ML20004C487

Responds to Generic Ltr 81-10 Re Emergency Response Facilities.Facilities Will Be Completed Prior to Fuel Load & Final Documentation Will Be Provided in Future Revisions to FSAR

ML20004C487
Person / Time
Site:	Wolf Creek, Callaway
Issue date:	06/01/1981
From:	Petrick N STANDARDIZED NUCLEAR UNIT POWER PLANT SYSTEM
To:	Harold Denton Office of Nuclear Reactor Regulation
References
SLNRC-81-38, NUDOCS 8106040160
Download: ML20004C487 (18)

Text

{{#Wiki_filter:_ 09# r SNUPPS Stenderdised Neeleet Unst Power Mont System Nicholas A. Petrick s choke Cherry Ross v tend 20060 Executive Director ' June 1, 1981 SLNRC 81-38 FILE: 0278 SUBJ: Emergency Response Facilities /Mr. Harold R. Denton, Director Office of Nuclear Reactor Reijalation jp '. /N U. S. Nuclear Regulatory Comission [v 8g i., (j$ Washington, D. C. 20555 W,' g J [1 Docket Nos. STN 50-482 STN 50-483, and STN 50-486 bl C ' e. % /gg F;

Reference:

NRC Generic Letter 81-10, dated February 18, 1981 N; '#$/4g, ,fj'

Dear Mr. Denton:

' ; 3.m.Tf.X' / The referenced letter provided recommendations for an Emergency Response Facility. The attached report describes the plans for these facilities for the Callaway and Wolf Creek plants. The facilities will be completed prior to fuel load and final documentation will be provided in future revisions to the SNUPPS FSAR and the Callaway and Wolf Creek FSAR Site Addenda. Very truly yours, l fr%~OC v / Nicnolas A. Petrick RLS/mtk Enclosure cc: J. K. Bryan UE G. L. Koester KGE D. T. McPhee KCPL W. Hansen USNRC/ CAL Q003 P T2 E. Vandel USNRC/WC II F .1.....> e

o c 1. ~ EMERGENCY RESPONSE FACILITIES e This report describes. the Emergency Response Facilities of the Callaway and Wolf Creek plants.s The f acilities for the two sites interf ace with the standardized SNUPPS power block and have many features in common. Aspects that are different between the 'two sites are specifically identified. ETechnical Support Center The Technical Support Center (TSC)lat each site is located within the pro-tected area and immediately adjacent-to the on-site building that contains the offices of the Plant Superintendent and the plant support staff. This building is termed the Service Cuilding at the Callaway site and the Admin-istration Building at the Wolf Creek Site. The locations of these buildings and 'the Technical Support Centers are shown. in Figures I and 2. ,This location for the TSC was selected because there is no suitable space within the power block ar.d because: - This location f acilitates activation of the TSC, since the persons designated to man the TSC have their offices in the Service or Admin-istration building. - There is ready accessibility to plant data available in the Service or Administration building which is not stored in the TSC (e.g., vendor manuals). - During normal plant operations the TSC will be readily available to the onsite engineering staff, which is quartered in the Service or Admin-istration Building. Distances from the TSC to the control room are approximately 700 and 450 feet, respectively, for Callaway and Wolf Creek. Walking times are estimated to be about three minutes and about two minutes, respective,y. At Callaway, travel between the TSC and Control Roan in the event of an accident would utilize motor vehicles (in part to minimize personnel exposure to radiation) and under these conditions the travel time should be less than two minutes. The TSC, shown in Figure 3, is a one story building of 5,000 square feet lo-cated at grade level. The walls are reinforced concrete 10 inches thick and the roof is reinforced concrete 6 inches thick. The structural design is in conformance with the Uniform Building Code. Within the ISC are 2900 square . feet of working space-that contain displays of plant status, meeting and dis-cussion areas, communications facilities, and document storage. The remain-ing 2100 square feet within the TSC are occupied by a mechanical equipment room, which centains HVAC equipment, a dedicated diesel-generator for the TSC, a _ computer room, and limited toilet, kitchen and access facilities. This is sufficient space for at least 25 persons, including five NRC personnel. For any extended duration of TSC operation, additional toilet, locker room, and . kitchen facilities in the Service or Administration Building will be avail-able.

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ra [. Page Two The HfAC systein for the TSC suppites outside air appropriately cooled or heated and has provisions to isolate inlet air and to operate in a filtered recirculation mode if radiation levels are high. The filter train contains l HEPA and charcoal-filters. Switchover to the filtered recirculation nsode is manual. -Radiation monitoring'in the TSC consists of cne area radiation monitor in 4 monitor which will detect concentrations as low as 10~g a radiciodine the main' work area, with a range of 1 to,10 mr/hr, an A.c/cc. The' radio- -iodine monitor will read the concentration in the inlet air to the TSC in wnichever mode of the HVAC system, open or recirculating, is employed. Electric power.to the TSC in a post-accident situation is normally provided by a transformer from off-site power. In addition, there is a dedicated standb,y diesel generator, rated at 218 kva, tnat is started manually utiliz-ing dedicated battery power. The diesel generator has sufficient capacity to power all TSC loads, including ~ the computer system, the communications system, H7AC ~and lighting. Essential equipment in tne TSC is also provided with power supplies to keep the equipment operable during a power interrup-tion, as for example, loss of offsite power af ter activation of the TSC and until the standby. diesel generator is started and asstsnes load. The computer system and communications systems have uninterruptable power supplies ('JPfi. Emergency lighting consisting of.tif-contained battery units is also pro-vided in the TSC. Protective clothing, breathing apparatus, and personnel radiation.~Jnitors to permit up to 10 persons to function within radiation areas will be stored within the TSC. Documents stored in the TSC will include the FSAR, systems descriptions and P&I0's, electrical schematics, general arrangement drawings and other appro-priate _ plant records. The conditions for manning the TSC are descrioed in general terms in the Emergency Plans for the Callaway and Wolf Creek sites. These are antained in Section 13.3 8 of the Callaway FSAR Site Addendum and Reference Document Section of the Wolf Creek FSAR Site Addendum, respectively. Detailed proce-dures are in the process of being developad, as Emergency P:an implementing procedures. Operations Support Center The location designated for the post-accident Operations Support Center (OSC) at Callaway is the Service Buiding and at Wolf Creek is the lccker room area of the Shop Building. These locations provide ample space for assemoly of personnel and have communications to the Control Room and TsC. Maintenance equipment, tools and protective clothing are also available. There is no special radiation protection because the OSC would not be utilized at times when outside radiation levels are high. k

Pag? Three Emergency Operations Facility The Emergency Operations Facilities (EOF) at Callaway and Wolf Creek are dif-ferent in layout and location with respect to the plants,--but the information systems are common'in design. At-Calla.vay the EOF is located approximately 1 mile from the plant, as shown in Figure 4. The building layout.is_shown in Figure S. It is a one-story building of 13,000 square feet, which is divided rougnty equally into (1) a Recovery Center and supporting services and (2) an Emergency Control Center and supporting services. At Wolf Creek the EOF is located approximately 2 3/4 miles from the. plant, as shown in Ftgure 6. The building layout is shown in Figure 7. It is a one- ~ story builoing of 14,000 square feet, which serves as a Visitors and Training Center during normal conditions and as are E0F in' emergency conditions. In each case the EOF working space is sufficient for at least 35 persons, con-sisting of 25 persons design?tw by the licensee including state and local officials,-9 persons from 'ae NRC and one person from FEMA. At each site the structural design of the EV is in conformance to the Uniform Building Code. Walls are concrete, approximately 10 inches thick and the roof consists cf double-T pre-cast concrete sections with a minimum concrete thickness of approx-imately 6 inches. The structure provides radiation shielding equivalent to a protection f actor of about 25. The HVAC system for the EOF is similar to that of the TSC, except it contains only HEPA and no charcoal filters. Radiation monitoring in the EOF is the same as described for the TSC. Electric power for the EOF is normally provided by a transformer from offsite power. In addition there is a dedicated standby diesel-generator, the rating of which has not yet been established, to operate the EOF in the event of loss of offsite power. The standby diesel generator is started manually, utilizing dedicated standby power. As in the TSC, the computer and ccmmunications systems are each provided with a Uninterruptible Power Supply. Emergency lighting con-sisting of self-contained t,attery units is also provided in the E0F. Because of the radiation protection factors provided by the building structures, it is the judgement of Union Electric and Kansas Gas and Electric that a backup E0F, as described by NUREG-0696, is not necessary. Emergency Response Facilities Information System (ERFIS) The Emergency Response Facilities Information System (ERFIS) conceptually consists of the: 1. The plant data acquisition system which supplies data to the: a. Control room CRT's b. Safety Parameter Display System in the control room c. Control rcom printer d. ERFIS communications processor

1. IPage Four. 1 2 '. ERFIS communicatlons processor wh'ich suppl'ies data to th'e: a., TSC subsystem -b.' ' Nuclear: Data Link (if required) '3.. TSC-subsystem which supplies data to the: . a.L - TSC peripheral equipment b. SPDS in the TSC ~ c. EOF peripheral equipment d. SPDS in the E0F A.- Safety Parameter Display System (SPDS) The Safety Parameter Pisplay System.is being designed jointly by'a group . of Westinghouse NSSS utilities of which SNUPPS is a member. - This con-cc.'tual dcsign i.s ' presented in Attachment'! to this submittal. - i -The control room SPDS is being designed for a 99% availability, and will not be seismically qualified. The function.of the SPDS does not warrant seismic qualification because of the low prol ability of a seismic event concurrent with the need for the SPDS. function,~ given tha availability of seismically' qualified dis-plays for key safety parameters in the control room. Further, a separ- ' ate additional concentrated display is not required as' a. backup for a nonseismic-SPDS and is conceptually contrary to good human engineering practices. -Qualified indicators are available and with proper training.of the opera- -tors are adequate for controlling the plant under all conditions. 'The SNUPPS Control Room has been subjected to a human f actors review and improvements are being made. The requirement to install separate addi-tional seismic displays compounds the human f actors problem and is in ~ conflict with the design criteria of Reg. Guide 1.97 which encourages that the operator use normal operating displays during accidents. This use of existing displays is most desirable since the operator will always get information to perform critical and normal operating functions from i the same location. The SPDS, by definition is intended to concentrate a P minimum set of plaat parameters' to aid the operator in the rapid detection of aonormal operating events. However, it is reasonable to use the normal qualified displays as a backup for this purpose. NUREG-0659 encourages that "....desig i improvements which can be accom-t plished by techniques of enhancement...be implemented prior to the comple-i tion of the control room design review...". Enhancement techniques are 1 4 those which can be most easily accomplished. The ease of implementation ' does not equate to benefit or human factors improvement. Such enhance-ments could have significant impact on the operators performance and the control room review. It is recommended that enhancements be implemented during t.;e course of the control room review only if it is determined that ? + a L i y r --u- =, ea-- r r r--- -p--

'Page Five ~ this rapid. implementation will produce a significant improvement and can be accomplished without a negative impact on operator performance. The existing seismically qualified instrumertation in a well human engi-neered. control. room can fulfill the functions required for the SPOS backup. -The ongolng control room reviews to improve human factors considerations will assure that the requested' functions for an SPDS backup will be satis-fled by qualified safety parameter displays. A separate concentrated seismically qualified backup SPOS in the control room is unnecessary. B. Nuclear Data Link The Nuclear Data Link has no definition at this time since the protocol and definition required has not been established by the NRC. The develop-ment of the required software and hardware for the other portions of ERFIS to_ meet NRC requirements prior to fuel load may make implementation of the -Nuclear Data Link very difficult. A complete system regeneration along with software changes will most probably be required. Further, the ERFIS hardware may be inadequate. C. Technical Suoport Center Sy' stem (TSC3) The TSCS conceptually consists of a CPU, CRT's and a high speed printer. SPOS displays along with the displays currently available in the control-room will be available in the TSC. The complete data base from the plant data acquisition system along wi'.h the data from the Radiation Release Information System and the Post Accident Sampling System will= be available. This data base includes the Regulatory Guide 1.97 parameters. The TSCS is being designed for a 99% availability. Details of Mstrumenta-tion quality, accuracy, and reliability have not yet been established. D. Emergencvjperations Facility System (EOFS) The EOFS conceptually consists of CRT's and low speed printers. SPOS displays along with the displays currently available in the control room will be available in the TSC. Current plans are to have the same data base at the TSC available at the EOF. The TSCS CPU drives the EOFS peripherals through a data link. As a result of the distance to the EOF, the data link has to be compatible with the environment. The E0FS is being designed for a 99% availability. Details of instru-mentation quality, accuracy, and reliability have not yet been established. , Task Functions for the TSC and E0F Both Union Electric and Kansas Gas & Electric have submitted their Radio-logical Emergency Response Plans to the NRC as part of the FSAR Site Addenda. Details of task functions and manning can be found in tnese plans. - l

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Attach.nent I ~ SAFETY ASSESSMEN1. ofEM GENERIC CONCEPTUAL DESIGN DESCRIPTION LO GENERAL CONSIDERATIONS The Safety Assessment System.(SAS) meets the requirements of the Safety Parameter Display System (SPDS). This report-describes that portion of the SAS which meets the SPDS requirements of NUREG-0696. It provides a centralized, flexible, computar-based data and display system to assist control room personnel in evaluating the safett status of the plant. This assistance is accomplished oy providing the operator ard other Emergency Response Facilities (ERFs) a high-level graphical display containing a minimum set of key plant parameters representative of the plant safety status. More detailed plant information is provided by several secondary displays. All graphical displays are presented to the control room operator on a bi;"'--- ' "':' multiple-color

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CRT. All data displayed by the SAS is validated by comparing redundant sensors,checkingthevalueagainstreasonablelimits,calculf'atingrates

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of change, and/or cticking temperature versus pressure curves. All displays of the SAS have been carefully designed by persons with plant operating experience and evaluated against human factors cesign criteria. The concepts used in the SAS design will be verified using data recorded from

Attachicant 1 Pagte Two ~ 6.similar power plant simulator. The intent of the.SAS is to present to the control room personnel a few easily understandable disp' lays wnicn use color co'ing and pattern recognition teschniques to indicate off-normal values. d These displays are updated and validated on an essentially real-time basis. Tne SAS will be operable during normal and abnormal plant operating conditions. The,SAS will operate during all SPOS requireo modes of plant operation. The normal operation mode will encompass all plant conditions at or above normal operating pressure and temperature. When the reactor coolant system is intentionally cooled below normal operating values, the operator will select the Heatup-Cooldown moce which alters the limit checking algorithm for the key parameters. An additional mode may be provided to address concerns ~ of cold shutdown plant conditions.

2. 0 DISPLAY-HARDWARE LOCATIONS AND OPERATION The SPDS portion of the SAS may be implemented on a single CRT located smys..%% ~L\\e~

tF ::ntr:1 ~m visible to th. ::ntrd r;:- ^^^-s** yY,. in a control location O' f end Uiw R iicr E ::ter Oper:t m This CRT contains the high-level display from i which the overall safety status of the plant may be assessed. A d;d

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r... a:, operator select W several predetermined second n;;'- _ater level (trend) displays at any time.

The SAS has ceen designed such that control room personnel can utili:e its features without recuiring adoitional coerations personnel. l

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Tne SAS displays ~will be provided to ;th;r ER": :::h :c the Technical racility.

Support Center and Emergency Operations

3. 0 DISPLAY CONTENTS The primary display consists of bar graphs of selected parameter values, digital status indicators for important safety system parameters and digital values. The parameters indicated by bar graphs and digital values include:

RCS pressure, RCS temperature, pressurizer level, steam generator levels and steam generator pressures. Status indicators are provided for centainnient i environment and secondary system radiation. Reactor vessel level N <a@644,coreexittemperature,amountofsubcoolingandcontainmentradiation are indicated by digital values. P J In addition, there is a message area which will be used to indicate that an appropriate secondary display provides further information in case an off-normal value is detected or an event is occurring. Each of the bar graphs indicates wide-range values. -If a parameter's value is outsice the normal range, the bar color will turn red. Arrows next to the car will indicate the trend direction (increasing or decreasing) based on data smoothing algorithms. During normal operation, the message area will be used to display L g average power, reactor core average temperature, daty.xtime. :-d ; '* ***+ ,g. __ These messages may be displaced by' higher priority messages as required. 1 9

f~ + .e e iAttachment I ~Page Four Seconcary.displ'ays may be selected by tne ooerator. Trend grapn -groups of selectec parameters, snowing tne last tnirty minutes of plant operation are availatie. 'These trend groupings were cnosen to keep like parameters or related parameters on one display "page".1 4 4.0 HUMAN FACT ^RS CONSIDERATIONS Human factors engineering and industrial design techniques have been_ effectively combined to establish man-machine interface design require-ments, maximize system effectiveness, reduce training and skill demands, and' minimize operator error. The CRT color graphic formats and functional key board designs have been developed through an interdiscipliancy team of senior operational, human factors, industrial design and computer interface personnel. Minimum use of color combined with simplified format throughout tne CRT presentation have key design features to provide both normal and off-normal pattern recognition. The operator, wno is the end user, has been directly involved from the conception to insure that man-machine interface goals of SAS have been satisfied. Human factor engineering standards and testing verification have been used which are consistent with accepted practices. 5.0 VALIDATION AND VERIFICATION The SAS is implemented on a digital computer system which_. includes a peripheral display generator computer for' color graphic displays. The l

I4 J - Attachment 1 Page Five software tnat controls the sensor data. validation, key parameter construction, and display formats has been developed under strict verification and validation procedures. ~ Tne original development of the SAS software began with a functional specification chat was developed over a period of 18 months by a technical com-mittee comprised.of members from a number of utilities and consultants. These functional specifications are transformed into a' design specification. Reviews of tne design specification will assure conformance of the SPOS portion of the SAS to those functions discussed in ilVREG-Cd96. The basis for selection of the primary display parameters will be a part of the final project documentation. During the course of software development, a set of static test cases will be developed wnicn test the key features of each software module. Further-more, static system test cases will be developed and used to verify the correct operability of tne total system. A set of dynamic test cases will be generated oy recording nuclear plant simulator data on magnetic tape from a number of - different plant transients which test the dynamic behavior of the system under "real" conditions. A design review that compares these test results to the original functional and design specifications will be performed. A selected numoer of the static test cases will be " frozen" such that they could be used to l verify future changes to the software. In summary, verification and validation was addressed and designed into the SAS software from the beginnin" tc provide a nignly reliable product and a mechanism for identifying and controlling future changes. 1 -}}