ML20012D909
| ML20012D909 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 03/15/1990 |
| From: | Parker T NORTHERN STATES POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), Office of Nuclear Reactor Regulation |
| References | |
| REF-GTECI-A-47, REF-GTECI-SY, TASK-A-47, TASK-OR GL-89-19, NUDOCS 9003290080 | |
| Download: ML20012D909 (5) | |
Text
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Northem States Power Company 414 Nicollet Mall -
Minneapolis, Minnesota 554011927 Telephone (612) 330-5500
' March!15, 1990 NRC Ceneric Letter 89-19' Director of Nuclear Reactor Regulation
-U S Nuclear Regulatory Commission
-Attn:
Document-Control Desk
. ashington, DC-20555 W
Y PRAIRIE ISLAND NUCLEAR GENERATING PIANT
. Docket Nos. 50-282-License Nos. DPR-42
'50-306
.DPR 60 Response to NRC Generic Letter 89-19 Unresolved Safety Issue A-47 " Safety' Implication of Control' Systems in LWR Nuclear Power Plants"-
Our response to Generic Letter 89-19 is attached, Please contact us if you have any questions related to our response, Thomas'M; Parker-Manager Nuclear ' Support - Services.-
c:. Regional Administrator - Region III, NRC
-Senior Resident Inspector, NRC NRR Project Manager, NRC G Charnoff
-Attachments
- 1. Affidavit
~2, Response to NRC Generic-Letter 89-19 9
9003290080 900315 A
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UNITED STATES NUCLEAR REGULATORY COMMISSION
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PRAIRIE' ISLAND NUCLEAR GENERATING PLANT DOCKET NO, 50-282 50-306 O
j UNRESOLVED SAFETY ISSUE A-47 " SAFETY IMPLICATION e
OF CONTROL SYSTEMS IN LWR NUCLEAR POWER PLANTS"'
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-Northern State.s Power Company, a Minnesota corporation, with i this letter -is submitting information requested by NRC; Generic Letter 89-19.
e This letter contains.no restricted or other defense information.
NORTHERN STATES POWER COMPANY By
-Thomas M Parker
-j Manager, Nuclear Support Services-i I
4 #8 before me a notary public in and for said County, personally appeared Thom[as M Parker, Manager, Nuclear ~ Su On' this/
ay o and;being~first duly sworn acknowledged that he.is authorized to execute this document on-behalf of Northern States Power Company, that he knows the contents thereof, and that t'o the best of his knowledge, information, and belief the statements made in it are true and that it is not interposed for delay.
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. JUDY L. MLAPPERICK NOTARY PUBtlC-MINN0014 5 ANOKA COUNTY My Commaston Expres Sept 8, M
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PRAIRIE ISLAND NUCLEAR GENERATING PLANT RESPONSE TO NRC'CENERIC LETTER 89-19
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y The original l design for steam generator overfill protection at Prairie _ Island
- Units 1 and'2 is defined by-Group I on page 4 of. Enclosure-2-to Generic Letter 89-19.
This is ione of the " Plants that have an overfill-protection system i
initiated on 'a steam generator _ high-warer_-level signal based on :.
a 2-out-of-3 initiating. logic which is safety grade' but uses one out of the three
' channels for both control and protection." This system isolates' feedwater to the
-steam generators by tripping the main feedwater pumps and closing the feedwater
. main and bypass control valves.
k The Generic Letter concludes this design is acceptable if:
o A.
Tho' feedwater control system is not powered from the same source as overfill protection.
B.
Overfill protection and feedwater control are not located within the same cabinets.
C.
Overfill protection and feedwater control signals are routed such that, a fire is not likely to affect both systems.
D.
Plant procedures and Technical Specifications include requirements to periodically verify operability of overfill prorection.-
Each criterion is individually addressed below in the context of the original Prairie Island design; then the added benefit of the upgraded feedwater control system, with a Median Signal Select, is discussed.
1 A.
-System Power: Overfill protection is provided through trip bistables in the Reactor Protection analog instrumentation racks, which are powered
-from Red (Panel 112), White (Panel 111),' Blue-(Panel 113), and Yellow (Panel 114) 120 VAC instrument buses. Upon bistable actuation, Train A and Train B 120 VAC Protection relays (normally powered by the bistable) are deenergized, and the relay contacts (configured in a 2 out of 3 matrix) close to energize 1.25 VDC safeguards relays, resulting -in j
overfill protection' actuation.
1 In general, the control systems are powered from the same 120 VAC-panels; however, different source breakers are used to provide 1
separation from the protection supply. An instrument bus failure would
-l cause the 120 VAC relay to deenergize, actuating one channel of overfill protection.
This aspect of the licensed design of the plant has been discussed with Westinghouse personnel and it is believed the design is adequate regarding overfill protection.
B.
Location:
Overfill protection and feedwater control are physically located in separate cabinets.
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Routing: - No control' signal cables are routed' with protection cables,
(Red,' White, Blue, Yellow,-Train A, or Train B).
This is the case for feedwater control / overfill protection; thus the likelihood of a fire affecting both systems is minimized.
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D.
Technical Specifications / Surveillance Procedures: Steam Generator level.
channel testing is required to be performed at a ; frequency specified by Technical Specifications (Table 4.1-1, Item 11). The only aspect of the channels required to be tested per Technical Specifications (Table 3.5-2, Item.12) is -the Low-iow Level actuation functions. Thus, testing of the high level trip actuation (feedwater pump trip and valve closure) is not specifically addressed by Technical Specifications.
However, this function is presently tested by Prairie Island procedures.
Individua1' protection channels are-tested via monthly analog. protection testing (SP1003).
The 2 out-of-3 safeguards relay matrix is tested monthly via safeguards logic testing (SP1032A). Verification of relay energization is tested via the Integrated SI Test (SP1083). every outage.
Field actuation is tested via feedwater isolation testing (SP1143) every -
outage.
p Feedwater Control System Unerade Prairie. Island has recently upgraded its feedwater control system on Unit 2 (4/89)_and Unit 1 (1/90).
The control system now uses all three narrow range g-steam' generator level channels for protection and control, with. the implementation of a Median Signal Select on steam generator level control channels to provida signal validation.
This prevents protection / control interaction upon failure of a single level channel and is used to satisfy IEEE R27, Paragraph 4.7.3 criteria for reactor protection (low level trip,
Reference:
Westinghouse WCAP-11931). In addition, the NRC has accepted the use of a median
- signal select, implemented in control grado equipment, to address steam generator overfill concerns at another Westinghouse plant (
Reference:
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J Although there are no licensing requirements to test the median signal select, it is implicitly tested through the proper operation of automatic steam generator i
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, level control while monthly surveillance of protection channels is occurring.
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Additionally, the median signal select function for narrow range steam generator level is specifically verified (and documented) by ensuring that a failed channel (high or low) will not be selected for control.
l The reliability of the new control system is supported by redundant 120 VAC power supplies (from the aforementioned instrument buses). The failure of a single 120 VAC bus will have no effect on automatic feedwater control.
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- Thus, the new contro1~ system characteristics enhance.the licensing basis of the e
plant with' regard to steam generator overfill protection. This: design has been
. accepted by the' NRC via an approved License Amendment for Prairie Island; however, it should be noted that CL 89-19-does not recognize this design.-
The following actions will be taken:
A.
A License Amendment Request will be submitted to:
l Revise Table TS 3.'5-4 to include Feedwater Isolation, as shown below:
FEEDWATER ISOLATION a.-
Hi-Hi Steam Generator Level b.
Safety Injection c;
Reactor Trip with Low T,y, (Main Valves only)
' B.' Feodwater = isolation curve 111ance procedures will be revised to ' verify that control valves move from full open to full closed in-5 seconds or
.lcss, upon deenergizing either air supply solenoid. valve. Each of these-i procedures will be revised prior to the next refueling outage for which
.the surveillance is performed.
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