ML17299A408
| ML17299A408 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 06/18/1985 |
| From: | Knighton G Office of Nuclear Reactor Regulation |
| To: | Van Brunt E ARIZONA PUBLIC SERVICE CO. (FORMERLY ARIZONA NUCLEAR |
| References | |
| TAC-56654, NUDOCS 8507090180 | |
| Download: ML17299A408 (20) | |
Text
Docket Nos.:
50-52 5 - 29 and 50-530 gag 18 lid5 Mr. E.
E.
Van Brunt, Jr.
Executive Vice President Arizona Nuclear Power Project Post Office 'Box 52034 Phoenix, Arizona 85072-2034
Dear Mr. Van Brunt:
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION - PALO VERDE SAFETY PARAMETER DISPLAY SYSTEM By letter dated February 27, 1985, you provided the safety analysis report for the Palo Verde Safety Parameter Display System (SPDS).
As a result of the staff's review of this report, we have. determined that additional information is needed for the staff to complete the review.
The specific information required is discussed in the enclosed request (Enclosure 1).
We ask that you provide the requested information and inform us within one week of receipt of this letter as to when this information will be provided.
The staff also plans on performing a site audit of the SPDS which will take two full days.
The audit plan and agenda are provided as Enclosure 2.
After you review this agenda, we ask that you propose a reasonable date before September 30, 1985 for the audit.
If you have any questions regarding this request and/or the proposed
- agenda, you should contact Mr. E. Licitra, the Licensing Project Manager.
Sincerely,
Enclosures:
As stated cc See next page DISTRIBUTION NRC PDR LPDR LBA'3 Reading EJordan BGrimes MLey ELicitra JLee GWKnighton George W. Knighton, Chief Licensing Branch No.
3 Division of Licensing 85070>
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Palo Verde Mr. E.
E.
Van Brunt, Jr.
Executive President Arizona Nuclear Power Project P. 0.
Box 52034 Phoenix, Arizona 85072-2034 Arthur C. Gehr, Esq.
Snell 5 Wilmer 3100 Va1 1 ey Center Phoenix, Arizona 85073 Mr. James M. Flenner, Chief Counsel Arizona Corporation Commission 1200 West Washington.
Phoenix, Arizona 85007 Charles R. Kocher, Esq. Assistant Counsel James A. Boeletto, Esq.
Southern California Edison Company P. 0.
Box 800
- Rosemead, California 91770 Ms. Margaret Walker Deputy Director of Energy Programs Economic Planning and Development Office 1700 West Washington Phoenix, Arizona 85007 Mr. Wayne Shirley Assistant Attorney General Bataan Memorial Building Santa Fe, New Mexico 87503 Mr. Roy Zimmerman U.S. Nuclear Regulatory Commission P. 0.
Box 239 Arlington, Arizona 85322 Ms. Patricia Lee Hourihan 6413 S. 26th Street Phoenix, Arizona 85040 Regional Administrator - Region V
U. S. Nuclear Regulatory Commission 1450 Maria Lane Suite 210 Walnut Creek, California 94596 Kenneth Berlin, Esq.
Winston 5 Strawn Suite 500 2550 M Street, NW Washington, DC 20037 Ms. Lynne Bernabei Government Accountability Project of the Institute for Policy Studies 1901-Que Street, NW Washington, DC 20009 Ms. Jill Morrison 522 E. Colgate Tempi, Arizona 85238 Mr. Charles B. Brinkman, Manager Washington Nuclear Operations Combustion Engineering, Inc.'910 Woodmont Avenue Suite 1310
- Bethesda, Maryland 20814 Mr. Ron Rayner P. 0.
Box 1509
- Goodyear, AZ 85338
eO eO
REQUEST FOR ADDITIONAL INFORMATION CONCERNING THE PALO VERDE I, 2, AND 3 SAFETY PARAMETER DISPLAY SYSTEM ENCLOSURE 1
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 requirements and the staff's review of the SPDS.
The staff reviewed the SPDS Safety Analysis and Implementation Plan provided by Arizona Public Service Company (REF. 1).
The staff was unable to complete its evaluation because of insufficient information.
The information required to continue the review is described in the text below.
A.
Variable Selection The selection of SPDS variables was made by the applicant based on PVNGS emergency operating procedures (EOPs).
The PVNGS EOPs are based on the Combustion Engineering Owners'roup's "Combustion Engineering Emergency Procedure Guidelines" (CE EPGs).
The variables selected by the applicant are summarized in the attached Table I (grouping provided by applicant).
1.
The applicant has not provided an evaluation of the relationship/
of these parameters to the CSFs in NUREG-0737, Supplement 1.
The applicant's submittal should be expanded to address this area.
The staff has reviewed the applicant's Safety Analysis Report provided with Reference 1.
While we find the variables selected do comprise a generally comprehensive list, we note that the following variables are not proposed for the PVNGS SPDS:
Neutron Flux (Source Range)
Cold Leg Temperature Steam Generator (or Steamline)
Pressure and Radiation Shutdown Cooling Flow.
2.
Neutron flux is a fundamental variable for monitoring the status of'lant reactivity control and should be monitored for all power ranges.
However, from the variables listed in Table I, it appears that the source range is not displayed on the PVNGS SPDS.
The applicant should clarify whether the neutron flux level for all power ranges is in fact displayed on the SPDS, or justify
that the proposed arrangement provides adequate indication for use under all conditions.
3.
The cold leg temperature, in conjunction with RCS pressure, is a
key variable for brittle fracture considerations.
In addition, cold leg and hot leg temperatures are used in determining the status of natural circulation as a mode of heat removal.
TPe applicant should provide a commitment to add cold leg temperature to the PVNGS SPDS, or provide alternate added variables with justifications that these alternates accomplish the same safety functions, or provide justification that variables currently on SPDS do in fact accomplish the same safety functions.
4.
The steam generator pressure is a key indicator of the integrity of the secondary system for the Core Cooling and Heat Removal CSFs.
For example, in the Excess Steam Demand event in the CE EPGs, the steam generator pressure is used to identify the affected steam generator.
The applicant should provide a commitment to add steam generator pressure to the PVNGS SPDS, or provide alternate added variables with justifications that these alternates accomplish the same safety functions, or provide justification that variables currently on SPDS do in fac't accomplish the same safety functions.
5.
Prior to isolation, steam generator radiation, in conjunction with containment radiation and plant vent stack radiation, provides a rapid assessment of radiation status for the most likely radioactive release paths to accomplish the "Radioactivity Control" CSF.
The applicant should indicate how radiation in the secondary system (steam generator and steamline) is monitored by SPDS when the steam generator and/or their steamline are isolated.
6.
During shutdown cooling and ECCS modes of cooling when the steam generators are not available, shutdown cooling flow is a key indicator to monitor the status of the heat removal system.
The applicant should provide a commitment to add shutdown cooling flow to the PVNGS SPDS, or provide alternate added variables with justifications that these alternates accomplish the same safety functions, or provide justification that variables currently on SPDS do in fact accomplish the same safety functions.
B.
Variable Validation The applicant's submittal describes the methodology used to perform and collect data for the validation and verification of the PVNGS SPDS design.
The method used to validate the SPDS design was programming an actual SPDS with transient data to simulate CSF changes (and challenges) on the SPDS displays.
The applicant simulated the following four scenarios:
small break LOCA, back-up power failure, large steam line
- break, and large steam generator tube rupture.
The description does not address validation of the SPDS variable relationship to the CSFS or the useability of the SPDS display during transients'nd accidents for a rapid assessment of. plant safety status.
The applicant should provide a
description of this'variable validation program discussing the following items:
1.
Relationship to the CSFs by additional consideration given to control room walkthroughs of transients and accidents utilizing the SPDS.
2.
Useability, demonstrating that the following criteria are met:
a ~
b.
The selected transient and accident test cases should exercise the SPDS variables to the fullest extent
- possible, including representative beyond design basis scenarios.
The proposed transient and accident. test cases should cover the instrument setpoints for systems actuation (e.g.,
ECCs actuation) and operator actions (e.g.,
RWT level for switchover) identified in the EOPs.
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REFERENCES 1.
- Letter, E.
E.
Van Brunt, Jr.
(APS) to D.
G. Eisenhut (NRC), dated February 27, 1985.
2.
Supplement 1 to NUREG-0737, "Requirements for Emergency
Response
Capability (Generic Letter 82-33)," dated December 17, 1982.
3.
Safety Evaluation of "Emergency Procedure Guidelines," for CEN-152, Revision 1 (SER, dated July 29, 1983).
Attachment:
Table 1
TABLE I SAFETY FUNCTION VARIABLES PAI 0 VERDE NUCLEAR GENERATING STATION$ UNITS I, 2$
AND 3 Safet Function Reactivity Control (RTV)
Heat Removal (HRV)
Pressure 5 Inventory Control (PIC)
Indirect Radiation Release
( IRR)
Containment Integrity (CIN)
Variable CEA Position Log Power Linear Power HPSI Flow to RCS LPSI Flow to RCS Sub-Cooled Margin CET-T Hot T hot - T cold (Loop I)
T hot (Loop I)
T hot - T cold (Loop 2)
T hot (Loop 2)
Outlet Plenum Level SG-I Level SG-2 Level Steam Flow - Feed Flow 1 Steam Flow - Feed Flow 2
.Sub-Cooled Margin Vessel Head Level RCS Pressure Pressurizer Pressure Pressurizer Level HPSI Flow to RCS LPSI Flow to RCS Plant Vent Stack Condenser Vacuum Exhaust Fuel Building Exhaust S/G I Blow Down Radiation S/G 2 Blow Down Radiation Essential Cooling Water Radiation Control Room Vent Radiation Nuclear Cooling Water Radiation Containment Isolation Verification Containment Pressure Containment Spray Flow Containment Temperature Containment Level Containment Radiation - High Refue]
Pool Radiation H2 Concentration
TABLE 1 (CONTINUED)
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Maintenance of Vital Auxiliaries (VAX)
Variable HPSI Flow to Loop 1
HPSI Flow to Loop 2 LPSI Flow A to Loop 1
LPSI Flow B to Loop 2, CS Flow A CS Flow B Aux. Feed Flow to SG1 Aux. Feed Flow to SG2 Steam Flow - Feed Flow 1
Steam Flow - Feed Flow 2
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ENCLOSURE 2
AUDIT PLAN FOR EVALUATION OF THE PALO VERDE 1, 2, 5
3 SAFETY PARAMETER DISPLAY SYSTEM Back round Staff evaluation of safety parameter display systems (SPDS) consists of reviews of the applicant/licensee's documentation (i.e. safety analysis report and implementation plan) and audit meetings/site visits.
Three separate audit meetings/site visits, as described
- below, may be arranged through the Division of Licensing Project Manager.
As dictated by the comprehensiveness of the applicant/licensee's documentation and the schedule for design and implementation of the
- SPDS, the objectives of these au'dits may be met in fewer site visits.
Design Verification Audit:
The purpose of this audit meeting is to obtain additional information required to resolve any outstanding questions about the Verification and Validation (VIV) Program, to confirm that the'V8V Program is being correctly implemented, and to audit the results of the V&V activities to date.
At this meeting, the applicant should provide a thorough description of the SPDS design process.
Emphasis should be placed on how the applicant is assuring that the implemented SPDS will:
provide appropriate parameters, be isolated from safety systems, provide reliable and valid data, and incorporate good human factors engineering practice.
Design Validation Audit:
After review of all documentation, an.audit may be conducted to review the as-built prototype or installed SPDS.
The purpose of
'his audit is to assure that the results of the applicant/licensee's testing demonstrate that the SPDS meets the functional requirements of the design and to assure that the SPDS exhibits good human factors engineering practice.
Installation Audit:
As necessary, a final audit may be conducted at the site to ascertain that the SPDS has been installed in accordance with the applicant/licensee's plan and is functioning properly.
A specific concern is that the data displayed reflect the sensor signal which measures the variable displayed.
This audit will be coordinated with and may be conducted by the NRC Resident Inspector.
Based on the advanced state'f the Palo Verde design, the staff plans to do a
combined Design Verification and Design Validation audit at a mutually agreeable date.
Audit Schedule The staff anticipates that the audit will take two full days.
After reviewing this agenda the applicant should propose an approp> iate date for the audit.
The staff will attempt to accommodate any reasonable date between April 1 and September 30, 1985.
Since final revision of the design will be complete by June I, 1985, the staff suggests that a mid-June audit may be appropriate.
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NRC Audit Team The NRC Audit Team will consist of representatives from the Human Factors Engineering
- Branch, Procedures and Systems Review Branch, and from the Instrumentation and Control Systems Branch.
In addition, the staff will be assisted in the audit by Science Applications International Corporation (SAIC).
~Ae ad a Day 1:
8:30 AM-10:30 AM Introductions, short entrance briefing (10 minutes) by NRC, overview of SPDS-design program and current status by Arizona Public Service (APS) including:
1) human factors analysis, standards, and criteria used in the design process, with emphasis on plant-specific aspects 2) reliability - a) design characteristics b) methods used to estimate reliability 3) data validation methodology, and display of validation information 4) operator training and. system operating procedures for the SPDS 10:30 AM-2:30 PM (one hour break for lunch during this period) 2:30 PM-4:30 PM Description of Verification of Validation (V&V) program including:
1) description of V&V team and demonstration of independence from the design team 2) available documentation for completed tasks and phases 3) validation of SPDS parameters and how the results demonstrate the representativeness and useability of selected parameters 4) human factors aspects of V&V Program 5) method and results of dynamic simulator testing NRC questions and review of V&V documentation Discussion of isolation devices, including:
a.
For each type of device used to accomplish electrical isolation, describe the specific testing performed to demonstrate that the device is acceptable for its
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application(s).
This description should include elementary diagrams where necessary to indicate the test configuration and how the maximum credible faults were applied to the devices.
b.
Data to verify that the maximum credible faults applied during the test were the maximum voltage/current to which the device could be exposed, and define how the maximum voltage/current was determined.
c.
Data to verify that the maximum credible fault was applied to the output of the device in the transverse mode (between signal and return) and'ther faults were.
considered
(-i.e.,
open and short circuits).
d.
Define the pass/fail acceptance criteria for each type of device.
e.
Provide a commitment that the isolation devices comply with the environmental qualifications (10 CFR 50.49) and with the seismic qualifications which were the basis for plant licensing.
f.
Provide a description of the measures taken to protect the safety systems from electrical interfence (i.e.,
Electrostatic
Day 2:
8:30 AM-9'30 AM Short tour of control room or simulator 9:30 AM-12:00 PN Demonstration of SPDS page formats in TSC/EOF/Simulator.
Walk-through of a plant-specific scenario that involves confirmation of containment isolation, and monitoring of radioactive releases.
1:00 PM-3:30 PM 3:30 PN 4:00 PM NRC audit of displays, display formats, interface devices, access and response times, etc.
NRC caucus Exit briefing
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