 11 12 1 13 b~( E 't& @ $ " ,h u *.' \\ qsc e rno) (=t if'*N f ',' g b A g r o 5) '(g < w.1-- r t%.wc . h g 6Aw u r-4 -,x,h m t**'*.E

- I J w@-m@e :

~i =, M D ' ,M G 'C 5 I Ja!. "= 1 y t -..t3,. ck h"'t t

N r~rE )U as

-*-M 913 fL'*M I 6

")

] h = "T @ *M-w a ti2

O'T.a

= p i.,,

-;.w' m

<m . m .J,r;D Qg f[ ". {; g f cm.,.. ; .. Em pg- - p g i g ,g: e.tse I g' y s s sv.ats. ,,_ r 6 s svoev3 g, ';g jp g B _4?H9 go.1= h ' ',M' 4i lb.. 5 ),, h ,3,f i E a h , s,m e, 'pe.nooi>f G j V., .3 o-a . i., b ^ v V 4 g4 . og t g l .s e nN Y ie 5 b b d b M f e'r"Al._w.*'1 % "g M. y r a, '"~J L t. 48g y u .<,.m., 3 g ; ~ <- i..e L q q g

m..., <

gf e 1 k E d.3l .......to.6 ( h W 36 savt pouwe coa L J"*h]yg 4pgg' 3g,. c q ^ 'f_..-. ..s..- . r;- { t avr a't. ar, E 3,,p1 4 as v. m f '#h Q*g. (4'6C mc4 (

1a w >%

c q1g @~k/~ eq

iPii* a$1Wh51 Ni !N-,f,P-..[T 9

Ced...ra il , c t y 1 ,..'=.,_ 6,'1%i ,i @.c-8 b n p.

t n.4/ '"h D

n +,.. u - w.=, e g! g.gj j'y 't=" ~h'# erm o s earre n.t e waa*w=J M

h g

g I e .$4 Nap u..ua s,wsat entes p'.g fM.q vr-y E h). i s e -- - - -o-na-l ~ j @f ' ' l" ",ll,,',L",,,7 ".;*C,"' "" '"" '" p A

g. $.

3 n .....~e,, r-- T V a

s. easiese,ev.ais sa. sis st. 67so-.rsco b.d

(,J d [ 'y ae e gijg" [3,,[h h ' % ",=%" "*,7 i. ;", E . @ "p:::'"'u..f,s.L

.. r,=.r.t.....',r,.,'o.';"" "" "

4 h> ,,q gn

=

.g.. _f). M7bjj gm) g #h~ 1 ->m@#W g erg j o-r e-. e_," d E @ d) F _ W W 'f_ d #, d._,, s.MA9e%~dp .'.y fi ~ Y~-- "'9 3 0 g' D C savioo.vu--"gigd 9 e ? a* g 3 og n t g L g4,A,l, z u s,_n .r u m X --'u~ c ooomer, 9* a.3oo.pN 8'"' j w i n, e sc= aar,a

rros 6

o g. <.... s J, $s 2A,,,7_#-g,. 2 c. m g.V 2, 2 % n< i; g t cm wG ma"1st t{ f Q,J3 n.s siyta. - Y @ =d~w"#. ;* MNUNATED PIPlHG ~ r-sutovin ta neo g

9. m o o

k pip!NG t INSTPJMENT D'AGRAM H HF MQR., U,',S? + p" m + m m) "'* Y F" '.. p.<= s* i SAFETY INJECTION ( CC.dTAlWEN'

==.Re sts r*=Sa oc wroot ) _'3; $ PRAY SYSTEM 6.

s....

ge UNIT NO. 2 " ^ '. *'0*} . Ya n.tc.TRIC PRODUCTION PLANT D$rY "y ~" CALVERT CL.F71 UNIT I(2 .o.s 1r s, co.se um

~

B ALTI.M.O.,R.E ~CA.S.A.M D. -(,L.ECTi.tI,C CO.

=

4lf 1 i W BECHTE -CgPyATION j 10 [k "" *" Y 4a ,3.gg ~ M 6 '-257-E 76 '-+ l 7 i 8 i 9 l 10 l 11 j 12 ( 13 170A 00

F.' M!!.::....TC..... ' '.." ' I ,t.(_;- y-. F,.s h*P'-i;p*- g/s *,A

pm,adu.-[ N,a @

t f - 4', "',' j' J m,.J 3 95N9 .y ,f g 7 f < - -F - - l (.N.a,Y-G g .@h SMINh'.

E ~

i ?%% "5:w 6Q.c,,'rkn? O_$', $ k.- 9(1r'.kN,0?'..','* q ( - ----Q, g.w (.w. r u l w' e e,d v. }- :! _8' ' '~ '" ... - g,, *@y__)j, s.gji25 ;, g ,,,,, gi.z.,, f.u o"M ~ ' t Lx 'd W t !e *7" ~ - g p q o, .) ...<y.raj. r. l 4.1 <;md II,)T;==n Pr>-o ; r %~A M b - e. g d- @ j @,,, y,

M 9 ;g

~ i< ecc.,g b'3 '== )4'*aT/> e-o e %, t 46*-- ap-: I M , c: ~, El i d', 'a #6-v w ee<' [' a- "f" 6 < f~T ~F#EE) 9

N r

u r => h .( g 4 9 b-I o e 5 . t,,; i, n, m'. l P, " M asa; $ 4"' ur o.' I ne" g=no@ - ~ ~ &7n 3 ~@m

+

g' =' L-

=, %

gmg 3:qmwj<~m.> v gg 1j i II h;.. h ,t oa@., t,b @) ;. _ e.4 m,.: g-8 f ~, g tca v i ~E py_ ta M ' !T,' @'q! ![{ (- I + E $PsJ4 f.#y.oA. F__.a4 - -f' w J %u g l.!. e; MM.d,. m.h",Q\\ FfiGD ?W[Ec,.f..~!"#

t$j' e

m ..m$ 5 g we ma ,e m~ o g lhus m n l 7 I t N~ f k* i l f s% tR '= d[.$@ 3ENM3Ih kkhf D ' M i } s e

a-tg b

JT@r %'ip

! /(?,4 s - s. *T-IE., Ys l h c x t -, M *i J. r

n9

%->bm,;.:ad ~~ .,q d M "d$ @(a i pl p* A r, " A*.e M,. I euj. m=~ h

[

I' y t l, Pi g l__ j-2_j Mrs b,

@b.@fV{efh4)@ llg}hmD--. g Er.!(p/ lIld,hy

'5W r I j .p =r-I-j M

s y m@h k

a . m e w$ w. -, a c ]9.: m u. 8 e !La I gg_.. - y.. f %go (re [j! t _s.9 = F-- imlF [ 5 e1 it :. .I A

--- y: r,:,,],

3] p<;,u.a ,q g m.<h p d3 3-I g 4Q e g2 x r c. y. e. e 3 yr gu;m_a gg p 2. g m, v m $f 'l. [ I ' A' Y f II wa.a: ';7 1 }" (I '9 M* 9 r ' N =m 5 g F#m a \\

^,J s

r- ' *2 71 y vu > Q_ l((g;n.or)g

g 6

Mu r euz) anuaa m ei t [ __.i s_f.3 m = M r_ a"e t. a -, e ! !. w> <2 g, ~ r om g,..,y.e p w. i -[ T.i, " 7.1.g'm a ra -1 I g ~ 596%., ' ! jg p .s,jr 8 [ - W,d,% g.. o n . g...; x , Y ". ,q w =,,, y' g, o i m 7 Y 1704 044 I l 2 l 3 l 4 l 5 l 6

0,,i ,i, . L.g.r .. 3..,Wm/_L__ f 9. W.._ _ _., 9 t.. ..., p._. m.:.. g - m. &.. w,i +o 15-. LJ.... ,m - --L_..m..,,

n

.bm,, 's.... .7 i ;w. ;s;, - y 4 nN ..c e > . ~ - -,.s., .. fv h,. ,M*CA f 3 yh C E. 2 S *- 3, r.t [ o.., p..,"" Q im g Se i *. s e,,, I ma e.w 8. r. = ~ . l, m,:1 p, + 1 d j I %.ur-

2. A.m.,.fgi g. C. pig. a,-

p,,ig.- - _ _ _ e., g y,3,,

y..

r o ,, Is. d- %m.a.. - ~ um A v: n >....u. w - yec N ~ -- 0- _IT. r Y %e tr^4 t.% =7.,. p g, f u l =...,.,. j,.2,,, i4 s 3, ,J4s p.. a!j,.%,. ..m.,.- g,3. a=.L.e 4., s n r...-:...-. c o q,.

3. m e

y. e.,.

..y w... w. . ~.. V.. t ~. .m........ 3 n.3,. Q, v

~.,fv, v

[I n w.q re e e a < q" vs.p s...g v.<.. y I e +w ?, - ~ - r

x. -

, m:=n \\ - 1. - r - r gc ~m ... L

c. MC

.e 4 30,s a t4 s Y C # 2,' M< a 'y Y T, il E y <t, m. , E r. M l" l, ;, D,, TT. . -..Y..- k i < raw. 3.1,. ,4 4-1_c 1+.,..,,,.,..m g3 m,_-%,,, s g ...s m......=: r..:. ma-7.. .x....>

w..,,

a f p m __ .ph +.. - .P m ,.. _ _.-q..% n

a. :..=..

o;..y__c,.m>,. -w Ccmnot . ~s ,3, y ~,~ '"[ i * ") 6HioTAb/ E " *"?W"* k ' *f"*) l. , }a5Wt - T,"st_.- .w. 0 D 67.ii~.b----

v. ' Lf n_=.-.b_,<,.,r,.n,
V; 6

W H * ** .~.- -.a ch 4 -.n),,, ,,/ l y @. t. Q, 1 ' _- <- f '- )tA. g d,.,. k .,.n...m....... 5 Y e

%.m

...n.........m.m...,m....m- -wre. w., b.,wg-

4. r?

m .._i e u. g. fa ... -- n. m. - gh) o .g. -gp: ............m m m ,, a 3,, y : 4 " *, %.; 1 e. e-o................ wg --4 ( y-< 6 w..

m..<ma

-.. no-a -... o. a m > ws .4 '.*,;;""y" ' ;"',",' ';', "'""" " '" f -- -<tmHE M,nC(') c'd @s l t.wr -w G .I L 1-~ 7-- '.~'.c&; (L. a ', t - f_ ' ' u ~ " '- '"a - '" a ~ ~ a '- - a "" +- I ~.,; ti ,e M.[.,pg,,,$c-m.... w @i q. :,,, ~..; 2 ( -. g# g.', 4: . h. x + $..~r*,i W &, $5? Wi%t* *.*.s=> Y x- -x ~ g,,,,y s: _. - *-* -x x x - r ~a n. r,.g s> (*. 'c. y o O..U4PsitQi.t11 f e ,1f, i.'d) gw 4 ht p ,,. r"-- ++-w - - 9*~ m.c.& 'n I N k d A r ^# n. -, t l> " ' ~ N h,r[MQ ".,c.<h5k.., u[,,,9 f U P"'"b.":='

r F'%

x N, M " % 2 W '~"'"" d'.". 3,.s(( ,","') .g# -,~ m. I. k=,[.>".],Ae.--.-, ~ _l Q, D i j i t %,o s 4e .n <g, %s %. m, @y. ,.e s

a.

m, w;. n eg m.m. L,,e <m..o e--.m, 4 + 4 - r.. (Yy u, a,q u -~. "k a ,=r,. ,me.J,, .r " E u ~s rnD. 4 &W -? WV a-3 x y.aw.T M. g 4 ^ 5%-gD u s J' bw D P n r I G* M b'},;t r - CORE FLU.5t t in m' M $ <- k m,. "" H "k h.b IN Rto a .f T:.. P. q@.. a;d )' -@' m ?,o [f-J ij %e #H q ee

v.

a .,p t:. .k Q PIPING 8 IN51RUD.CNT CICAAM fh # -] h h E _s R m h., b [ CFCwCAL Ato \\OLUMC CONT AOL SY5 TEM h 2 2 y y { g YJ L') f ,J, a*i*Mr'- ;-' ~~~~ UNIT NQ 2 A'- R' '. .d.s.a.~ e.e- .---.mp . A,m

5

,(, I L Y o 3 C64.arAt 'll C - ' ' ^ ' 3C ".E '8 Mm.C CN,'LANT D Y )' 74 -* s_M_v ' w.c u, hI 4 CAlvurt curTs ut.rrs e a a i T5.,'s%_ (&*Mn --" '* f.e,. ' E> U n* ~ - ~A BILTIMo,R.=E CAS.n.ND..E..L.EC.=T.R.I,C CO. N' ,w.i. F E M~{ '8aG-MtA --M euc e== s X] mAzrgu. max +-- -- h,M M.d.h,'-g$5- '4.g k.k r,$ y

h BECHTEL POWER C.ORPORATION J

,1 ^ .--x

a"=2 m** a *'un g

y 4 g-- m e' y - e -- m.-< ~ l.%--41g ' FIGUP6 p. .*'" I N m. x s M'O M.'M?.'.'.? FW = = .n = < a.,.a.a,, ' SM s2 2Sb.E k l i 7 l 8 l 9 l 10 l n l u g 1704 045

I 2 3 l 4 l 5 l 6 A 4: 3 2 n.., l, L, a B C f sia. _1 ., -l l . m. J i RJ' i t.i?O ! D t n 7 - -- @H i% h# L,N o a w M'# iE~~ ~O m-( . y i g e= _=3~ o O '.'",'4!%. = se 0,o 0 VL UNIT I E G432 f c

% s iq\\L h

'7. g;h5. S' 447 F ..mm. \\\\x s-w r- _R m ,.,. a = og d ~r w se., a E ' imw ::ref H \\> l 1704 046 1 l 2 l 3 l 4 l 5 i 6

l 7 l 8 l 9 l 10 11 l 12 l 13 N s 4 - f ( i B KEY P L /+ LJ

m w N..,

L DOSE #ATES AGE Mm are AND AP( Aaunart irwn our enote o, werucc. C 100sf maits met at tuet 't% apite ag;oggg, 3 mt(=r 16 Pe t/wi 2 q-l

4. SNAC(D AF( A5 (C4 TAM ([fqT4.NATED PreeC,.

S. NFe wtLD mata 10 Pet ve%f statA., meet, af0 (W'D.re ,,d u-N2,b'h o a. /^ Mr.CR Myg. -m gg> sd E m== gr- ,M "/X" w.s: - % c* : t 1 9% ..w. iema F ~ =..,... ~,. . mN/ s oC@ w" N&-] -e 1, =m= a = = w" rd 1704 047 BALTIMORE GAS AND ELECTRIC COMPANY CALVERT CLIFFS NUCLEAR POWER PLANT F%T-ACODFNT GCATION DOSE 5 RANT StfLCING [tb60N FEVIEW AL*1UARY W ALDING RAN 4 FL, t )tO'-(7((PS -O nea:pr 3 [ is 1 7 I 8 I 9 l 10 l ii 1 12

i 2 l 3 1 4 l 5 I 6 ( bhhhhk)l 8 \\ {--

o. 3"1rf 7l ili c

m-u L. A -..:._ w i 1-.o., l g ra su s ~ r~~ 1XEG r ngvi -i i-,._..m. i 4 o a.. a ',AA'S cn w Lr mm.n ~ q wg g-w. ppm ggg4 WL4 0 emi.dP-*- C~ [W

'~

Q' E b / 2 9 M*.

2?la" *'* " @

N.' ~ i' %eW/ 't D t j;;,.,,ege @) x gu B Tj .x:n ,p o grrj q 'i .g ? , z.e.4:.w .s A -% M y x i %ssA (- '\\, -.i m 1 /. ' } m ma sw ME f[""**" .E f I, av::re== ? \\ so m y., S v j j qc..aw. a.m s.a*w w. 2 f 'Q N s% W ,.t i N r *y (k iO11Yb $,y 'J ' NM / *f w'r s y(ag,. 4~ _m w yv.E*= I ' l~ j n egn; e G)u sd 3 - g,_g y u=w. = =o o iE A '. '~ uranm acc.. i he Lta; O O -,ss - ' _;e g j~ / ~ m-yb#E m, 13 \\_ $$.m-J o G ! Ut. - W , [' (7 1 8 y \\ H t 9, ~ J 1704 048 1 l 2 j 3 l 4 j 5 l 6

l 7 l 8 l 9 l 10 l 11 l 12 l 13 l l +N- + N)- a ? I' .1 i W r tr. e .~m t ocu was A.t s+w% ne a=c ard _..., _,, em...~. m d 'I .45 ~ 2.CC54 jdT& t A4( Af let inM0 AF f t1 A (f Dt % T,

1. At(W(%f 15 de WT 2.

/ j.h k 4

s..Mai(D.e.Aad A$ ( A14pe (C %the#*b4 a.,tx o.

i / . - - Sy '/l c .,m s to -.., r. m,.. m" .. - v. e e... r p/ g.a

6. u.

um,. -, o, ._s..

g. m..:_T

/ 7 vm--/) > /gA.Ly .,,a_.... ...c.. m / 'i 3 rk j? Q 9 ~ ' g, m. c an 3s

. #i=-k ~N.

?. Dg0D 0 }L w f ",= s'rtyLy f U R D

x:.~

y-p m.., - A A3

==-=f"j 'g\\,, v j4,,, c ^97 Leas q / r} l \\ '"g ,7-C- /p j 'gL., ~ / L j g.. g.-. -. ~ \\ - - m b-E / , +1, y -.= 2[.,.hfe\\ yy,KEIAA [hf ~ Il. ( yp, 3 y .e. _Tm<. j.. n,x / n m m z-o c n==a n:: ..a f Drr&.- m e

,= =. _. <. a --

n. e s

1 o q,/g c H, 1704 049 l _ J BALTIMCOE GAS AND ELECTRIC COMPANY CALVERT CLIFFS NUCLEAR POWER PLANT POST A' Yf NT leOaTIO,4 C/yf 5 PLANT Wa% (15t>4 DE vt vv AtwuAE r MDr6 R as At EL Sk~ F (m if. l 7 l 8 l 9 l 10 l 11 l 12 l 13

i 2 i 3 t 4 i s i 6 A d l E 5 33 1 j j 3 3 r 3 s i a m = t. J . e a t h '3 ~gg, c lb d V _n %,+ s_m _. _ = m *vuvtcwt Ji A AL L AT ! L. 35-C/ m O. >J t i g .,a ewiecwo a samass o ,r q ty .mm,e n 1 ~ n-1 m.. o -mm, =w.)w,, Q g.v. m u j g w... a r a o*-.... m. __. -,.,,'**} m

i-

-/--- WrikJ + W s

b. l._ /n

% *.' -~ Q _ y .. e .~g aaw-s, x. g =:.. ( T . _ e- ~

E

( s mai'. L ji{ -- d. M... '- - ...;a - a a. +. y . t.. 3._ v. n 6 Jd"l'$l f_ .r. b- {4 .!gl@suw L i ME s/a. bngE., LW M e ..m J -.. r y ; g " w ",Y & E $ b,7. ~.i m, m G J m,,,,,,, / .:r- ?! /M (stoown. cat. i n sp ~~ . N / jg ' jl : .X" 'I t t - /E\\ l l H ' ' ' ' ' ' ~ ~ " ' 1 t i J 1704 050 i i 2 t 3 i 4 5 i 6

l 7 l 8 l 9 l 10 l 11 l 12 l 13 +p i h a -l a ( s ( s.) I. l I XE.,'~Pt.M a ~ NOTFR-e o#4 e. s,.at vs..w i=Dar e..w.st .a 4.CT*d it.'(*.4 Int.1 p0 AP Yl4.Cu{,I m .WM7 Erf.am if8Ab.l ( 3.tt.e f NT 5 pe U'w t.2. .,. o.. -,e ~...,, _. _ m.sm e.-..... h*Te ansq 4,4 rm ,s - x..:u.e::~=,.,,v,r.~m:wwa = ,_.,e.., 9'T sa.5 hew WN ( % f Li, AN.4 4$ 0 ' $. ' ] [ [' t e *C aM.G h P a *.. a m 41 g n 2 _IL-1, E @c -r = _._ v o m... u@ m.= g e,t a. ~ e f'.. ~ 4, )[+% cuanannam em. - C Q {gw na ~. - s,a o bSkOh0k khsIh, ~k jan,_ 't e v.

_.4._.__

t = s rdv y, j' O *. - - s{v i O.gM., ' } -- 1 I; f. ).\\ a, x-2.a = m v... /. k ,f NEO. \\.* .q En d ' A; =:,' pd CM-h)v /A g g e% ",. . 'qqg,- / dn; i +,2 u. rm t fl I. -r, Q[l.E. ' .131e;fLf -/ tt.5dlW u u.=]Q /q a mx . ~. .N H' .=::.: a.. M Ae, =. - 1704 051 r BALTIMORE GAS AND ELECTRIC COMPANY CALVERT CLIFFS NUCLEAR POWER PLANT POST-ACCIDEN T RADIATION DOSES J PL ANT 5HIELDING DESIGN PEviEW AUXILIARY BUILDING PLAN AT EL272d resuar s l 7 l 8 l 9 l IC l 11 l 12 13

i i 2 i 3 i 4 i s i 6 00k 0 Dpg ^ I \\

y

/ !?"..,. 4 ' '@~ B , :.m./ PL AN AT EL C _. a. QL' M 't >. ~! 1 vmqzi ,y ) '/ ! P

r% 4* 3 = ~$ b, aw c ar' L4-t

.g m ~ ~.. _, -,_.,._._ w. h[,..Y Nb ~ g g -y= fm s-m u_,. s. e m .- - 3 an.. u._ qxg... 'f... .;}j.g v. = ,= L s v

t. 1 n,..,.g.

4, e m ~ u.... \\ u. 2 e (_' UNIT I nc~ ! -,. : n:x m am.y ""# f ~g'""' * *1Ic==- &y"

h-F

=- b , :=. j ym Ln g - -Q+ .ne w ,ym n.-- I \\. Ed + NE d.. _._ _ _ i !EE:1 -- 4 ~ g ,f_~ r _*_*## G h W. - h j: i ll T g' _ , g M ***cw c g84 cams & maas%2 cess.g ] E E >*E I_,l =".... -. H 1 .\\ J 1704 052 l s l 6 1 1 2 3 l 4

l 7 l 8 l 9 l 10 11 l 17 l 13 t A I. .__.1 gg

r. q..1

-p p-se p-se p.

~ L. e6 to-be" .. 5 d.* B w:Y pt.Al e 6.. v .m

f. l*

~

1. 0t'4 nAf t).h{ Aw e p.

.g gqg e a :cwr.a. c r c o x,,*,..,e C 2.t('d 447tl seit A' f ast f(840 as Ttt 4(tstigt eBa.. nos.b.)

5. ACC';i4T 1:3 m Las t 2.

gin... 4, f a rear ag1 stof m (4.Wwatfir 80 esp /s.. I,1y.am.M_ .s.~ .,o wiu.,.m, is se c, c,c,se u nt.e r w as Per aar. (t a H p. , l w,y 5 1 4 _, } . ~,,, ... ~.., m., u ts;t-ya.y ~ a w-- /I nr-N .,,,, M.a m a. o p; o o a ~ w.- m h D y

x.,._ _ _

o, mq,,; e ; 7. ,..ww ~ bs h Uj,",) / ;;[$ O_=, =. c d. ,,.e nw 4-o s, 4_.. a_i:pm UNT2 v%m. man, r_ g, y m,,#=

4 p h yfq.A)"(Jy)e m

,e. I I h 1. \\ an. m R TN i r 4e gQi,3,,"gp N,[IntiV."=. y -. - w q _2a m

-w y

m p. e

'.711

-l o L,,. i f =J '#,ld', . ren r "g. ans. 6=_.w = KW}_3n,' 'T K -.7 1704 053 i _ C BALTIMORE GAS AND ELECTRIC COMPANY CAL'JERT CLIFFS NUCLE AR POWER PLANT 3 POST #C0ENT RADATON 07ES RANT 9fELDNG CEKN MWW ALNUARY IUllCNG RAN AT Lt. 460' s <o g s. l 7 l 8 l 9 j 10 l 11 12 l 13

i l 2 1 3 1 4 I s I 6 D**D

D A

^ 4 oo o 3 I t J B

msw" *,

d.__ _ ' x=- ; ;',.4,n w.s m.m . vn, is. \\ .r n fh h tu 1 %, m, JI x a C

\\. C Q

tar,-.<~. c # w..., .v. N4 .- g .. ~. 3, m- =,==-:==,__.__ p, j w_ii n'L..-' T. 4 L A , cr Tg # a y- -Q l :.... /- I m, n i j"i i r yt fM i w i w y D I( ,'t 'I ~~ - w a.. a * - 57 p ) / a ~~ ~ %v,,_ 9.. ( /. .cH. ; 4 ci &yt, -.. tL n_ afi e l g...-.-y ' l ,t': - '=tta - 1 U..n.. j, j 3 l N 7 LJ L 1 e a-M, ~~;;,-c.,; } x rl O.

mi

= / _] [ -- - q .c = l K.4. Il ii r, % = ( '3 E,......h....i p e---- gif .g a,_ ty

ft V ';

.I:3, .,g?_"_ 17f g~ s i b ..,L.. hI l lP}WW ' I uc. f, ~ l g* g tes =.. i v- ,4 ) $f.e%%364bleI'_ s_ , n. bcB&U. [ ___-_e,, .m T ]p~1 IT ~T

w.. /

1 ,,aj+

w lf a :X[--

'1, >.rcr e m n e,.~ n ,n y y a, -P

w..

a i h y',} q _v a gg m.t a-C, ma x g;7 y ,, t >~a w. A ( r a L2., em 'a ~ L .a y s ,a i T-... u, w.. m a=*;us ~N, ,.~w t- - = ./ un. ss. s, o- -e. J cssa ~.o { 1H - y A,-...* Pd = s, -, - 'y)W4* - _ _ _7 l 1 G . ~ i . me u_ .____] r 4 l

== 1turmas.ase Ts,es {v *{ .a / 1%'.*? g L g - - an, m m, + ma.s u m --,L ~ 'A 6.w* -- H I!'~Y**U w-I .\\ 1704 054 1 l 2 l 3 l 4 j 5 l 6

I 7 I 8 l 9 l 10 l 11 l 12 l 13 <ip ^ A +\\/N? 1 t l 9 10 p-31 ( _. g Vd ur s Mf Y PL AN w-a -- to rcem ,",q"'"

i. W.ar,s a.1 we= pir 4%04, b

,. m.an c,.....,.. c.;.. _.. .um,. r . m,..... s. MV~ % s n v.,s z., t%* 4.=aw smho es 7: * *. sinta.a%c. C Mfy I, -m -ass

6Ill w

7,t " y' =. ~"' ~ W l I-1 ng ~sy -. ~ ~ m 99 w. D 7 JL (*t w fi *(r 0 }e0 , % v'j /.._.." w "P' i ,/. = ~/m c hJ_e f , e n J2<.< a.... ' y, ___! I.._- , m- ~ \\ M*,'M F

rn,a.u.,

t.m.tg, a r> \\ L,. ~.. ~. s.~ n.g, u ,$ 4 _l.? -A- ~ ~ , f*= )f A sh f cr;.;._ ) _, g, j t

o

.p.,,, j H I =y.l-TX% .a e, r r,cr; r--, d-. j/ gl =. n4 p .a L = 5,1 ^ q* n-c, e .n : s.a j __.x ~ - =. .w ,,h i ._i_.."-**~ y _7.

3. 5n.....

- ~5*=

n-7 := m q,__. _ =__ u j-l o _ ', '. AL=.e,;x ~ (..~. _ m. \\ x- ' '.=~., a. a. lj r H 1704 05 3 / 8ALTIMORE GAS AND ELECTRIC COMPANY CALVERT CLIFFS NUCLE AR POWER PL ANT S R%r AIGNT R.NTEW 0%ES fuNT %Rfet ft9(,N RFygn 4fxtaWf FutDtNr., h4N at Ft,40' FIGUlit 7 I 7 i 8 l 9 l 10 4 11 l 12 13

Appendix C to Enclosure 1 CONCEPTUAL DESIGN FOR POST LOCA CONTAIINENT AIR SAMPLING CALVERT CLIFFS UNITS 1 & 2 BALTIMORE GAS & ELECTRIC CO. CONTENTS: I. INTRODUCTION II. DESIGN III. COMPOITE?."rS IV. INTERIM M*E URES FIGURE 1. P&ID POST LOCA CONTAINMENT ATMOSPHERE ANALYZER / SAMPLER 1704 056 ,..~e-3 7,

Page 2 I. Introduction In response to section 2.1.8.a of NUREG 0578, a design review has been performed for the existing containment radiation monitoring system to take and analyze an air sample from the containment af ter a LOCA in which significant core damage and large radiation fields result. The containment radiation monitoring system installed at Calvert Cliffs includes equipment for air particulate monitoring, charcoal filter analyzing, and containment air gaseous monitoring. The original design of the equipment did not include system operation af ter a design basis accident. Moreover, the design temperature, pressure, and activity ranges of the monitoring system are not compatible with the postulated post LOCA conditions. The system consists of a single containment pene-tration with automatic isolation valves radiation monitoring and sampling cabinets, and a blower which discharges sa= pled gases to the auxiliary building exhaust systems. The cabinets are located on elevation 5'-0" of the auxiliary building in the area where large radiation field is postu-lated by the plant high radiation shielding design review (103 - 104 R/ hr). These dose rates are based on the source term requirements of NUREG 0578 items 2.1.6.b. In consideration of the above design features of the existing system and NUREG 0578 requirements, it has been determined that the present system can not be used during or af ter the postulated LOCA. A new system, to be designed specifically for LOCA conditions, is therefore presented here which can be located in an accessible area and which can provide remote operation and grab sample capability. The position stated in NUREG 0578 section 2.1.8.a indicates that prompt quantification of accident related radionuclides would require the new system to have the capability to mea-sure both gaseous and particulate activities under all possible contain-ment accident conditions, including design pressure and te=perature. In addition, Regulatory Guide 1.3 or 1.4 source terms must be used in the design to predict the best possible plant arrangement and shielding con-figurations for achieving the lowest practical dose to operating personnel. II. Design A P&ID of the conceptual new containment air analyzing system is shown in Figure 1 The existing containment penetration piping, sample piping and isolation valves will be utilized. The connection to the new system equip-ment is located downstream of containment isolation valve CV-5292 (eleva-tion 33 '-0"). The new system equipment will be installed in an area with acceptable dose rates - either in the service water pump room or in the switchgear room on elevation 27'-0" or 45'-0". However the exact location of the analyzers will depend on size and weight of the equipment. The new system will feature a return path to the containment as release to the aux-iliary building ventilation system is undesirable due to the postulamd noble gas concentration. A positive displacement air pump will be used to ensure a timely purge and proper air flow to the analyzers. A spare con-tainment penetration will be used to return the purge air tc the contain-ment. The existing system, used only during normal operation, will be 1704 8 [7

Page 3 isolated by SV-4 when the new system is placed into operation to prevent the release of contaminants in the auxiliary building. Also included in the design will be the capability for obtaining a con-tainment air grab sample. Design details of the grab sample are not yet final as this will depend on the design of the commercial analyzing module. Access to the module itself will. only be required for taking grab samples. Sufficient shielding will be incorporated into the design of the analyzer module as well as the sample container to ensure as low as practical dose to the operators. All analyzer operations will be controlled remotely. Remote indication of gas and paticulate activities will be provided in the main control room with continuous recording on the remote control panel. Proposed design data for the containment air analyzing system is as follows: Temperature 40-300 F Relative Humidity 0-100*/. Sample Air Flow 25 CFM Pressure -2 to 60 PSIG Gaseous Activity Range 10 10 ci/cm x,133 5 3 3 Particulate Activity Range 10 105 ci/cm Safety Related No AC Power Supply Diesels III. Components a. Analyzer Module The analyzer module will house analyzer detectors which will have the cap-ability to perform radioisotopic identification of the containment atmosphere accident related constituents. It is proposed that an intrinsic germanium detector be used for this purpose. The detector is a liquid nitrogen cooled device which has excellent gamma ray spectrum resolution. It would not pro-vide identification of isotopes that are beta emitters. Beta emitters would have to be estimated from a comparison of the measured gamma emitter concen-trations and the typical fission product spectrum. The detector will have to be equipped with a collimator rather than solid shield in order to resolve the weak gn=ma from Xe133 A stainless steel collimator may be used solhat the hole does not corrode shut.

Page h If the commercial analyzer module is not equipped with an integral sample pump, then a pump vill be located in some other area to reduce the radia-tion does during grab-sample operations. The analyzer module vill be designed to provide more than one remote indication of the containment atmosphere activity. In the October 30 NRC Short Term Lessons Learned clarification letter, there is an implication that future requirements for remote indication in the proposed technical support center (TSC) may be imposed. Tharefore, the design of the containment air sampling system vill be finalized with this as a consideration. b. Remote Control Panel The analyzing system vill be designed to be operated and monitored remotely. The control panel vill be located in the existing reactor coolant sample room as this room has been shown to be accessible after a LOCA and vill be used for monitoring the status of other remotely operated plant sampling systems (post accident and normal). IV. Interin Measures Until the above modifications can be installed, a method has been established for the obtaining of a containment air sample. This will be acco=plished by routing a sample of the contairJ3ent atmosphere to a E2 gas analyzer. Due to the high dose rate, this analyzer vill have remote indications. This sample vill then be routed to the main vent where it vill be diluted and subsequently sampled for activity. Details concerning this sampling have been described in RCP-1-503. 1704 059

,m-x m t " m-JtiUL

z.,- 8 i j 2 t' Stas CLOSIS Si\\

5]j?} r/-tL C L o n_ _ sau +2 wi7u ov2 pyin

1yg 2.,,.

it_d/ m. '- c,. y ,e P'J%I AIR s x2 529r 6 2XaOAST bucT '

h) g 3

y 2-7!,i N c_ O N-I-M>; 1._ Sy 6277 }) V' N v ' E$

/- t L 0b3 :1

-? c; E, ta y]E CC r tTA:NMDiT Q_ "S '* M-g. /-T M C $DI NI

c. S

~s 2 >.3' # ' gg-5 gN ,, g ANAL'(2 5R MODO L"E ' h5 l 'N l (F __'/ 1 ' gg 3a s I l .sf I

r I

r' as sT ,] l v" gE / n )o lOJ A C04 a e

, c hS l

} e r~ tJ l g } l $ O'i O DI C E* 5' l l l %2 AHALY2_i; L F l l L_ 3 i l 43 l ,.3n L - _ _ _ _ _ __. _ _l I C,6 %.,,;> c) o l 1 $5 S A:iPLE l [ 43 1 5 -2 i l l c l I c-k/ O 5 l V'w

s L _ _ _

_] k s,,,...n< !! :i g c.

f.

e' p-04 vtpa sv.3 5' AS Clos 2s s v.2 J4 P 7/iT I O'/r%:bi j -2 S:1- $ C L O 7 6-h -c - - p c-2 wi-(p osig A ; 9 2 x )g c+_- T i" s s oa ~ ,q j r r ~

g' v.

~ 1704 060 i 1 s n,tu v u m,.,i o n s n.c.. . C l4 G. --~.- - -,, -.-.. ~.-

cu s s om o Dw. a.j v. ~ =- A i ~g g' l' / y ? S/ (?J E W Vt. VE) '-J 4 s -2 r r, A..r.,. g k.* ..j n O 4 y 3 y' N 6...: - s.w~ cs 110:1 tTO 'sS S'/- l 9,i V ~ r.x is rm sys, m N Z f/ S'/ 5 T i M I . rs o T) A H._ R21ST C.O N i rs O L. l ubi c ATio N SwncH35 J _. l _ _ _ - -,- - - _ _ _.I 1 n I a I 1704 061 up

,, fm

.. ! a, r f. ~. s.1.. :.. L. U. r. p,. !._..1n 1.1.. W. n.... '. t ..l(.. F,.4. E.,...,..,T... .i - C ALYZ ; c C.I.'. e L" :! c i ;o.1 cc 2 L3.,.. ....,u,- ._.r l C ' ; (!..'!'. 3:!C. M.' C.'L T O /. .a. "p e, d i L &, - %r . -~ v C. O r ~. / ! r 4 N E N 'T A T t-) O G % 'I R E l pa. n a simr car.t > n f m. --.. ...l ' 4.. b a as..a 4. s us e 1. rv1 : s. I l d,6 s FIG o rA's I _,. _...- ~, - -... . ~,. ,. n n. ~, -

APPENDIX D TO ENCLOSURE 1 CQiCEPTUAL DFEIGN FOR POST LOCA LIQUID SAFTLING CALVERT CLIFFS UIETS 1 AND 2 BALTIMORE GAS AND ELECTRIC COSTANY CONTENTS I. INTRODUCTIGi II. DESIGN CRITERIA III. PROPOSED MODIFICATIGI IV. SYSTEM COGONEtifS V. INTERIM FddURES FIGURE 1 P&ID POST LOCA LIQUID SAhTLE SYSTN 1704 062

Page 2 I. INTRODUCTI M The TMI-2 Lessons Learned Report cf the NRC, NUR m 0578, Section 2.1.8, requires the review of the plant's capability to obtain and analyze samples of the reactor coolant. Sampling capabiH ty is to be following a LOCA where 100% of the core is damaged and the following portions of the core inventory are released into the reactor coolant; 100% of the noble gases, 50% of the halogens, and 1% of the additional core constituents. An extensive redesign of the existing system would ling. This is due to the high be required for its use in post LOCA sq0 Rem /Hr) that will be generated dose rates (roughly estimated: 103 to 1 by the reactor coolant as determined in the plant shielding design review per section 2.1.6.b. Post LOCA sampling will require a large amount of shielding, mmg sample taking cumbersome and impractical for nomal plant sampling. Therefore, the existing system should remain as is and be used for normal plant sampling, and a new system should be installed for post LOCA sampling. The following is the planned design for a post LOCA liquid sampling station. The design is conceptual and requires additional analysis to finalize the design. The values given are an order of magnitude estimate. II. DESIGN CRITERIA The new sa=pling system will be designed to satisfy the following criteria: 1. The maximum permissible dose to personnel taking the sample will not be exceeded (3 Rems whole body dosage; 18-3/h Rems, extremity dosage). 2. In a timely manner (less than 1 hour), have the ability to obtain and analyze the reactor coolant. 3 obtain a sample of reactor coolant fram the primary loop and from the containment sump. 4. Sampling will not cause additional contamination of the auxiliary building, or centamination will be as low as possible. 5 Minimum numbers of camponents to decrease the chances of a system malfunction. 1704 063

Page 3 III. PROPOSED MODIFICATION The conceptual design is shown on Figure 1. The system will allow a sample to be taken, and analyzed, from the pressurizer liquid region, or the prtmary loop. In order to satisfy requirements for obtaining an analysis of the reactor coolant within one hour after an accident, in-line samplers will be installed. These will provide rapid on-site analysis of the reactor coolant. As a back-up to the in-line samplers and to provide the means for a more precise analysis, provisions will be made for taking a grab sample. The grab sample will be within a shielded cask which can be shipped off-site to a facility capable of performing the analysis. Figure 1 and the description below is for Unit 2 However, the design for Unit I will be identical, with the exception that the in-line samplers and the grab sample station will be located in the Coolant Waste Evaporator Room. For sa=pling from the primary loop or from the pressurizer liquid region, the existing sample piping within the containment can be used. The post-LOCA sample line will connect to the existing sam-ple line where the existing line exits the containment. A sample cooler will be used in the post-LOCA sample line to bring the fluid temperature down to that required for sampling. Up to this point, the equipment and piping will be located in the West Piping Pene-tration Room (el. 27' - 0") from which, the piping will be routed up to the. Solid Waste Handling Room (el. 45 - 0"). Here, the majority of the sampling equipment will be located. The fluid passes through a restricting orifice (RO-1) for pressure reduction. Sufficient piping will be provided, for taps and returns, for the sampling devices. The line will then be routed back to the West Piping Penetration Room, then into the containment, and finally into the containment sump. A flow will be established in this line to insure that a representative sample is taken. In the following discussion, this line will be referred to as the main sample line. Coming of f of and then going back into the main sample line will be a loop with a restricting orifice (RO-2), and a flow measuring device. This is to insure that the flow to the sampling equipment is at the proper rate. The in-line samplers will be connected down stream of R0-2. The samplers shall be connected in parallel, with their discharge directed back into the main sample line. Also connected to the loop downstream of R0-2 vill be the grab sample station. A shielded sample container will be used to take a sample. The container can be removed and shipped off-site for analysis. 1704 064

Page k IV. SYSTEM COVSONENTS The following are the major camponents required for the system (per unit). 1. In-line analyzers In line analyzing systems will be used for analysis of the reactor coolant. The individual analyzer elements will be grouped together in a shielded campartment which will be within the Solid Waste Handling Room. The individual analyzer electronics and remote indications / controls will be installed in an area protected from high radiation fields. The following is a list of the analyzers to be used. a. A highly collimated intrinsic gemanium detector for identification of radionuclides. b. A ther=al conductivity detector for identification of dissolved hydrogen. c. An electro-chemical cell for identification of dissolved cxygen. d. A boronometer for identification of boron. e. A pH electrode for pH measurement. f. A conductivity cell for measurement of conductivity. g. A chloride specific ion electrode for chloride identification. 1704 065

Page 5 2. Grab Sanple Station The grab sample station will consist of the following. a. Sample container The sample container will be within a lead cask. The cask will be in two parts (i.e., a cask within a cask). It would not be practical for the entire amount of lead shielding required for shipment to be in one cask. To reduce the dose from a one liter sample cawn to a 5 Rem dose one foot away, 6 inches of lead shielding would be required. A cask with this amount of shielding would be approximately: 15 inches in diameter, 24 inches high, and weight 1600 lbs. After a sample is taken, the cask can be moved to an area accessible by the existing monorail located in the plant area. The monorail will then be used to place the inner cask in the outer one. The sample will then be ready for shipment off-site. b. Sample enclosure The cask and fill line will be within an enclosure. The main purpose of the enclosure will be to prevent sample gases from contaminating the auxiliary build-ing atmosphere. It will also provide some additional shielding for when the cask cover is off. An exhaust duct will be connected to the enclosure to exhaust the gases. The enclosure will also be equipped with a viewing glass in order for the sample container to be observed during filling. 3. Control Panel A control panel will be installed locally to control all power actuated system valves, to provide system indica-tion, and to contain all pump controls. All system manual valves can be adjusted properly with the lowest practical does to the operator. The power supply will be from the plant AC system with diesel backup. 1704 066

Page 6 h. Restricting Orifices Restricting orifices, 'RO-1, and RO-2, will be sized to deliver the required flow against the maximum differential pressure that can be expected. In the event that the pressure differential decreases significantly below what the RO's have been sized for, the orifice bypass valves can be opened to decrease the resistance to flow. 5 Sa=ple Cooler The sa=ple cooler will be designed to cool the fluid from its maxinum te=perature down to that required for the in-line samplers and to maintain the grab sa=ple in the liquid phase. V. IITfERIM MEASURES Until all cf the above modifications are co=plete, the plarmed method of post, accident sa=pling of the reactor coolant is as follows: A shield wall has been erected near our present sink. A sample will be obtained by utilizing remote handling tools. The sample will then be routed to a holding tank where it can be diluted by re=ote control. After appropriate dilution, a sample shall be analyzed using app-repriate Rad Chem Procedures. Details concerning this sampling have been described in procedure RCP-1-503 " Post Accident Sampling of Reactor Coolant, Containment and Main Vent Systems". 1704 067

CONTAINMENT / 2c _ g ** P*4 2CV PENETRATION NOlA/ 54 s eNEW VRNt s ! 'w s 3g cw3a ,f %~CC-8)X3} s W e1 g s5 uoon u.a1 T -Y ruer:x. O ?/4"CC 2 con vN t' I e

l g

-~ ~~ ping] 4 SC [ t I W3T [ + S?ICC 3[CC / i N

- N m

(v2 2 D m 3 M 'P E PANEL W2 36 FC su a uma I

As c'.cs 2s

[3 [ 1 *0'"Lt - t 1:pp.E ROOM EL 45'-O" i I I ]1 7 I lh I v GATE g } $. Ih 5(CC =E OSIAs ctosES

2 WITu CvE ARtDa kk k

M 3 "CG B-2031 + c" SINGE / GST PLPING _ 4 2g MCoouNC, WATER ( li PENETMTd N P@M UtilT ?. l- - 3l ) Coolltro WATER) g p] l sxio - ~. _19 4 I T { PT -4 h e .I lg F -)Nif 2 IN- - ' FC 1 LINE SNiPLlNG r------ gj EST N p g X i g i UmT 2 EL.U-I1 l I SFA D RAD 10NUCLlDE I PENE $5 l \\DENTIFICATION I m I sy,, 5 I l Ym Fc 1 -{><3- $EMFlCRTI0N-bX3-F-----I & t--2 "8 >I l l SAM I 4 I DISSOLVED Oz 5 M IDENTIFICATION Il1 l h i CAS l 11 y l~ dL -t*a-l an us "a j-1 i .g l pH CHLORIDES $ I l ) IM DEN 1A ON t E t.___________.__________ _i

i L C 2002 SAS CLO5ES a, 5:A5 Lc5t5 M vi;Tu ovsstalbE M SPEC vgg );3 oggggia c_imt-i les/ _ _ (( - Y * * '2?9 suso / TO UNIT ?_ < M, . pg s/ ECCS PUMP \\ ROOM SUMP ut4l T 2 -%-2W2 h patsss.am rc.a gy; 2-2320 ~ [i C ) i uc,o c,y,..exzt,T SUMP f{c]3 P D ET FATiON h0 6 'El K LEGEND kEOCl EA STUG PIPE s ED 2 SX ' T ST TEW PtPE 5' CONN M/= M/JL%L VAWE WITH EEACH = 'tg2 ; / ROD u IAS CLOSES r-51AS CLOSES FC b L -1 7 _ _.L q ECCS PUMP "'T FC Asv UNIT 2 W <y CONTAINMENT i g M o_ __ J SUMP SAMPLE PUMP i E-SS TUSNG CONTA1 E T SUMP o [) ,E D ~4% MSN RATION /N 4 9 FC ]h 6'j d a AL 1704 069 mo m' Q ?LE ENCLOSURE D g\\g BALTIMORE GAS & ELECTRIC CO. ELECTRIC PRODUCTION PLANT f UNIT ?- CALVERT CLIFFS UNIT No.1 & 2 i LOCA GRAB ) 4PLE STATION BECHTEL POWER CORPORATION. j{ GAITHERSBURG, MARYLAND h -I /,} UNIT 2 POST LOCA LIQUID SAMPLE STATION... ~ SEC. TION ?.. l.8. Q-

HAUST 00cT-g ag.as, g=%

i .ll865 FIGURE ~1 J ~ -- s L}}

ML19260C721