IR 05000400/1988012
| ML18005A531 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 07/21/1988 |
| From: | Decker T, Sartor W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML18005A530 | List: |
| References | |
| 50-400-88-12, NUDOCS 8808150219 | |
| Download: ML18005A531 (23) | |
Text
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UNITED STATES NUCLEAR REGULATORY COMMISSION
REGION II
101 MARIETTASTREET, N.W.
ATLANTA,GEORGIA 30323 Report No.:
50-400/88-12 Licensee:
Carolina Power and Light Company P. 0.
Box 1551 Raleigh, NC 27602 Docket No.:
50-400 Facility Name:
Shearon Harris 1nspecti on Conducted:
tiay 1P-19, 1988 I n spec tor:
<~'l g<Zc-~
'M.
YI. Sartor, Jr.
License No.:
NPF-53 Date Signed Accompanying Personnel:
D.
Adams (Battelle)
F.
McHanus (Battelle)
. Milliams (NRC, NRR, PEPB)
Approved by: ~~/mW T.
R.
Dec er, Section Chse Division of Radiation Safety and Safeguards 5 >i c~S'te Signed SUMMARY Scope:
This special, announced inspection was an Emergency Response Facilities (ERF) Appraisal.
Areas examined during the Appraisal included a review of selected procedures and representative records, the ERFs, and related equipment, and interivews with licensee personnel.
Selected activities were observed during the 1988.annual exercise to ascertain the adequacy of the ERFs and related equipment.
Results:
No violations or deviations were identified.
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TABLE OF CONTENTS 1.0 Assessment of Radiolo ical Releases 2.0 3.0 4.0 1.1 1.2 Meteorolo Technical 3 ~ 1 3.1.1 3.1.2 3.1.3 3.1.4 3.2 3.2.1 3.2.2 3.3 3.4 3.4.1 3.4.2 3.4.3 3.4.4 3.4.5 3.4.6 3.4.7 3.4.8 3.4.9 3.2.4.10 3.5 Emer enc 4.1 4.2 4.2.1 4.2.2 4.2.3 4.3 4.3.1 4.3.2 4.3.3 4.3.4 4,4 Evaluation of Source Term Methodology Dose Assessment ical Information Su ort Center TSC Variable Availability Adequacy of Regulatory Guide (RG) 1.97 Variables RG
~ 97 Variable Availability and Sufficiency Computer Data Manual Data TSC Functional Capabilities TSC Power Supplies TSC Data Analysis TSC Habitability TSC Data Collection, Storage, Analysis and Display Methods of Data Collection Data Displays in the TSC Time Resolution Signal Isolation Data Communications Processing Capacities Data Storage Capacity Model and System Reliability and Validity Reliability of Computer Systems Environmental Control Systems Implementation of RG 1.97 0 erations Facilit EOF Location and Habitability EOF Functional Capabilities Data Analysis Adequacy Backup EOF EOF Rel iabi1 ity EOF Data Systems RG 1.97 Variable Availability 8 Sufficiency Computer Data Manual Data Data Adequacy EOF Data Collection Storage, Analysis and Display
5.0 Persons Contacted 6.0 Exit Interview 7.0 Licensee Actions On Previousl Identified Findin s 8.0 Glossar of Acron ms and Initialisms
Assessment of Radi pl o ical Rel eases Since there was no meteorologist assigned to this inspection team, the portions of this section pertaining to atmospheric transport and diffusion models and the adequacy of the meteorological data system were not performed.
However, the inspector noted during his evalua-tion that the Harris plant was using a simple, straight-line Gaussian plume model for its transport and diffusion codes.
This model appeared inadequate since it does not have the capability to make the appropriate changes to the dose projection which resulted from changes in the meteorological conditions (e.g.,
changes in wind direction after the radioactive release begins).
However, the item could not be properly evaluated because there was no team meteorolo-gist to perform an evaluation of the transport and diffusion codes.
A meteorologist should evaluate the transport and diffusion codes for the Harris dose projection models for both the manual and the computerized dose projection systems to determine if these codes are adequate to determine the magnitude of and for continuously assessing the impact of a
radioactive release to the environment ( I F I 50-400/88-12-01)
.
Evaluation of Source Term Methodolo There were six primary gaseous effluent release points at the Harris plant:
the main plant ventilation stack (stack ¹1), the filtered stack on the waste processing building (stack ¹5), the unfiltered stack on the waste processing building (stack ¹5A) and the PORVs and safety relief valves on each of the three steam generators.
A review of the one line drawings given in the figures in the Harris FSAR, Section 9.4, as well as the information provided in Section 11.5, determined one additional effluent release point; the turbine building stack (stack ¹3) which released effluents from the steam condenser.
With the exception of the main steam line monitors which are used to monitor the PORVs and the SRVs for each steam generator, all of these release points were monitored with wide range noble gas and radioiodine monitors.
In addition, all of these points, including the main steam lines, were equipped to obtain a grab sample from the effluent stream.
Stack ¹3 is not considered for source term determination because the monitors in the main steam lines also covered this release point and the percentage of the release that might occur through the PORYs or the SRVs would not be included in the effluent being monitored in this stack.
The only identified unmonitored release pathway was the possible leakage through the reactor containment structure.
The relationship between radioactivity in the primary coolant, core exit temperature and other variables which determine the amount of damage to the fuel were given in EPIP, PEP-362.
The methodology for this procedure was based on the westinghouse Owner's Group,
"Core Damage Assessment Methodology",
November 1984, Rev.
2.
A manual system using a
hand held calculator and a computerized program using
a PC in the TSC were available to the Assessment Team to perform this analysis.
The computer program only handled the analysis of the radionuclides in the PASS samples for the reactor coolant system, the containment atmosphere, and the containment sump.
All of the other methodologies were performed manually using graphs, tables, and simple calculations.
The radionuclide analysis was considered to be the most accurate and reliable method of assessing core damage.
The technical basis for performing dose assessment was described in Annex B of the Harris Emergency Plan which described the rationale for the assumed radionuclide mix for source terms derived from direct radiation measurements.
This rationale was based on the Oak Ridge Isotope Generation and Depletion Code and assumed that the release is 855 noble gas and 15%
radiohalogens.
The primary method for performing dose assessment calculations in the Control Room, TSC and EOF was the DOSESHP program which was run on a
PC as described in EPIP, PEP-343.
"Automation Of Dose Projection-IBM PC", Rev. 4, dated Harch 26, 1987.
The DOSESHP program calculated a source term based on plant data from the Emergency Response Facilities Information System (ERFIS).
The ERIFS display information was generated by both the plant systems computer and the emergency data acquisition system with the upper level displays comprising the SPDS.
The data from ERFIS must be manually entered into DOSESHP.
Source term information is gathered from several different methods.
Radiation monitors used to determine source term for airborne radioactivity were the main steam line monitors for each steam generator; and the wide range noble gas and radioiodine monitors on plant stacks Pl, 85 and 85A.
Grab samples can be obtained from the six monitored release points.
Containment radiation leakage can be measured with the containment monitor of the PASS system.
Environmental monitoring data are also used to predict source terms.
There was no provision in the model for default source terms or handling an unmonitored release pathway and the computer program could be enhanced with a menu of default source terms for use in the EOF based on the accident analyses in Chapter 15 of the Harris FSAR and NOREG-75/014,
"Reactor Safety Study (WASH-1400)".
The manual system for performing dose assessment was provided in PEP-341
"Manual Dose Calculation",
Rev. 4, dated December 2, 1987.
This procedure was used as a
backup for the computerized method during equipment failure.
The manual procedure was essentially the same as DOSESHP and used the same rationale and data inputs.
The data inputs and necessary computations were outlined on a series of forms provided with the procedure.
The data inputs for the dose assessment model for both the manual and computerized system did not have the capability to incorporate particulate radioactivity in the dose projection even though the grab sample analysis could indicate the presence of such radioactivity.
The licensee should consider modifying. DOSESHP to permit the
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inclusion of data on particulate radioactivity in the dose assessment calculation.
The licensee depends heavily on offsite environmental monitoring to val'idate his dose assessments and provide
'a source term for an unmonitored release.
Between the Harris plant and the Harris Energy and Environmental Center (HEEC), five mobile offsite monitoring teams could be dispatched within one hour after an emergency had been declared.
These teams were well equipped with a wide range ion chamber, a micro R meter, a
GM instrument with an HP-210, a high volume air sample equipped with AgK filters, two way radio, personnel dosimetry, and protective clothing.
Each team was equipped with a four-wheel drive vehicle.
A small number of preselected monitoring points (18) were located within one to two miles of the plant site and were for the plant monitoring teams who are less familiar with the site environs than the HEEC personnel.
This concept should be considered for expansion by the licensee to include locations out to ten miles from the plant site.
Also the statement on pages 3-14 in Section 3.9.6 of the Harris Emergency Plan that charcoal filters were used to measure radioiodine in the plume under accident conditions was in error because the technique is not effective and the licensee used AgX which was the appropriate filter material.
This statement should be corrected.
The licensee had extensive radiochemical and environmental laboratories located in the HEEC and a well equipped mobile laboratory which were used to support the offsite monitoring system during emergencies.
The availability to the grab sample stations and the PASS under severe accident conditions was evaluated.
Calculated whole body and extremity doses that would be received for the various PASS samples did not exceed 243 mrem and 919 mrem respectively.
An inhouse evaluation of the dose that potentially could be received from taking grab samples from stack bl under severe accident conditions, dated April 8, 1986, estimated the whole body exposure rates in a range from 10 to 15 mR/h.
Using the exposure rate estimates for a design basis accident (DBA) in the Harris FSAR, the exposure rates in the vicinity of the sample station for stack 81 did not exceed 10 mR/h (Figure 12.3 A-10).
DBA exposure rates for the stack k'5 and 85A sample stations were less than 110 mR/h (Figure 12.3 A-20).
The grab sample points for the main steam lines are located adjacent to the RHR heat exchanger and under DBA conditions the exposure rates could be as high as 800 R/h (Figure 12.3 A-8) and would be inaccessible under severe accident conditions.
A note stating the possible unavailability of these grab sample stations for severe accident conditions should be included in PEP-341 and 343.
Based on the above findings, this portion of the licensee's program appeared to be adequat.2 Dose Assessment 2.0 The dose assessment model used a transport and diffusion code based on a
simple, straight-line Gaussian plume.
The model used the assumptions by D. B. Turner for determining stability class from delta-T and Reg.
Guide 1.145 assumptions for atmospheric dispersions.
Gamma, whole body dose conversion factors for noble gases were taken from NUREG/CR-1918,
"Dose Rate Conversion Factors for External Exposure to Photons and Electrons."
Thyroid dose factors from radioiodine inhalation were taken from RG 1.109 assuming that a child (age 1 to 10 year)
characterized the critical population.
The computerized dose assessment system (DOSESHP)
must be given the source term information manually by inputting the radiation monitor reading by location in mR/h or microcuries/sec.
Grab sample results (in microcurie/cc by radionuclide)
could be substituted in the procedure.
Dose projections are routinely obtained for the plant area and the site boundary (0.47 and 1.3 miles)
as well as 2, 5 and 10 miles; however, the centerline whole body and thyroid dose and dose rate could be provided for other selected distances or intervals.
The manual dose assessment system was set up to use either radiation instrument readings or total radioiodine or noble gas activity and was arranged to calculate centerline whole body and thyroid dose at the five standard distances.
Based on the above findings, this portion of the licensee's program appeared adequate.
Meteorolo ical Information The met tower was located about 1.1 miles NE of the power block.
A 10 meter and 60 meter platform contained sensors that provided wind speed, direction, and atmospheric stability data to the processor located at the base of the tower.
The data was transferred to the Emergency Response Facility Information System (ERFIS) computer which provided 15 minute averages to all ERFIS displays.
ERFIS provided the ability to retrieve historical data as described elsewhere in this report.
Information provided by Carolina Power and Light Company (CPSL) indicated that the system had a greater than 905 availability.
In addition, CPLL maintained its own professional meteorological staff and facilities in Raleigh.
Dose projection was based on the information provided by ERF!S for meteorological conditions at the plant.
However, CPSL's professional meteorological staff provided analysis of weather patterns and systems as they affected plume travel and dispersion.
Information from the National Meather Service, located at Raleigh-Durham airport, was available by phone to the emergency facilities.
Based on the above findings, this portion of the licensee's program appeared adequat.0 Technical Su ort Center (TSC 3.1 3.1.1 TSC Variable Availabilit Ade uac of Re ulator Guide (RG 1.97 Variables Based on a
review of the SER, NUREG-1038 issued May 1986, and inspection of the RG 1.97 variables available in the ERFIS system, all parameters of interest during emergency incidents were available to the TSC personnel.
However, two parameters, accumulator level/
pressure and RCS soluble boron concentration, were under review by NRC staff to confirm adequacy as stated in the SER.
Level/pressure was available but not qualified to
CFR 50.49 requirement.
Boron concentration'as available by sample.
For purposes of this inspec-tion, the data was adequate.
However, NRC staff was reviewing these parameters and additional measures may be required.
3. 1.2
,RG 1.97 Variable Availabilit and Sufficienc Variable values were available in the TSC and EOF via the plant ERFIS.
Radiological information was provided by the plant wide information gathering Radiation Monitoring System (RMS) which feeds to ERFIS
~
The Safety 'Parameter Display System (SPDS)
was a
software-based subset of the ERFIS and was available as a top-level display at any ERFIS terminal and on a
SPDS dedicated display in the Hain Control Room (MCR).
The inspector compared RG 1.97 required variables to the variables available in the ERFIS.
3.1.3 Based upon the above findings, this portion of the licensee's program appeared adequate.
Com uter Data ERFIS consisted of two redundant computer systems and was configured to failover to the backup unit for critical hardware.
The system was powered from an uninterruptible power supply (UPS).
Variables were tagged as to quality to inform users of any questionable data.
Data storage was provided by disk media for short-and mid-term data and by magnetic tape for long-term data.
Trending could be provided on any ERFIS parameter from the short-and mid-term memory and from long-term tape by running the tape.
Based upon the above findings, this portion of the licensee's program appeared adequat.1.4 3.2 Yanual Data In the unlikely event ERFIS was not available, data would be supplied from the control room via a
phone system using trained phone communicators.
Data was recorded at a minimum of every 15 minutes on status boards in the TSC and was available for emergency personnel for determining plant conditions and emergency action levels.
This system was demonstrated during this inspection and it functioned adequately.
Based upon the above findings, this portion of the licensee's program appeared adequate.
TSC Functional Ca abilities 3.2.1 TSC Power Su lies The TSC was supplied power from two separate sources:
(1) start-up transformer 1A and 1B or (2) the unit transformer to 6.9 KV auxiliary buses to 480 Y buses via an auto transfer switch to the TSC loads (HVAC, normal power and lighting, radios, telecopiers, and radiation monitors).
A similar power scheme was used to supply power to the ERFIS computer loads in the facility via UPS with.5 to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> battery backup.
CPKL, during preparation for this inspection, determined that the TSC would require an additional power source during station blackout.
It was determined that a vendor-supplied diesel generator could be obtained and made operational within approximately 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.
During the course of this inspection it was determined that an alternate path for power supply to the ERFIS loads was available from a 1E qualified source.
However, the feasibility of using this path was not determined.
While the existing TSC power supplies appeared to meet the requirement of NUREG 0737, Supplement 1, for a reliable supply, it was noted that the functioning of the TSC would be severely limited if the present power supplies were lost.
The licensee agreed to reanalyze the potential need for an additional power supply and take appropriate corrective action based on the results of the analysis (IFI 50-400/88-12-02).
The existing power supply transfer switch was not functionally tested periodically to verify reliability.
CPSL had identified the need for a functional test of the power supply transfer and committed to having a
test procedure in place by June 30, 1988 (IFI 50-400/88<<12-03).
Communication systems in the facility consisted of the following:
Inplant PABX with normal and alternate diesel power supply and emergency battery backup power supplies.
Emergency PABX (located in the EOF) with normal and alternate (EOF diesel)
and emergency battery backup power supplie.2.2 Auto ring-down phones to State and local agencies.
Nicrowave phone capabilities to all CPSL facilities.
Radio communication to States and county agencies.
Security radio communications to local law enforcement agencies.
Environmental team radio network.
EHS ph'one with alternate diesel power supply.
HPN phone.
Sound-power phones for communications between TSC-OSC-NCR.
ROTE:
both PABXs tie to the commercial Southern Bell exchange located in Apex, NC via dedicated phone lines.
TSC Data Anal sis 3.3 Data trending was provided by the ERFIS computer system and its SPDS'ubsystem.
The licensee demonstrated the SPDS trending ability which permitted trending of preselected parameters of interest.
Selected trending automatically retrieved 8 minutes of historical data and subsequent data scans were displayed.
Long-term (archived)
data could be retrieved from tape storage.
In addition, the SPDS top-level parameter display provided up, down, or horizontal arrow indicators to provide trends from the previous data scan.
In the unlikely event ERFIS was not available, data trending would be accomplished using grid type parameter/time status boards with manual input by communicators collecting data from the Control Room via the plant telephone system.
Nechanical and electrical system drawings, plant operating manuals, FSAR, corporate/plant/State/local emergency plans, and a
document control library were available within the TSC envelope to aid the emergency team in analyzing plant accident and off-normal events.
Based upon the above findings, this portion of the licensee's program appeared adequate.
TSC Habitabilit The TSC was located within the protected area at elevation 324 feet in the fuel handling building.
Malking distance to the NCR was approximately 400 feet.
Protective clothing and breathing apparatus were provided in both the TSC and NCR for personnel who must traverse between the two.
Protection was provided for personnel in the TSC by air filtration and pressurization.
Radiation sensors in the ventila-tion system automatically routed the TSC atmosphere and outside
3.4 3.4.1 makeup air through a single prefilter, HEPA, and charcoal filtration train.
The system was functionally tested every 18 months to ensure proper operation by Administrative Periodical Test (APT)
110
"TSC Performance Test."
A review of APT 110 conducted on September 4,
1987, indicated that proper pressurization and air flows were achieved.
A demonstration of the emergency ventilation system was conducted during which the system failed to achieve adequate pressure.
Investigation revealed a flow control valve out of position, Maintenance was performed on the system to properly balance the air flows.
A second demonstration was conducted with satisfactory results.
An analysis of TSC habitability was performed by EBASCO in September 1983, which concluded that doses to personnel would be below General Design Criteria 19 for a design basis LOCA event.
The direct radiation exposure analysis states that six feet of concrete shielding and a minimum of 88 feet separate the outside surface of the reactor containment and the TSC.
The radioactive piping which would present a problem under these accident conditions in the auxiliary building is also separated from the TSC by six feet of concrete shielding and a
minimum distance of 110 feet.
The contribution to direct radiation dose from either of these sources is insignificant.
The facility is maintained at a positive pressure to prevent the inleakage of the airborne radioactivity except when the access doors are open.
The primary source of direct radiation to the TSC would come from the ventilation filters for the TSC.
The dose rates from these filters were computed to be in the range of 0.4 to 13.2 mrem/h depending on the location of the measurement within the TSC during the first day of the accident.
Doses over a 30 day period for the assumed occupancy factors and which include the exposure due to the inleakage from the radioactive plume are within the 5 rem whole body and 30 rem thyroid and skin dose requirements.
Based upon the above findings, this portion of the licensee's program appeared adequate.
TSC Data Collection, Stora e Anal sis, and Dis la The Emergency Response Facility Information System (ERFIS), acquired from Science Applications Inc., provided information to the Technical Support Center (TSC), the Emergency Operations Facility (EOF),
and the Control Room.
ERFIS provided SPDS capability, balance of plant processing, and nuclear steam supply information to each of these locations.
ERFIS gathered information from field instruments and the Radiation Monitoring Computer System, processed the information, and provided displays to approximately 25 terminals.
Methods of Data Collection ERFIS gathered 1100 (1600 maximum)
analog data points and 1200 digital data points in real-time and provided this data to redundant hardware trains.
These signals were conditioned and digitized in 17 multiplexor cabinets before feeding data to two data
concentrators (data acquisition processors).
The data concentrators decoded and concentrated data needed by the Central Processing Units (CPUs).
The system, configured in redundant trains,.used two Gould 32/67 CPUs each with 6NB of memory, two 675YiB data storage disks, two tape archival devices, and two SONB program storage disks.
The CPUs both processed information in a "primary" or "backup" mode depending on the system configuration and failover status.
The CPUs transfered information using a high speed data link so that in the event of a CPU failure, loss of information was minimal'nformation from the primary CPU was passed to a digital switch linked to the control room, TSC, and EOF peripherals.
The digital switch also passed information from the peripherals (keyboards, for example)
to the CPUs for processing.
The link to the EOF used dedicated fiber optics.
ERFIS was powered by an uninterruptible power supply with one half hour battery reserve.
If the air conditioner failed, the system would be operable for one half hour before temperature limits are exceeded.
The operator interface with the system was through 20 interactive data terminals ( IDTs),
3 video image and store and copy (VISC)
devices, 8 printers, and one high speed line printer.
ERFIS sampled each of the 1100 analog and l200 digital data points ten times per second.
Data was processed depending on a variable scan table within 30 seconds.
Internally, the input sensor values were combined mathematically to provide the operator with calculated point information.
The total number of points within the system (analog, digital, and calculated)
was approximately 5300 points.
During an emergency, data would also be recorded manually on data lists or charts on the walls of the TSC and EOF.
The information would be supplied via telephone from the control room.
Based upon the above findings, this portion of the licensee's program appeared adequate.
Data Dis la s in the TSC The TSC used the same SPDS displays as were used in the control room.
In fact, virtually any display was available on any terminal.
In the TSC within the Site Coordinators Staff Room, two interactive data terminals (IDTs) were suspended from the ceiling on movable swivels.
A third IDT in this room was used to acquire data and control the display on one of the swiveled IDTs.
The display on the other swiveled IDT was controlled from the Accident Assessment Room, across the hall from the Site Coordinators Staff Room in the TSC.
The IDTs were large units with relatively good contras ERFIS provided the user with extensive display capability, including data trends or plots.
Of particular importance were the SPDS displays..
These displays were plant-specific versions of the 11estinghouse Owners Group displays.
The
"SPTOP" (top-level SPDS display)
display provided discrete information related to plant status.
At the bottom of all SPDS displays was a Critical Safety Function l'latrix.'or each critical safety function (e.g.,
subcriticality, core cooling, etc.),
there was a status tree display showing the inputs to the logic used for the critical safety function.
Also displayed was the "data quality" (good,. bad) of the status tree information used in determining the color (green - good, etc.)
in the matrix for the critical safety function.
The status tree display contained entry points to the plant operating procedures.
A third level display provided pre-defined trend plots of SPDS system data.
Based upon the above findings, this portion of the licensee's program appeared adequate.
Time Resolution ERFIS scanned all field points 10 times per second.
Internally, a
variable scan table processed the data at rates of no longer than
seconds, depending on the information requirements for the particular data point.
During the emergency response drill, a request for the
"SPTOP" display required 1 minute and 45 seconds to process, a relatively long period of time.
Based upon the above findings, this portion of the licensee's program appeared adequate.
Si nal Isolation The Safety Evaluation Report approved the Shearon Harris isolation device, a fiber-optic cable, to interface the SPDS with safety-related systems, The NRC required that the Class 1E multiplexor cards within their associated divisions must be powered by that division.
Modifications were to be confirmed prior to fuel loading (Shearon Harris SER 4, page 18-8).
Based upon the above findings, this portion of the licensee's program appeared adequate.
Data Communications Communications to the TSC and EOF were provided via optical RS-232 data links.
Display formats were resident in each of the IDTs so that only refresh data needed to be transmitted.
Error correction was provided using various methods depending on the link involve Given the frequency of data updates on the screen and the speed with which displays were generated once the screen started to be drawn, the communications bandwidth appeared to be adequate (under steady state conditions - minimal load on the system).
Based upon the above findings, this portion of the licensee's program appeared adequate.
Processin Ca acities ERFIS CPU, memory, and disk use could be monitored in real-time from the computer room for each system, primary and backup.
During the exercise in steady state conditions, the CPU use was approximately
to 60 percent of capacity (average),
the memory use was 70 percent used and the disk was observed to be 50 percent used.
System response under these conditions was fair to good.
No documented tests were provided to determine system capacity under accident conditions.
Furthermore, users from the many terminals could at any time request additional plots, logs, or group point sumnaries.
Scan rates or processing rates could also be changed.
Each of these requirements on the system could reduce the processing capacity such that during an accident the system might be totally used and unable to respond to Control Room, EOF, and TSC needs.
As a result of this concern the licensee agreed to examine the need for any user restrictions to assure that ERFIS has sufficient capacity to respond to EOF, TSC, and Control Room information requirements during an accident (IFI 50-400/88-12-04).
Data Stora e
Ca acit ERFIS provided short-term, mid-term, and long-term archival capabilities.
Data storage was provided by the two 675MB disks for shorter term data storage and by tape drives for long-term data storage.
Short-term data (data for the last 8 to 10 minutes)
was available at the terminal with good response times during steady state conditions.
Based upon the above findings, this portion of the licensee's program appeared adequate.
Model and S stem Reliabilit and Validit Validation of the SPDS included SPDS/Emergency Operating Procedure walkthroughs and dynamic simulation tests.
The Verification and Validation plan is referred to in the Safety Evaluation Report (SER 4, 18-8)
and was based on NSAC-39.
The SER confirmed that. the Verification and Validation Team was independent of the Development Team and guality Assurance Progra.4.9 Dose calculation software was well documented and verification was performed by comparing the results with other dose models used by the State of North Carolina as well as the NRC (IRDAN).
Differences in the models were noted and accounted for by personnel other than software developers as described in a
CP&L internal report dated March 1988.
This dose software was used by other Carolina Power
&
Light facilities and is controlled by a specific group within CP&L.
Based upon the above findings, this portion of the licensee's program appeared adequate.
Reliabilit of Com uter S stems Computer system availability was analyzed during development and predicted to be in excess of 99K.
Availability records were being kept in the computer room, and the reported average availability from November 1987 to Hay 10, 1988 was 99.845.
Based upon the above findings, this portion of the licensee's program appeared adequate.
3.4.10 Environmental Control S stems 3.5 Air conditioning was reported by licensee personnel to be functional in the computer room.
If the air conditioning system failed, the computer can continue to operate for approximately one half hour.
Based upon the above findings, this portion of the licensee's program appeared adequate.
Im lementation of RG 1.97 4.0 The implementation of RG 1.97 was acceptable to the NRC with the exception of the accumulator tank and containment sump instrumenta-tion.
This issue was still under review by the NRC.
Emer enc 0 erations Facilit 4.1 EOF Location and Habitabilit The EOF was located in the Harris Energy and Environmental Center approximately 3400 meters from the plant and is designed to exceed the Supplement 1 to NUREG-0737 requirements.
Emergency ventilation was provided by filtered air (HEPA and charcoal)
and the building was pressurized to prevent infiltration.
Analysis conducted in 1983, by EBASCO concluded that adequate protection was provided to personnel for a design basis loss-of-coolant accident.
The direct radiation exposure analysis to determine the PF for the EOF took credit for the azimuthal sectors of gamma contribution and other geometrical factors.
The results of this evaluation assuming 1.0 HeV gammas, indicates a
PF of 18.5 for the first floor of the EOF and 5.1 for the
second floor.
Except for toilet facilities and a vending area, all of the EOF functions are located on the first floor.
This facility is equipped with both charcoal and HEPA filters and is maintained under a
positive pressure during conditions when emergency ventilation is used.
The evaluation of the EOF against GDC 19 and SRP 6.4 indicates that the calculated 30 day whole body dose would be 20 mrem and the thyroid dose 1.6 rem.
During the exercise, the ventilation failed to achieve the specified 0.125 inches of water pressure; it appeared that this was a result of inadequate control of the access doors.
During a
subsequent test conducted for NRC observation, proper pressurization was demonstrated.
CPKL identified a potential source of air entry into portions of the filter train through small unvalved filter train drain lines.
This small amount of air would be partially filtered by the filters downstream of the entry point.
CPSL has instituted corrective actions that will install isolation valves in the drain lines.
4.2 Hased upon the above findings, this portion of the licensee's program appeared adequate.
EOF Functional Ca abilities 4.2.1 Data Anal sis Ade uac 4.22 4.23
, Data analysis was performed using the same systems (ERFIS with status board backup)
as the TSC.
This provided CPSL the ability to cross
, check decisions and conclusions independently in both the TSC and the EOF.
Mechanical and electrical system drawings; plant operating manuals; FSAR; corporate, plant, State, and local emergency plans; and a document control library were available in the EOF to aid the emergency team in analyzing plant accident and'off-normal events.
~Backu EOF CPIIL does not have a backup EOF and has requested an exception from this requirement.
NRC staff was reviewing the request.
EOF Reliabilit The EOF was powered from the normal commercial power grid which was backed up by a dedicated 175 KW diesel generator.
During the inspection, CPSL demonstrated the diesel operability and associated automatic transfer of the EOF loads to this emergency supply.
This demonstration was satisfactory.
Weekly operability test were conducted to insure the continued reliability of the diesel generator.
The inspector noted that the normal power supply.light on the automatic transfer.
switch was out.
Maintenance personnel indicated that this condition,had existed for a considerable period
.of time and would be corrected.
One light on the emergency ventilation control panel was out and was immediately corrected.
Communications capabilities in the EOF were similar to the TS.3 4.3.1 Multiple communication systems throughout the emergency facilities provided a reliable method of communications within CPSL and to Federal, State and local agencies.
Based upon the above findings, this portion of the licensee's program appeared adequate.
EOF Data Systems RG 1.97 Variable Availabilit and Sufficienc 4,3.2 RG 1.97 variables were available in the EOF on
ERFIS terminals.
The ERFIS system was described in the TSC section and the terminals in the EOF have the same capabilities.
See section 3.1.2.
Com uter Data The EOF was equipped with 3 ERFIS terminals linked to the plant via a fiber optic cable.
ERFIS data was adequate for the functioning of the EOF.
See Section 3.1.3.
4.3.3 Nanual Data Status boards (parameter vs.
time grids) were provided in the EOF.
Data was entered every 15 minutes by trained communications personnel using phone links with the TSC or Control Room.
The status boards were maintained and used in a manner similar to the status boards in the TSC.
Visibility was good for all personnel requiring information from the boards.
4.3,4 Based upon the above findings, this portion of the licensee's program appeared adequate.
Data Ade uac The data available in the EOF was from the ERFIS system, which shared the same data base as the other Shearon Harris emergency facilities.
The capabilities in the EOF were similar to the TSC.
EOF Data Collection, Stora e
Anal sis and Dis la The EOF used the same system (ERFIS) for data collection, storage, analysis and display as the TSC and Control Room.
The integrated use of the computer systems, communication systems, and data analysis capabilities provided CP8L the methods and tools to adequately access reactor data during emergency conditions and off normal events for determination of protective action recommendations to State and local agencie.0 The same computers supporting the TSC supported the EOF from a common data base.
The same types of CRTs, keyboards, and hardcopy devices were used in the EOF.
Comments made previously concerning the TSC applied to the EOF.
Persons Contacted 6.0
- R. Baldwin, Senior Specialist, Corporate EP
- R. Black, Manager, Emergency Preparedness J.
Bozeman, Environmental Laboratory Supervisor
- J. Bullock, Principal Engineer
<<J. Collins, Manager, Operations
- T. Drum, Senior Scientist
- G. Fouhand, Director, QA/QC
"A. Garrou, Senior Specialist, Emergency Preparedness
- C. Gibson, Director, Program and Procedures
- J. Harness, Plant General Manager
<<M. Hindman, Jr.,
Manager, Harris Project Administration
- R, Indelicato, Project Specialist, CPLL
"B. McFeaters, Project Specialist, Corporate EP
- B. Meyer, Principal HP Specialist B. Morgan, Senior Specialist, Radiation Control, SHNPP A; Poland, Project Specialist, Radiation Control, SNHPP L. Ratliffe, Senior Specialist, Health Physics, CPSL
- E. Steudel, Principal Engineer, Special Projects
- D. Tibbitts, Director, Regulatory Compliance
- R. Van Metre, Manager, Technical Support
'R.
Matson, Project Vice President Other licensee employees contacted included engineers, technicians, operators, and security force members.
Nuclear Regulatory Commission
- G. Maxwell, Senior Resident Inspector
- Attended exit interview Exit Interview The inspection scope and findings were summarized on May 19, 1988, with those persons indicated in Paragraph 4.O 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 repor,0 Licensee Action On Previousl Identified Findin s (Closed) Inspector Fol1owup Item (IFI) 50-400/87-11-01:
Delayed Telephone Call From Control Room Requesting Offsite Fire Assistance.
An inspector noted that Control Room personnel were knowledgeable about the requirements to promptly request offsite fire assistance if needed.
8.0 b.
(Closed)
IFI 50-400/87-11-02:
Failure to Aggressively Pursue the Coordination of Information Prior to Press Releases.
Corrective action was previously noted by an NRC observer during FEMA required offsite news media corrective action.
c.
(Closed)
IFI 50-400/87-11-03:
Security to Owner Control Area Mas Not Maintained Throughout the Exercise.
An inspector noted that security personnel maintained control into the owner controlled area during the exercise.
Glossar of Acron ms and Initialisms APT CR DBA EOF EPIP ERFIS FSAR HEEC HEPA IFI KY MCR PC PEP PORV RCS RMS SER SPDS SRV SER TSC UPS Administrative Periodical Test Control Room Design Basis Accident Emergency Operations Facility Emergency Plan Implementing Procedure
. Emergency Response Facilities Information System Final Safety Analysis Review Harris Energy and Environmental Center High Efficiency Particulate Air Filters Inspector Followup Item Kilo Volts Main Control Room Portable Computer Plant Emergency Procedure Pressure Operated Relief Valve Reactor Coolant System Radiation Monitoring System Safety Evaluation Report Safety Parameter Display System Safety Relief Valve Safety Evaluation Report Technical Support Center Uninterruptible Power Supply