ML20210R401
| ML20210R401 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 05/02/1986 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | Wilgus W FLORIDA POWER CORP. |
| References | |
| NUDOCS 8605160411 | |
| Download: ML20210R401 (6) | |
Text
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May 2,1986 DN Docket No. 50-302 Mr. Walter S. Wilgus Vice President, Nuclear Operations Florida Power Corporation ATTN: Manager, Nuclear Licensing
& Fuel Management P. O. Box 14042; N.A.C. H-3 St. Petersburg, Florida 33733
Dear Mr. Wilgus:
In our letter of December 17, 1985, we made a request for additional information on the Safety Parameter Display System (SPDS) for Crystal River Unit 3.
Per your letter of January 20, 1986, your response will be provided by June 30, 1986. Since the original request, our staff has asked for further information to complete our review. This additional request is identified in the attached enclosure.
Please provide your response to this request as well as the earlier request by the June 30, 1986 date. This request for additional information affects fewer than ten respondents; therefore, OMB clearance is not required under P.L.96-511.
Sincerely,
Enclosure:
As stated cc w/ enclosure:
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9 Mr. W. S. Wilgus Crystal River Unit No. 3 Nuclear Florida Power Corporation Generating Plant cc:
Mr. R. W. Neiser State Planning and Development Senior Vice President Clearinghouse and General Counsel Office of Planning and Budget Florida Power Corporation Executive Office of the Governor P. O. Box 14042 The Capitol Building St. Petersburg, Florida 33733 Tallahassee, Florida 32301 Mr. P. McKee Mr. F. Alex Griffin, Chairman Nuclear Plant Manager Board of County Comissioners Florida Power Corporation Citrus County P. O. Box 219 110 North Apopka Avenue Crystal River, Florida 32629 Inverness, Florida 36250 Mr. Robert B. Borsum Babcock & Wilcox Nuclear Power Generation Division Suite 220, 7910 Woodmont Avenue Bethesda, Maryland 20814 h
Resident Inspector U.S. Nuclear Regulatory Commission Route #3, Box 717 Crystal River, Florida 32629 Regional Administrator, Region II U.S. Nuclear Regulatory Commission 101 Marietta Street, Suite 3100 Atlanta, Georgia 30303 Mr. Allan Schubert, Manager Public Health Physicist Department of Health and Rehabilitative Services 1323 Winewood Blvd.
Tallahassee, Florida 32301 Administrator Department of Environmental Regulation Power Plant Siting Section State of Florida 2600 Blair Stone Road Tallahassee, Florida 32301 Attorney General Department of Legal Affairs The Capitol Tallahassee, Florida 32304
o REQUEST FOR ADDITIONAL INFORMATION CONCERNING TFE CRYSTAL RIVEP UNIT 3 SAFETY PARAMETER DISPLAY SYSTEM Each operating reactor shall be provided with a Safety Parameter Display System (SPDS). The Commission approved requirements for an SPDS are defined in NUREG-0737, Supplement 1.
In the Regional workshops on Generic Letter 82-33, held during March 1983, the NRC discussed these reouirements and the staff's review of the SPDS.
The staff reviewed Florida Power Corporation's SPDS Safety Analysis (Ref. 1) and was unable to complete the review because of insufficient information.
The following information is required to complete the review.
Parameter Selection As a result of our review, the staff noted that the followirc variables are not displayed in the Crystal River SPDS:
ag 1.
Neutron flux, power range
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Neutron flux, intemediate range 3.
Reactor Core System, water level 4.
Residual Heat Removal System, coolant flow rate 5.
Containment Sump, water level 6.
Status of Containment Isolation 7.
Containment Hydrogen Concentration 8.
Steam Generator Radiation.
The rate of change in neutron production (neutron flux) is a fundamental neutronics variable for assessing the status of the Reactivity Control Critical Safety Function (CSF) and should be monitored for all power ranges.
From the data in Reference 1, it is not clear if neutron flux, power range, and neutron flux, intermediate range, are monitored and displayed in the SPDS.
Neutron flux, power range, is an important variable to monitor during an ATWS event. Neutron flux, intermediate range, is an important variable to monitor during a re-criticality event subsequent to a reactor trip. The licensee should clarify whether these variables are displayed on the SPDS, or provide justification that the current design provides adequate indication for use under all conditions, i
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. Reactor core water level is a fundamental variable for monitoring the Reactor Core Cooling CSF during large break and small break loss of coolant accidents. The staff recommends that the Reactor Core Water level be added to the SPDS, or the licensee should provide justification that the current design provides adequate indication for use under all conditions.
During emergency operations, when the steam generators are not available, the Residual Heat Removal System (RHR) and the Emergency Core Coolant System (ECCS) remove heat from the core and keep it cool. RHR coolant flow rate is a critical variable within the Heat Removal CSF. The licensee's SPDS does not contain this variable. We recommend that the variable be added to the SPDS, or that the licensee justify that the current design provides adequate information to evaluate the Heat Removal CSF under all conditions.
Containment sump level is a key variable to identify a LOCA-type breach of the Reactor Coolant System Integrity CSF, particularly for small leaks during which RCS pressure remains constant. The licensee's SPDS does not contain this variable. We recommend that containment sump level be added to the g~
SPDS, or that the licensee justify that the current design provides adequate information to evaluate the Reactor Coolant System Integrity CSF.
Our review of the licensee's safety analysis noted that the SPDS does not display a direct indication of containment isolation. A displayed variable on containment isolation is important in making a rapid assessment of
" Containment Conditions."
In particular, a determination that known process pathways through containment have been secured provides significant additional assurance of containment integrity. Therefore, the licensee should add this data to the SPDS, or provide justification for its exclusion.
Containment Hydrogen Concentration is a key variable to monitor in evaluating the Containment Condition CSF.
For some accident scenarios, hydrogen can be produced and released in containment.
Combustion of large amounts of such hydrogen may result in failure of the containment structure. The licensee's SPDS does not monitor Containment Hydrogen Concentration. We recommend that the licensee add this data to the SPDS, or provide justification for its exclusion.
Prior to a steam generator's isolation, steam generator radiation data and plant vent stack radiation data provide information for monitoring the most likely radioactive release paths. These data are monitored and displayed by the licensee's SPDS. However, the licensee should also indicate how radiation in the secondary system (steam generator and steamline) is monitored from the SPDS when the steam generator and/or their steamline are isolated.
. Data Validation The licensee's safety analysis did not describe how data are validated prior to their display and use by control room personnel. We request that the licensee describe the technique (s) used to validate data, and to describe how valid / invalid data are coded for display in the SPDS.
Human Factors Program The licensee's safety analysis states that the SPDS will be located in the control room. The SPDS provides information to control room personnel via selectable displays and automatically displays alert signals. The selectable displays include:
(1) the Low-Range Pressure-Temperature (P-T) format, (2) the ATOG P-T format, (3) the Inadequate Core Cooling (ICC) format, (4) two " normal" power operation formats, and (5) fcur pages of alphanumeric data, which contains several key process variables ard system variables. The location of data within the SPDS was defined in Table 2.1, " Parameters Required to Monitor the Five SPDS Safety Functions," of the safety analysis.
Nn illustrations of these display formats were provided in the safety analysis.
Our analysis of the data contained in Table 2.1 identified several potential human engineering discrepancies (HEDs).
For example, we noted that information on the Reactor Ceolant System Integrity CSF is centained in:
(1) the Low-Range P-T display format, (2) the AT0G P-T display format, (3) the ICC display format, (4) the alphanumeric display format, and (5) alerts. We fail to s?3 how a concise display of data for the Reactor Coolant System Integrity,CSF is achieved. We request that the licensee provide a description of the human factor principles and practices used for the design and development o,f the display system, or to provide justification that clearly illustrates how each of the critical safety functions may be rapidly and reliably evaluated by a user of the display system.
Design Verification and Validation Program The licensee's safety analysis did not contain a description of the Design Verification and Validation Program used to develop the system.
In Section 3.0, " Applicable Events," of the safety analysis, a general discussion on six initiating events used in the AT0G development program is presented. However, from the material presented, it is not clear that the SPDS was validated. No evidence was provided to demonstrate that the SPDS does allow a user to rapidly and reliably evaluate each of the CSFs for a wide range of events. We reouest that the licensee provide the staff with the Design Verification and Validation Program used in the development of the SPDS.
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i REFERENCE i
I Letter from G. R. Westafer, Florida Power Corporation, to John F. Stolz, NRC,
Subject:
NUREG-0737, Item I.D.2 and Supplement 1, Safety Parameter Display System, dated August 30, 1984.
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