IR 05000395/1988006
| ML20153G513 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 05/03/1988 |
| From: | Decker T, Sartor W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20153G506 | List: |
| References | |
| 50-395-88-06, 50-395-88-6, NUDOCS 8805110255 | |
| Download: ML20153G513 (89) | |
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UNITED STATES
/ nato o NUCLEAR REGULATORY COMMISSION
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REGION 11
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101 MARIETTA STREET. N.W.
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tg 0 51988
Report No..
50-395/88-06 Licensee:
South Carolina Electic and Gas Company Columbia, SC 2921S Dcoket No.-
50-395 License No..
NPF-12 Facility Name:
V. C. Summer Inspection Conducted:
March 15-18, 1988 Inspector:
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6/2/68 W. M. Sartor, Jr.
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Dite' Signed Accompanying Personnel:
G. Bethke (Battelle)
G. F. Martin (Battelle)
K. G. McBride (Battelle)
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Approved by:
_T. R. Decker, Section Cnief Datt. Signed Emergency Preparedness Division of Radiation Safety and Safeguards Summary Scope:
This special, announced inspection was an Emergency Response Facilities (ERF) Appriasal.
Areas examined during the Appraisal included a review of st'ected procedures and representative records, the ERFs and related equipment, and interviews with licensee personnel.
Selected activities were observed during tne 1933 annual exercise to ascertain the adequacy of the ERFs and related equipment.
Results:
No violations or deviations were identified.
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k TABLE OF CONTENTS 1.0 Assessment of Radioactive Releases 1.1 Source Term 1.2 Dose Assessment 2.0 Technical Support Center 2.1 TSC Variable Availability 2.1.1 Documentation for Regulatory Guide (RG) 1.97 Variables 2.1.2 Regulatory Guide 1.97 Variable Availability and Sufficiency 2.1.3 Computer Data 2.1.4 Manual Data 2.1.5 Data Adequacy 2.2 TSC Functional Capabilities 2.2.1 TSC Power Supplies 2.2.2 TSC Data Analysis 2.3 TSC Habitability 2.4 TSC Data Collection, Storage, Analysis and Display
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2.4.1 Methods of Data Collection 2.4.2 Data Displays for the TSC 2.4.3 Time Resolution 2.4.4 Signal Isolation 2.4.5 Data Communications 2.4.6 Processing Capacities 2.4.7 Data Storage Capacity 2.4.8 Model and Systen Reliability and Validity 2.4.9 Reliability of Computer Systems 2.4.10 Envi-onmental Control Systems 2.5 Data Acquistions Systems 3.0 Emergency __Cperations Facility 3.1 EOF Variable Availability 3.2 E0F Location and Habitability 3.3 EOF Functional Capabilities 3.3.1 Data Analysis Adequacy 3.3.2 Backup EOF 3.3.3 EOF Reliability 3.3.4 EOF Data Collection, Storage, Analysis and Display 4.0 Persons Contacted 5.0 Exit Interview 6.0 Licensee Actions on Previous 1r Identified Findir33s 7.0 Glossary of Acrenvms and Initialisms
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1.0 Assessment of Radiological Releases 1.1 Source Term
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There were three primary gaseous effluent release points: the main
plant vent, reactor building purge exhaust, and atmospheric steam dumps.
The following potential release streams were routed through L
the main plant vent: waste gas decay tanks, condenser exhaust, fuel handling building exhaust, and auxiliary building exhaust. A review of the FSAR, Section-11.4, indicated that the primary gaseous release pathways were monitored and that the potential release pathways feeding 'to the main plant vent were also monitored.
The only unnonitored release pathway identified was leakage through the containment structure.
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Predetermined relationships for estimating core damage were contained in Chemistry Procedure CP-308, "Core Damage Assessment Methodology,"
Revision 2.
-The methodology for this procedure was based on the Westinghouse Owner'<. Group Post Accident Core Damage. Assessment Methodology report.
The primary method for performing dose assessment calculations in the EOF and TSC was the Emergency Assessment and Response System (EARS).
The backup method was contained in Emergency Plan Procedure EPP-005,
"Offsite Dose Calculations,"
Revision II, May 28, 1986.
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Procedure EPP-005 also contained the nomographs which were the primary method by which Control Room personnel performed dose calculations. A release source term could be calculated with EARS from the following inputs; post-accident sample analysis, grab
samples, plant vent monitor, containment purge monitor, main steam line monitors, containment leakage information, and FSAR based
default assumptions.
The manual method of Procedure EPP-005 determined a release source tern from the following inputs; plant vent monitor, high range containment monitor, steam line monitor,
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ccitainment leakage, grab sampie results, environmental monitoring I
data, and FSAR based defaults. The nomographs in EPP-005 used by the
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CR staff could calculate doses based on the following data inputs;
grab sample results, steam line monitors, plant vent monitor, and
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reactor bui' ding purge exhaust monitor.
Corporate Health Physics l
Procedure CHP-309, "Emergency Operations Facility Offsite Dose l
Assessment Guidance and Conduct of Operations," Revision 6 provided a method for EOF dose assessment personnel to calculate a source term from field monitoring data,
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j During the emergency planning exercise, an inspector observed that
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the dose assessment personnel in the EOF did not use the dose
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calculations from the steam line monitor readings to recommend l.
emergency classifications or protective action recommendations. An EARS calculation during the exerci>e using a steam line monitor reading of.5 mR/h during a release through the steam dumps yielded a whole-body dose rate of 11 R/h at the site boundary. Dose rates at
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the site bcundary reported by field teams were background. The EARS results were approximately 1E4 higher than field monitoring results.
The steam line monitors (3) were ion chambers with a range of
.1 to IE7 mR/h gamma.
Each monitor was shielded by 2 inches of lead to reduce background and provide collimatior The plant FSAR, Section 11.4.2, stated "The anticipated dose response, calculated for the detector due to an equivalent concentration of Xe-133 in the discharge effluent from the steam line is 6.1 X 101 mR/h/pCi/cc."
Discussions with licensee health physics personnel revealed that sensitivity calculations performed by a vendor indicated that even slight increases in steam monitor readings imply major core damage which translated into large site boundary doses.
This lack of sensitivity in the monitors (RMG-19a, b, c) makes them ineffective as indicators to classify events or provide input for protective action recommendations until PAGs have been exceeded.
Based upon the above review, the licensee committed to:
Review the specifications and documentation on the steam line monitors to determine if the monitors can be made ef.fective in providing dose assessment data at lower than PAG limits (IFI 50-395/SS-06-01).
1.2 Dose Assessment The manual dose assessment method t>ed a straight line gaussian plume model while EARS used a segmented gaussian plume model. Both methods used Regulatory Guide 1.23 assumptions for determining stability class from AT, Regulatory Guide 1.145 assumptions for atmospheric dispersion from ground level releases, and Regulatory Guide 1.109 assumptions for whole-body and child thyroid inhalation cose conversion.
In addition, EARS used Regulatory Guide 1.111 assumptions for dry deposition and simple exponential wasbout.
The EARS computer code was supplied and supported by a vendor.
No documentation supporting the validation and verification (V and V) of this code was readily available in the plant docum9ntation system.
Also some of the documentation describing methodolcgies used in EARS
and the manual methods were difficult to locate. Personnel felt that
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some portion of this documentation might exist onsite in personal l
files or might be available from the vendor.
A new plant procedure l
(Jan. 88) covering software V and V and maintenance was being applied to existing software in an attempt to gather and record appropriate
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documentation.
The EARS code had not been brought into compliance l
with this procedure at the time of this inspection.
Some poorly documented cases of calculational ccmparisons between EARS, State, and NRC dose rodels were found.
The comparisons indicated fair agreement within the limited cases presented.
No comprehensive and fully documented calculational comparison between l
EARS, back-up manual models, State models and NRC models existe.
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Based upon the above review, tne licensee committed to:
Locate all available documentation pertaining to validation, verificati:n, and methodology for all dose assessment models (computer and manual) and centralize the maintenance of it into the Plant Record System (PRS) (IFl 50-395/83-06-02).
Establisn a periodic calculational comparison between dose assessment models (e.g., EARS, EPP-005, State, NRC) and document the reasons for significant differences (IFI 50-395/88-06-03).
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2.0 Technical Supoort Center (TSC)
2.1 TSC Variable Availability 2.1.1 Documentation for Regulatory Guide (RG) 1.97 Variables South Carolina Electric and Gas (SCE&G) received an SER for Regulatory Guide 1.97 variables it. November 1937.
The November SER contained only three rcmaining questions concerning SCE&Gs implementation.
Two of the questions were related to the environmental qualification of Accumulator valves and Main Steam Line Monitors and the third question concerned a slightly more narrow range on the Containment Temperature Monitors at V. C. Summer than what is speci fied in RG 1.97 (50 to 250 degrees F vs 40 to 400 degrees F).
For the purpose of the Emergency Response Facility (ERF) Appraisal, adequate RG 1.97 variables were available.
This does not eliminate the need for SCE&G to resolve the emaining questions in the RG 1.97 SER with the cognizant NRC Division.
2.1.2 RG 1.97 Variable Availability & Sufficiency
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Plant variables were available to the TSC and the EOF via the Plant Safety Status Display /Onsite TSC (PSSD/GTSC) computer system.
The PSSD/0TSC was the original Westinghouse design for a combination SPDS and ERF data acquisition system. The inspector reviewed the computer system point list against RG 1.97 and found no missing variables.
In addition to the PSSD/OTSC system, the TSC had a functional analog instrument panel which provided a limited set of reactor coolant system and steam generator variables.
2.1.3 Corputer Data The PSSD/0TSC system was operated by licensee and inspection team personnel during tne a p p ra i.^.a l.
Most of the few computer points which were invalid or of poor quality had been removed f rom scan by the licensee, thus precluding operators and ERF managers from being mislead by erroneous data.
The PSSD/0TSC system had a trending capability which allowed plotting 4-hour historical trends of points L
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during normal operation and up to 24-hour trends in the post reactor trip environment.
SCE&G is cautioned that this ERF Appraisal report in no way influences the need to complete SPD3 requirements as outlined in the November 19S7 SER (e.g., The report on SPDS deficiencies requested by NRC from SCE&G three months prior to the next refueling outage).
2.1.4 Manual Data Dedicated telephone communicators and status boards were used in the TSC as a backup to the computerized data acquisition system. The TSC was located in the Control Room envelope and had a window which overlooked the Control Room.
Data could be passed by hand from the Control Room to the TSC.
TSC data could also be passed to the EOF via a telefax machine.
The inspector observed all of these manual techniques being used during the annual exercise which was performed concurrent with the ERF Appraisal.
2.1.5 Data Adequacy All RG 1.97 variables were available in the TSC and adequate.
Based upon the above findings, this portion of the licensee's program appeared adequate.
2.2 TSC Functional Capabilities 2.2.1 TSC Power Supplies All TSC lighting, receptacles, computer systems, printers, and other equipment were supplied with power from a 4SO/277 VAC distribution panel (APN 4021).
APN 4021 was supplied by a battery backed uninterruptable power supoly (UPS) which in turn could have been powered from either the A or B train 4S0 VAC engineered safeguards feature (ESF) buses.
Tnerefore, all TSC power was extremely reliable. Telephone systems in the TSC were of two major types; some were a R3LM system maintained by SCESG and the others were a SLIC 85 system maintained by Southern Bell.
Both categories of telephones were provided with backup power from battery / charger units.
The SCE&G microwave system was also battery packed.
2.2.2 TSC Data Analysis In addition to the data systems previously described, the TSC was supplied with a complete set of plant drawings, reference material, and other aids for data analysis.
Based upon the above findings, this portion of the licensee's program appeared adequat. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _,
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2.3 TSC Habitability The TSC was contained in the Control Room environmental envelope and as such was designed to maintain the same level of habitability as the Control Room. An analysis and discussion of radiation doses and exposure rates for the TSC/ Control Room was contained in the V. C. Summer Final Safety Analysis Report (FSAR) Sections 15.4 and 12A.4. The gamma radiation shielding provided appeared adequate for tne accident conditions discussed in the FSAR.
The TSC was serviced by the same ventilation system as the CR and consisted of redundant trains A and B which each contained a normal air handling unit and an emergency air handling unit. The emergency section of each train contained a series of high efficiency particulate air (HEPA) filters and charcoal adsorbers which were automatically placed in series with the normal section of the train upon detection of gaseous radioactive material in the supply header or upon a Pnase A Containment Isolation signal.
The inspector reviewed system drawings and test procedures and observed several system actuations. The HVAC system maintained approximately 2 inches of H O positive pressure in the Control Room envelope when operating
in the emergency mode.
Based upon the above findings, this portion of licensee's program appeared adequate.
2.4 TSC Data Collection, Storage, Analysis and Display At the time of this Appraisal, computer systems providing ERF support were being upgraded. The upgrade effort was in the conceptual phase and details of tne upgrade were not available and will not be discussed in this report.
Findings described in this section are based on the evaluation of existing computer systems and their use to support ERF functions.
2.4.1 Methods of Data Collection Real-time data acquisition, display, and storage to support ERF functions were performed by redundant Digital Equipment Company (DEC)
PDP 11-44 ninicomputers.
Each 11-44 had 1.75 Megabytes (PB) random access memory (RAM), a 67 M3 hard disk unit, and two shared 1600 bits per inch (BPI) magnetic tape drives.
The bulk of the ERF sof tware was written in FORTRAN 77 witn some routines written in assembly language.
Supporting documentation (e.g.,
a user's guide and a programmer's reference manual) were found to be comprehensive and professionally done.
Computers gathering ano transmitting data to the ll-4?s included: 3 DEC PDP 11-03s with 64 KB (kilobytes) RAM and no peripherals (these computers were front ends to the 11-44s and read plant sensors); and 1 Hewlett Packard 1000 linked to the ll-44s via a DEC PDP 11-23 with 12S KBs RAM and no peripherals (the HP 1000 gathered meteorological data).
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The V. C. Summer plant also had available a VAX cluster of four VAX computers ranging from the 7S5 to $530.
These computers did not directly support ERF functions; however, they were available for special purpose data analysis and display of ERF data if needed.
The following is a list of analog (continuously variable) and digital (2 state) plant sensors:
Analog Digital Sensors Se_nsors Ra d / t'_e t Total Sensors 250 aD
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2.4.2 Data Displays in t_be TSC Four Aydin 5216 display generators controlled ERF display Cathode Ray Tube's (CRTs).
In the TSC there were 5 Aydin display CRTs, 1 Tektronix hard copier, and I line printer.
Users ce"1d display safety parameters or parameter sets of interest. On request a hard copy could be generated.
Display options have been irplemented into three basic types:
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PSSD_(Plant Safety St a t u s D i s p1_a_y_1 The PS$D displays were designed to provide a quick overview of plant
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safety status, detect and evaluate abnormal conditions, and provide the ability to r.onitor and help ritigate abnormal events. One of the PSSD options was tne display of an octagon with the vertices displaying the expected values for S overview plant safety parareters. Wnen actual plant parameters did not natch the ideal the octagon shape would be distorted letting the user know immediately of off-standard cor.ditions.
Experienced users were familiar with most cistorted octagon possibilities and could relate a specific shape to a specific plant problen There were some la menu choices for PSSD displays, all of which were related functions.
OTSC (Onsite TSC)
OTSC displays have been designed for plant manager ent and technical suppm t personnel.
These displays have been designed to provide status of rajor systems, provide status of support systems, and provide history and trend information. The OTSC did not provice the craphic (iconic) display as did the PSSD but did provide some 22 menu choices for related plant systens, twelve of which were the same as the PSSD.
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BISI (Bypassed and Inoperable Status Indicator)
EISI provided the CR operators and ERF personnel with a concise and continuous indication of the status of the ESF systems and their support system.
BISI provided a means to evaluate the causes of alarm conditions using a three level hierarchical structure. The top level display, ESF systems, listed ESF systems sequentially along with their availability status. The second level showed the systems as subsystems and support systems.
The third level gave component status and position.
PSSD and 0TSC displays have been designed to: (1) give a top level summary of plant health; (2) show a second level graphic display of overall plant status; (3) give a third level of plant systems; and (4) provide alphanumeric format displays of sensor data.
Display functions were noted to be impaired by intermittent lockups.
Lockups were readily corrected from any available display CRT (re-bcoting computers was not required).
Apparently, licensee personnel have observed lockups but have been unable to correct the problem due to the difficulty of identifying the cause. Because the lockups are rare and can be easily dealt with and because of the availability of several CRTs in each ERF this is not viewed as a serious reliability problem.
Based upon the above findings, this portion of the licensee's program appeared adecuate.
2.4.3 Tire Resolution ERF supporting corputers read, analyced, and stored to hard disk data
' rom 735 analog and digital sensors.
The sampling rate for data acquisition varied be:ncen every two seconds and every 10 seconds for ERF related plant sensors. The rate for RAD data was 35 data points every 2 minutes and for PET data was 9 data points every 15 minutes.
The data rate was considered len to roderate speed.
The cata acquisition tasks were assigned a high priority and even when display tasks were observed to lockup, the system continued to collect and store plant sensor data without apparent data loss.
Based on the above findings, this portion cf the licensee's program appeared acequate.
2.4.4 Sicnal Isolation
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At the V. C. Su-er plant both optical and raonetic isolation were used to provide isolation sufficient to reet N'JREG-0737, Supplerent ;
requirerents.
This was verified by letter "Safety Evaluation Report
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by the Office of the Nuclear Regulatory Commission Regarding the SPDS South Carolina Electric and Gas Ccrpany, Virgil C.
Su~,er Nuclear Plant, Doctet No. 50-395."
Section Ii.5 of this report states "The
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S isolation for the SPDS at Summer meets the requirement of Supplement 1 to NUREG-0737."
Based upon the above findings, this portion of the licensee's program appeared adequate.
2.4.5 Data Co~munications Data communications capabilities were reviewed between the DEC PDP 11-44s and the other DEC processors. These included the 11-03 front ends, the 11-23 collecting RAD / MET data from the HP 1000, and the 11-24 at the EDF.
Error checking and correcting was reported to be done by modem firmware and operating system software.
DSta communications processors used high speed data links (approx'mately 56,000 bits /second).
Based upon the above findings, this portion of the licensee's program appeared adequate.
2.4.6 Processina Cagabilities The DEC PDP ll-44s and peripheral computer systems were configured to support plant safety monitoring and reporting needs. Processing was based on multitasking to allow several software functions to be processed concurrently. Data acquisition and storage tasks were high priority tasks and continued to execute even when other tasks were locked up (e.g.,
display tasks).
Licensee contacts reported the 11-44 central processing units to routinely function 75 to 89'o loaded. The heavier processing loading during an emergency situation would result in a slower respcnse tire. A much f aster response time will be available following the installation of a new computer system currently scheduled for calendar year 1990.
Eased upon the above findings, this portion of the licensee's program apoeared adequate.
2.4.7 Data Storace Caoacity Data storage capaclty met N'JREG-0696 requirements. Utility per:,onnel interviewed reported that at any time, 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of historical data was available to provide trending i n f o r.m a t i o n on critical plant para eters. On demand, plant analog sensor data could be stored up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> en disk. Also, on dem.nd, up to 200 digital sensor full sample set state changes cuuld be saved to disk.
If both PDP 11-44 computers should have to be re-bected because of haraware or software failures, historical data would be lost. Power failure will not cause this problem because the computers use autcmatic restart on power restore.
The automatic restart restored all files and cceputer states prior to the por,er failure.
The loss M
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g of historical data was considered very unlikely because of the redundant configuration and because of the availability of records.
Based upon the above findings, this portion of the licensee's program appeared adequate.
2.4,8 Model and System Reliability and Validity Documentation for model algorithms described' in the Programmer's Manual, written by Westinghouse Electric-Corporation on December 6, 19S5, were reviewed.
The equations and iescriptions were professionally documented.
M0 del algorithms we*e not checked for completeness and correctness during this Appraisal.
Based upon the above findings, this portion of thii licensee's program appeared adequate.
2.4.9 Reliability _of Computer Systems Computer system unavailability was reported by the utility for the ERF support computer systems to be.89% for the past year.
Based upon the above findings, this portion of the licensee's program appeared adequate.
2.4.10 Environmental Control Systems Air conditioning was report *d by licensee personnel to be functional in the computer room. The air conditioning system was reported to be set to maintain ambient temperature at about 80 degrees Fahrenheit.
Based upon the above findings, this portion of the licensee's program appeared adequate.
2.5 Data Acquisition Systems The licensee report on the implementation of RG 1.97 variables was available and was used as an appraisal information resource.
Conpared to similar electronic data systems reviewed, the PSSD/0TSC system appears to have a very high degree of reliability and a lack of spurious alarms and erroneour data.
The PS$D/0TSC has a very reliable power supply and did not exhibit
"lockups" during the course of the Appraisal. In the unlikely event the system would need to be "re-booted" during an accident, the present sof tware configuration causes all historical data on the computer hard disc remory to be erased.
The reason given by SCE&G for this feature is that the plotting routine cannot account for missing data points (which would cccur during the time of re-boot) by either f airing between pre-boot and post-boot data or by indicating
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missing data points with some special characters.
The licensee
agreed to consider eliminating this memory erasing feature during a future software modification.
Based upon the above findings, this portion of the licensee's program appeared adequate.
3.0 Emercency Operations Facility (EOF)
3.1 EOF Variable Availability
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The EOF used the same PSSD/0TSC systen for electronic data as that
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used in the TSC.
The telephone systems, status boards, telefax equipment, and esference material available in the EOF were essentially identical to that in the TSC.
Therefore, this section
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i Based upon the findings previously described, this portion of ti.e licensee's program appeared adequate.
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3.2 EOF Location and Habitability The primary EOF was located in the basement of the Nuclear Training l
Center (NTC) about 2.5 miles south of the plant power block. Ceiling and wall shielding provided a protection factor of greater than 5.
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i The E0F was served by a dedicated riVAC system consisting of one j
normal and one emergency air handling unit, associated condensers-
(chtilers), heaters and filters.
The emergency filter train j
consisted of roughing filters and two HEPA filter banks. No charcoal adsorbers were installed in the system.
The system was manually l
switched to the emergency mode upon activation of the EOF.
The inspector reviewed systen drawings and test procedures and observed ei several actuations of the system.
The system performed satisfactorily, with all fans and dampers actuating properly to i
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caintain a positive pressure in the EOF envelope.
NTC building maintenance records showed that the system was periodically tested
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The cognizant system engineer was in the process of having more formal j
test and maintenance precedures (e.g., Nuclear Training Center HEPA Filter Test) approved to insure continued reliability of the E0F HVAC
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Based upon the above findings, this portion of the licensee's program appeared adequate.
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i 3.3 E_OF Functional Capabilities
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3.3.1 Data Analysis Adequacy
'The electronic and manual data systems serving the E0F were identical to those previously described for the TSC, with the exception of the analog instrument panel which was installed in the TSC only.
Based upon the above findings, this portion of the licensee's program appeared adequate.
3.3.2 Backup EOF The backup EOF was located in Columbia, SC at the SCE&G Corporate offices.
The backup EOF was about 25 miles southeast of the plant site.
This location was previously approved by the NRC.
The inspector reviewed several electrical grid maps and determined that power for the backup EOF was supplied by a portion of the SCE&G grid which is not likely to be affected by any events which would disrupt power at the primary EOF.
The backup EOF received power from the NETWORK Substation which is connected to four different incoming 33 and 115 KV lines. The NETWORK substation is separated from the PARR Substation, which supplied the primary EOF, by at least 2 major substations on any possible connecting path.
The backup EOF used a SCE&G maintained Dimension telephone system which was battery backed.
Based upon the above findings, this ' portion of the licensee's program appeard adequate.
3.3.3
_ EOF Reliability The entire NTC was supplied with power from two redundant 23 KV sources.
Each of the incoming 23 KV lines had separate sources including 115 KV lines, a hydro station and gas turbines.
The incoming 23 KV lines were stepped down to 480 VAC through separate NTC transformers.
Either of the sources could be selected to power the building via manual breakers in the basement switchgear room.
Within the EOF, sensitive equipment such as computer terminals were supplied with conditioned power via a 15 KVA regulated power supply.
Based upon the above findings, this portion of the licensee's program appeared adequate.
3.3.4 EOF Data Collection, Storage, Analysis and Display _
The same computers supporting TSC ERF activitics supported the EOF.
These systems and details of their functions have already been described.
However, the licensee had implemented a DEC PDP 11-24 in the EOF for processing microwave communications from the PDP 11-44s at the plant.
The PDP 11-24 also controlled 1 Aydin 5216 display generator which in turn controlled 2 Aydin 8025 and 1 Aydin 8330
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l display CRTs. Also supported in the E0F were 1 Tektronix hard copier and 1 300 line per-minute line printer, EOF displays were exactly i
the same as the CR and TSC displays. Communications between the PDP 11-44s and the' PDP 11-24 have-been implemented using microwave equipment with a 56KB data link. Error checking and correction were reported in use at the EOF.
'The EOF used the same PSSD/0TSC system for electronic data as that used in the TSC.
The telephone systems, status boards, telefax equipment, and reference material available in the E0F was i
essentially identical to that in the TSC.
Therefore, this section
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contained no additional comments.
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Based upon the above findings, this portion of the licensee's program appeared adequate.
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4.0 Persons Contacted
- W. Baehr, Manager, Chemistry and Health Physics
- E. Baker, Systems Engineer G. Baker, HVAC Systems Engineer -
"R. Barton, Nuclear Computer Services Engineer
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- K. Beale, Manager, Nuclear Protr. ion Services l-
"R. Bender, Training Simulator Instructor
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- L. Blue, Manager, Corporate Health Physics
- 0. Bradham, Director, Nuclear Plant Operations
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- M. Browne, General Manager, Station Support
- R. Clary, Manager, Design Engineering C. Coleman, Telecommunications Specialist
- M. Counts Emergency Services Coordinator i
- H. Donnelly, Nuclear Licensing Senior Engineer
"J. Gesn, Engineer, Design Engineering i
"G. Higginbotham, Corporate Health Physics j
- A. Koon, Manager, Nuclear Licensing
- F. Leach, Manager, Facilities and Administration
- G. Liu, ISEG Engineer
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- G. Moffatt, Manager, Maintenance Services
- K. Nettles, General Manager, Nuclear Safety
- J. Proper, Associate Manage, QA
- R. Rusaw, Engineer, Systems Engineering
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- J. Skolds, General Manager, Station Operations
"G. Soult, Operations Manager
- S. Summer, Supervisor, Environnental Programs J. Wactor, Senior Electrical Engineer
- D, Warner, Manager, Core Engineer and Nuclear Com.puter Services
- B. Williams, Supervisor, Project Controls
"M. Williams, General Manager, Nuclear Services
"W Williams Jr., Special Assistant, Nuclear Operations Other licensee employees contacted included engineers, technicians, operators, and security force members,
.
.
.
Other Organizations J. Erster, Fellow Engineer, Westinghouse Electric Corporation Nuclear Regulatory Commission
- H. Dance, Section Chief, RII
- T. Decker, Chief Emergency Preparedness Section, RII
"J. Hayes, Project Engineer, NRR
- P. Hopkins, Resident Inspector, RII
- D. Prevatte, Senior Resident Inspector, RII
- Attended exit interview 5.0 Exit Interview The inspection scope and findings were summarized on March 18, 1988, with those persons indicated in Paragraph 4.0 above.
The inspector described the areas inspected and discussed in detail the inspection findings herein No dissenting comments were received.from the licensee.
Although proprietary material was reviewed during the inspection, such material was neither removed from the site nor entered into this report.
6.0 Licensee Action on Previously Identified Findings a.
(Closed) Inspector Followup Item (IFI) 50-395/87-03-01: Provide a method for effectively tra c ki ng, docu?entino, and posting repair and investigative reentry teams dispatched f rom the OSC.
An inspector observed that the licensee had established and used during the exercise an "In-Plant Entry Team Status Log" which corrected this previous finding.
b.
(Closed) IFI 50-395/87-03-02:
Assure that monitoring and surveillance procedures provided in field team kits are used in implementing field monitoring and surveillance requirements. An inspector noted that exercise observations did not idertify any procedural adherence problems with the field monitoring and surveillar e teams, c.
(Closed) IFT 50-395/87-EP-01:
Verify audibility of alarms in high noise area (79-BU-18). This item was previously closed by the resident inspector in Inspection Report 50-395/85-09, l
_ _ _ _ _ _ _ _ _
_ _.
-
.
I4 7.0 Glossary of Acronyms and Initialisms BISI Bypassed and Inoperable Status Indicator CHP Corporate Health Physics Procedure CR Control Room CRT Cathode Ray Tube DEC Digital Equipment Company EARS Emergency Assessment and Response System EOF Emergency Operations Facilities EPP Emergency Plan Procedure ERF Emergency Response Facilities
FSAR Final Safety Analysis Report HEPA High Efficiency Particulate Air Filters HP Hewlett Packard HVAC Heating, Ventilation and Air Conditioning IFI Inspector Followup Item (0 pen Item)
KB Kilobytes KV Kilo Volts KVA Kilo Volts Alternating MET Meteorology NRC Nuclear Regulatory Commission NTC Nuclear Training Center NUREG Nuclear Regulation l
OTSC Onsite Technical Support Center PAG Protective Action Guidelines PRS Plant Record System
,
PSSD Plant Safety Status Display RAD Radiation Absorbed Dose RAM Random Access Memory RG Regulatory Guide
!
SCE&G South Carolina Electric and Gas Company SER Safety Evaluation Report f
SPDS Safety Parameter Display System TSC Technical Support Center UPS Uninterruptable Power Supply VAC Volts Alternating Current V and V Validation and Verification Attachment
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VCSNS OVERVIEW l
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l DESIGN
!
J V. C. SUMMER NUCLEAR STATION HAS ONE OPERATING NUCLEAR UNIT
!
- WESTINGHOUSE PRESSURIZED WATER REACTOR
.
e 277S MWT(RX POWER)
'
e GILBERT / COMMONWEALTH WAS THE ARCHITECT ENGINEER l
COMMERCIAL OPERATION
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l e CONSTRUCTION PERMIT ISSUED IN MARCH,1973 e 5% POWER OPERATING LIMITATIONS MET IN AUGUST,1982
'
e 50% POWER OPERATING LIMITATIONS METIN NOVEMBER,1982
!
e 100% POWER OPERATING LIMITATIONS METIN MAY,1983
,
!
e UNIT 1 WENT COMMERCIAL ON JANUARY 1,1984 i
EMERGENCY PLANNING e INITIAL EMERGENCY EXERCISE ON MAY,1981
- PERMANENT EOF COMPLETE IN JANUARY,1983
i e NRC PARTICIPATION IN APRIL 1985 EMERGENCY EXERCISE e FINAL ACCEPTANCE OF SIREN SYSTEM SEPTEMBER,1986
-
_ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
.
ERF PRESENTATION
.
SPEAKER TITLE SUBJECT M. N. BROWNE GENERAL MANAGER, INTRODUCTION
$UPPORT SERVICES V.C. SUMMER NUCLEAR STATION OVERVIEW e
MAJOR MILESTONES AND AUDIT HISTORY A. R. KOON MANAGER, NUCLEAR LICENSING K. E. BEALE MANAGER, ERF OVERVIEW NUCLEAR PROTECTION SERVICES FUNCTIONAL ORGANIZATION e
REGULATORY REQUIREMENTS / GUIDELINES e
AND REVIEWS l
PHYSICAL FACILITIES e
EMERGENCY RESPONSE DATA SYSTEMS R. A. BARTON SUPERVISOR, PROCESS COMPUTER SYSTEMS
- GUIDANCE e DESIGN INFORMATION DISPLAY e
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_ _ _ _
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ERF PRESENTATION (continued)
!
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SUBJECT SPEAKER TITLE I
l DOSE ASSESSMENT G. E. HIGGINBOTHAM STAFF HEALTH PHYS!CIST
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EMERGENCY RESPONSE ORGANIZATION K. E. BEALE MANAGER, NUCLEAR PROTECTION
'
I SERVICES TSC
.
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ERF DOCUMENTATION OVERVIEW H. l. DONNELLY SENIOR ENGINEER,
REGULATORY INTERFACE i
I LOGISTICS H.1. DONNELLY SENIOR ENGINEER, REGULATORY INTERFACE
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i INTRODUCTION I
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SIGNIFICANT ERF PROGRAM ACTIVITIES
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e PARTICIPATES IN INPO EMERGENCY PLANNING WORKSHOPS
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PARTICIPATES IN SOUTHEASTERN UTILITY EMERGENCY PLANNING GROUP l
e UNDERTAKEN TSC COMPUTER UPGRADE
.
e MAINTAINS A CONSERVATIVE AND SELF-CRITICAL APPROACH IN EVALUATING THE VCSNS ERF l
PROGRAM
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INTERNAL AUDIT HISTORY l
!
j AUDIT DATES IDENTIFIER MAJOR FOCUS
-
3/17-31/83 CGSZ-282-SQA AUDIT OF EMERGENCY PLAN ACTIVITIES l
l 2/25/83 1-RG5-83-B OBSERVATION OF AN EMERGENCY PLAN DRILL
!
3/16/83 2-RGS-83-B OBSERVATION OF A RADIOLOGICAL DRILL
l 3/14-16/83 6-LCN-83-B OVERVIEW OF ANNUAL EMERGENCY PLAN EXERCISE 7/12-15/83 10-LCN-83-8 INPO ASSESSMENT OF EMERGENCY PLAN
!
3/26-30/84 CGSZ-389-SQA ANNUAL AUDIT OFTHE EMERGENCY PLAN
!
8/30/84 CGSS-13010-SQA QUALIFICATION REQTS. OF THE EMERGENCY COORDINATOR j
POSITION 1/21-29/85 CGSZ-493-SQA ANNUAL EMERGENCY PLAN AUDIT S/13/85 CGSS-14051-SQA EMERGENCY PLAN PROCEDURE INTERFACE 2/20/86 CGSS-628-SQA ANNUAL AUDIT OF EMERGENCY PLAN 1/7-2/4/87 CGSS-16502-SQA ANNUAL AUDIT OF EMERGENCY PLAN 8/24 -9/4/87 CGSS-17543-SQA TRAINING AUDIT (INCLUDES EMERGENCY COORDINATOR)
i 1/11 -22/88 CGSS-17708-SQA ANNUAL EMERGENCY PLAN AUDIT i
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I MAJOR MILESTONES l
l DATE EVENT 1981 MILESTONES (1) AND (2) OF APPENDIX 2 TO NUREG-0654, DEVELOPMENT OF EMERGENCY RESPONSE PLANS AND SUBMITTAL OF EMERGENCY IMPLEMENTING l
PROCEDURES, WERE MET AS DESCRIBED IN SSER #2
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1982 RECEIVED FINAL NRC AND FEMA APPROVAL OF VIRGIL C. SUMMER NUCLEAR j
STATION EMERGENCY PREPAREDNESS AS DESCRIBED IN SSER #3
!
1983 PERMANENT EOF OPERATIONAL 1984 APPROVAL OF BACKUP EOF LOCATED IN COLUMBIA, SC 1985 SUBMITTAL OF RESPONSE TO GENERIC LETTER 82-33, EMERGENCY RESPONSE CAPABILITY SUPPLEMENT 1 TO NUREG-0737 1986 FINAL FEMA ACCEPTANCE OF STATE AND LOCAL PROMPT ALERT AND NOTIFICATION I
SYSTEM I
1987 SAFETY AND TECHNICAL EVALUATION REPORTS RECEIVED RELATIVE TO REGULATORY GUIDE 1.97, REV. 3 REQUIREMENTS i
l SAFETY AND TECHNICAL EVALUATION REPORTS RECEIVED ACCEPTING THE VIRGIL C.
!
SUMMER SPDS FOR OPERATION
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i EMERGENCY RESPONSE FACILITY OVERVIEW
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KEN BEALE
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_ _ _ _ _ _ _ _
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FUNCTIONAL ORGANIZATION Chairman and Chief Executive Officer (SCANA)
John Warren President and Chief Operating Officer (SCE&G)
T. C. Nichols Executive Vice President Operations O. W. Dixon, Sr.
,
Vice President, Nuclear Operations D. A. Nauman Director, Nuclear General Manager, General Manager, General Manager, Plant Operations Nuclear Services Engineering Services Nuclear Safety Ollie Bradham Mike Williams Dave Moore Ken Nettles l
General Manager, General Manager, Station Operations Station Support Jack Skolds Mel Browne
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FUNCTIONAL ORGANIZATION
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l General Manager,
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Station Support Mel Browne
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Manager, Manager, Manager, Nuclear Protection Services Facilitiesand Administration j
Chemistry & Health Physics
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Bill Baehr Ken Beale
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REGULATORY REQUIREMENTS / GUIDELINES
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- NUREG - 0696 - DATED FEBRUARY,1980
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f e NUREG - 0654 - DATED NOVEMBER,1980 e NUREG - 0737, SUPPLEMENT 1 - DATED DECEMBER,1982
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PHYSICAL FACILITIES
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TSC e LOCATION -THE TSC 15 LOCATED ADJACENT TO THE CONTROL ROOM e DESIGN FEATURES l
HABITABILITY - HVAC, HEPA AND CHARCOAL FITLERATION, AUTOMATIC EMERGENCY SYSTEM ACTIVATION l
ON RECEIPT OF SAFETY INJECTION OR RADIATION MONITORING AND HAS OVER 4100 SQ. FT.
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DATA AVAILABILITY - REGULATORY GUIDE 1.97 VARIABLES ARE AVAILABLE ON THE ERDS CONSOLES. VOICE
COMMUNICATION SYSTEMS ARE AVAILABLE.
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i OSC LOCATION -THE OSC 15 LOCATED ON THE 448' ELEVATION OF THE CONTROL BUILDING.
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FUNCTIONALITY-A 2700 SQ. FT. AREA WHICH PROVIDES ADEQUATE SPACE AND FACILITIES FOR
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l LOCATION - THE EOF IS LOCATED IN THE BASEMENT OF THE NUCLEAR TRAINING CENTER (NTC). THE NTC 15 e
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HABITABILITY - HVAC, SEPARATE HEPA FILTER TRAIN, BY PASSED DURING NORMAL OPERATIONS AND HAS
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OVER 8,500 SQ. FT. OF SPACE. RADIATION PROTECTION (SHIELDING) EXCEEDS REQUIREMENTS OF A PROTECTION FACTOR OF 5
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DATA AVAILABILITY-SAME AS TSC. VOICE COMMUNICATIONS SYSTEM AND METEOLOGICAL DATA ARE AVAILABLE.
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B ACK-UP EOF, LOCATION - THE BACK UP EOF 15 LOCATED IN THE COMPANY CORPORATE HEADQUARTERS IN e
COLUMBIA, S. C.THE BACK UP EOF IS APPROXIMATELY 25 MILES FROM THE NUCLEAR STATION.
FUNCTIONALITY-A 2000 SQ FT. AREAWHICH PROVIDES ADEQUATE SPACE AND FACILITIES FOR
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ROOM e DIGITAL METEOROLOGICAL DATATHROUGH A
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READlLY AVAILABLE.
e JOINT DATA RECOVERY GOAL OF 90% (R/G 1.23) MET OR EXCEEDED SINCE 1983
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CURRENT UPDATE f
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e DIRECT ENTRY OF Cl/SEC OR ISOTOPIC uCl/ML AND l
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FLOW e RETURN TO FORECAST IN PROGRESS e FSAR CHAPTER 15 ACCIDENT SCENERIOS l
e END PROGRAM EXECUTION
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TSC/ EOF Communicator Data Sheet
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Time of Pr.
88/03/07 1530 2.
Class of Emergency based on Dose Assessment is GENERAL EMERGENCY 3.
Estimated duration of release in hours is
- See attached sheet for Protective Action Guides (PAG's) **
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Dose Projection Base Datai Current Total release rate (N.G.):
4.74E+03 Curies /sec Event Total release (N.G.):
1.75E+07 Curies Current To tal release aate (IOD.):
9.43E-05 Curies /sec Event Total release (IOD.):
3.51E-01 Curies Windspeed (mph)
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i Wind Direction fromi 300 Stability class (vert)
F Release heighti Ground level Temperature at sitei Dose Rates:
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Whole Body Child Thyroid
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Site Boundary /1 Mile 2.6E+02 R/Hr 1.5E+02 R/Hr 2 miles:
1.iE+02 R/Hr 6.5E+0i R/Hr 5 miles:
3.5E+0i R/Hr 2.iE+0i R/Hr 10 miles.
1.5E+0i R/Hr 8.9E+ 0 0 R /Hr Projected Integrated Dose to 88/03/07 1730
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Gite Boundary /1 Mile:
2 miles.
2.2E+02 Rem 1.3E+02 Ren 5 miles:
7.iE+0i Ren 4.2E+0i Rem 10 miles:
3.0E+0i Rem 1.8E+0i Rem Iq u i v a l en t Who1+
- se Factori 1.4E-01 (Rem /sec)/(Ci/m'3)
-:quivalent Child
.1 Dose Factors 8.4E-02 (Rem /sec)/(Ci/m*3)
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Environmental neasurements l
Location Time Dose Rates Sample Type Results
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Comments:
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- CURRENT CUMULcTIVE THYROID DOSE (arem).......__.....--. __..
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SECTOR W SITE 0-1.5 1.5-2.5 2.5-7.0 7.0-12.5 12.5-20
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10 Mile nap Roloace Start 88/03/07 1500 Wind Speedi 2.2 M/sec 300.0 Deg Pluno Update 80/03/07 1600 Wind D ir.
i Edgo Dose Rate & Type-6.iE-(,2 mren/hr (THYROID)
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n s:f ted area interior of the PLUME represents )i REM /He dose
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I EMERGENCY RESPONSE ORGANIZATION
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ORGANIZATION TECHNICAL SUPPORT CENTER e SUPPORT PERSONNEL
- OPERATIONS
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- PERSONNEL ACCOUNTABILITY
- COM MUNICATOR
- SHIFT ENGINEER
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ERGENCY RESPONSE ORGANIZATION OPERATIONS SUPPORT CENTER e SUPPORT PERSONNEL
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- MATERIALS e M AN AGEMENT PERSONNEL l
OSC SUPERVISOR
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M ECH ANICAL SUPERVISOR
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'!RGENCY RESPONSE ORGANIZATION EMERGENCY OPERATIONS FACILITY l
e SUPPORT PERSONNEL
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- COMMUNICATOR
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DATA LOGGERS
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i DOCUMENTATION AND LOGISTICAL
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OVERVIEW
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HAL DONNELLY
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a GENER......ct<GENCY PLANNING 00CUMENTATION l
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DOSE ASSESSMENT DOCUMENTAT!ON
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COMPUTER SYSTEM DOCUMENTATION
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REACTOR OPERATIONS DOCUMENTATION I
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SOURCE TERM DOCUMENTATION e:
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'T*.C. SUMMER ERF AVOIT QUESTION! ANSWER FORM
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Pers;n n;.cn was presented
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Person formulating answer:
Date:
Person Reviewing answer:
Follow Uo Action Required?
YES / NO (circle one)
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Completed By:
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MANAGEM ENT PRIMARY CC'NTACTS
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.. c: ions, concerns (or g eneral in formation), or a ny difficulty
"j of the people below, please contact Hal Donnelly, at ext. 4722.
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Nuclear Protection Services.......
Ken Beale *............... ext. 4268 Emergency Plan Coordination / Training......... Mark Counts............ ext. 4099 Operations Jerry Shepp ext. 4221
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Health Physics...............
Bill B a ehr............... ext. 4125 Corporate Health Physics and Environmental.............
L. A. Blue ext. 75-5002
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Core Engineering and Nuclear Computer 5 trvices......
Doug Warner ext. 4734
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Design Engineering............ Steve Cunningham....... ext. 4702 Systems and Performa nce Engineering................. Chris Osier ext. 4201
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QA..............
Jim Proper ext. 4088
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QC.............................
Steve H unt............... ext. 4128 Nuclear Licensing................ Hal Donnelly............. ext. 4722
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For long distance access, please dial 77 233 097 9-1-(Area Code) Number.
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l INSPECTION ITEMS LICENSEE CONTACT PERSONNEL-03.02 Assessment of Radiological Releases Corporate Health Physics & Environmental Prcgrams
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l (a)
Source fec.-
Glen Higginbotham 75-5009 (b)
Dose m.a i
03.03 Meteorologic..
,$ation Corporate Health Physics & Environmental..
(a)
Contro l in.... i..sormation Steve Summer
(b)
Represectative Data 75-5004
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(c)
Data Reliability
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(d)
Other Data Availability (e)
NWS Data Availability i
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Data Adequacy
(g)
Data Storage, Display. Analysis
(h)
EOF Data Handling Reliability 03.04 TSC Variable Availability Core Engineering and Nuclear Computer Services (4)
Documentation for Reg Guide 1.97 Al Barton/ Rick Rusaw Variables 4263 (b)
Reg Guide 1.97 Variable Availability and Sufficiency (c)
Computer Data
(d)
Manual Data
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Data Adequacy
l 03.05 TSC functional Capabilities Core Engineering and Nuclear Computer Services (a)
ISC Power Supplies Al Barton/ Gene Baker
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ISC Data Analysis 4263 l
03.06 TSC liabitability Systems and Performance Engineering j
Gene Baker 4504
03.07 ISC Data Collection Storage. Analyses and Core Engineering and Nuclear Computer Services Display Al Barton (a)
Review ISC Systems 4263 (b)
Data Acquisition Systems s
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INSPECTION ITEMS LICENSEE CONTACT PERSONNEL 03.08 EOF Variable Av.ilability Core Engineering and Nuclear Computer Services
(a)
Documestt.1: i... s'or Reg Guide 1.97 Al Barton/ Rick Rusaw Variabi..
4263 (b)
Reg Gui.t.
./ Variable Availability and Sufficie.. 3 (c)
Computer out.
(d)
Manual DeLa (e)
Data Adequacy 03.09 EOF Location and liabitability Nuclear Protection Services (a)
Location Ken Beale (b)
Meets Criteria of Supp. 1 4268 (c)
Habitability
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03.10 EOF functional Capabilities Nuclear Protection Services (a)
Data Analysis Adequacy Ken Beale
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(b)
Backup EOF 4268
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EOF Reliability 03.11 EOF Data Collection Storage. Analysis and Core Engineering and Nuclear Computer Services
Display Al Barton 4263 l
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',The mps of the Virgil C. Summer Station are being provided to reflect fac
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u d q u a rt e r s................ First Floor Conference Room in
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Copiers - Use Code 1075........... Located upstairs in the Security CraftTraining Building Cafeteria 111 Floor, Building # 5 (Aux. Service
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The Conference Room telephone extension is 4451.
If you have any further questions, please contact Hal Donnelly, at ext. 4722.
- ERF CONF ROOM i
Security and Craft Training Building
i Auxiliary Access Portal N
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I Psionary Aness Postal (PAP)
g 2 Serwece Buildeng 3 Tuetsine Buddeng 4 Water Treatenent Building 5 Ausdiasy5ervice huisdeng E Control uuild.ng 7 Intersnedsate g4
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11 Auxiliary Bu.ks...
12 Ileactor Budden.,
13 fuel finnd': ;;
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14 Hot ToolRoosn l45 l
15 Tool Room
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17 Circulating Water
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22 Securi Kraft Training Sidg.
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Cafeteria - # 5 First floor 1H.
Vending Machines:
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SITE MAP
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