ML19253B613
| ML19253B613 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 10/08/1979 |
| From: | Pilant J NEBRASKA PUBLIC POWER DISTRICT |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 7910160519 | |
| Download: ML19253B613 (8) | |
Text
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~ ~9 GENERAL OFFICE P. o. Box 499, COLUMBUS, NEBR ASKA 68601 Nebraska Publ.ic Power Distr. t TETE ~o~e<.on s.4-.s.1 ic 1j
__2 October 8, 1979 Office of Nuclear Reactor Regulation Mr. Harold R. Denton, Director U.S. Nuclear Regulatory Commission Washington, DC 20555
Subject:
Potential Unreviewed Safety Question on Interaction Between Non-Safety Grade Systems and Safety Grade Systems Cooper Nuclear Station NRC Docket No. 50-298, DPR-46
Reference:
1.
Letter frem Harold R. Denton to All Operating Light Water Reactors, September 17, 1979
Dear Mr. Denton:
Per your request contained in Reference 1, enclosed is an assessment of Cooper Nuclear Station (CNS) relating to the effect of non-safety system failures on safety system performance. The enclosed report also contains the more specific and comprehensive information and analysis requested by the NRC Staff during a briefing on September 20, 1979.
Efforts on the part of NPPD, General Electric Company and the GE BWR Owners Groups were utilized in making this assessment.
The assessment has not identified any impact on safety actions or. analysis conclusions which would increase the consequences (i.e., calculated peak cladding temperature, peak containment pressure, peak suppression pool temperature, or radiological release) of any FSAR events.
In particular, the assessment concludes that:
1.
No previously identified safety actions wauld be negated by the failure of non-safety equipment due to environmental effects of high energy pipe breaks (HEPB's);
2.
No previously identified safety limits would be violated by the subject effects; and 3.
Some additional operator actions could be helpful to more quickly mitigate the subject postulated effects.
In fully assessing this issue, it should be noted that:
g I
i i 150 202 7 91016 0 55' e
Mr. harold R. Denton October 8, 1979 Page 2 1.
An extensive evaluation of plant safety as regards HEPB's was submitted as Amendments 20 and 25 to the Cooper Nuclear Station FSAR. As a result of this previous evaluation, modifications to the station were made including pipe replacement, installation of pipe whip restraining structures, and installation of crea high temperature alarms annunciated in the control room. In light of the more severe criteria established in Reference 1, the Architect-Engineer for CNS has re-evaluated the previous safety audit and has confirmed that safe shutdown capability will be maintained in the event of a HEPB.
2.
The General Electric BWR includes a number of inherent character-istics which are specifically important in assessing this issue:
a) Thorough evaluation of outside containment line breaks for radiological reasons has resulted in a set of comprehensive sensitive leak detection and isolation systems on BWR's; b) The BWR does not depend to a great extent on non-safety equipment for safety actions:
c) The separation of protection systems from control systems has long been a rule relative to safety function reliability, and there exists almost a complete decoupling of the BWR nuclear steam supply and containment system from non-safety BOP equipment and functions; d) As previously noted, HEPB analyses have been performed and verified physically at BWR facilities; e) The BWR has treated intersystem relationships in considerable detail in a standard SAR section entitled Plant Nuclear Safety Operational Analysis. This systemacic evaluation of the BWR system has proven to be very valuable relative to environmental impact effects analysis; f) Transient and accident analyses of BWR's are conservatively bounded in most cases with respect to non-safety system performance.
In su= mary, this submittal is a complete and comprehensive re-evaluation of the potential impact of non-safety systems on safety functions. The previously approved safety evaluations for Cooper Nuclear Station remain valid.
Sincerely, M
Jay.!.
Pilant Director of Licensing 1150 03 2
and Quality Assurance JDW/cmk Enclosure
Mr. Harold R. Denton October 8, 1979 Page 3 STATE OF NEBRASKA )
) sn PLATTE COUNTY
)
Jay M. Pilant, being first duly sworn, deposes and says that he is an authorized representative of the Nebraska Public Power District, a public corporation an' political :ubdivision of the State of Nebraska; that he is July authorized to submit this information on behalf of Nebraska Public Powet-District; and that the statements in said appli-cat. inn are true to the best of his knowledge and belief.
N M y h. Pilant
'- x
[eb Subscribed in my presence and sworn to before me this day of October, 1979.
//)) JJ 2 J I'" O uY PUBLIC
/
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1 My Commission expires MWM stateat satrasu MARLYN R. HOHNDORF Nr h Esp. Oct.14. Iso 1150 ?04
ENCLOSURE EFFECT OF NON-SAFETY SYSTEM FAILURES (POSTULATED DUE TO ADVERSE ENVIRONMENT)
ON PERFORMANCE OF SAFETY EQUIPMENT The following table identifies the non-safety systems at Cooper Nuclear Station, and the effect of their postulated failure on safety system performance for a variety of postulated high-energy pipe breaks, locations, and sizes. No "1" entries were identified which would denote a possible adverse effect.
TABLE LEGEND 1 - Environmental Induced Malfunction May Provide Adverse Response (i.e., increase in: calculated peak cladding temperature, peak containment pressure, peak suppression pool temperature, or radiological release) 2 - Environmental Induced Malfunction Will Not Provide Adverse Response 3 - System Is Qualified For Adverse Environment 4 - System Will Not Experience Adverse Environment 1150 205
Page 1 of 3 LT1 CD r0 COOPER E'ICLtd - STATION O
ENVIRONMENTAL INTERACTION TABLE Ch MAIN STEAM FEEDWATER LOCA RWCU RCIC llPCI INSIDE INSIDE REACTOR TURBINE INSIDE REACTOR TUPBINE REACTOR REACTOR REACTOR SYSTEMS LOCATION SHAI.L LARGE IILDG.
BLDC.
BLDG.
BLDC.
SMAI.L LARGE BUILDING BUII. DING BUILDING Hecirculation System:
Pumps DW 2
2 4
4 2
4 4
2 2
4 4
4 V<alves & Oper.
DW 3
3 4
4 3
4 4
3 3
4 4
4 HG Set RB 4
4 4
4 4
4 4
4 4
4 4
4 4
4 4
4 4
4 4
4 4
4 Flow Control Sys.
CR/RB 4
4 4
4 4
4 4
4 4
4 4
4 Control Inst. Trans.
RB 4
4 4
4 4
2 4
4 4
4 4
4 Feedwater Delivery System:
Flow Elements TB 4
4 2
2 4
4 2
4 4
4 4
4 Level DW/RB 2
2 4
4 2
4 4
2 2
4 4
4 Pumps TB 4
4 4
2 4
4 2
4 4
4 4
4 Valves & Oper.
TB 4
4 4
2 4
4 2
4 4
4 4
4 Flow Control Sys.
CR 4
4 4
4 4
4 4
4 4,
4 4
4 Feedwater lleating Til 4
4 4
2 4
4 2
4 4
4 4
4 Instrument Air TB 4
4 4
2 4
4 2
4 4
4 4
4 Control Inst. Trans.
RB/TB 4
4 2
2 4
2 2
4 4
4 4
4
M Page 2 of 3 Cn O
COOPER NUCLEAR STATION ENVIRONMENTAL INTERACTION TABLE y
MAIN STEAM FEEDWATER LOCA RWCU RCIC llPCI INSIDE INSIDE REACTOR TURBINE INSIDE REACTOR TURBINE REACTOR REACTOR REACTOR SYSTEMS 1.0 CATION SMALL LARCE BLDG.
BLDG.
BLDC.
BLDC.
SHAll, LARGE BUII. DING BUILDING LUILDING_
Turbine Pressure Control:
By Pasa Valves TB 4
4 4
2 4
4 2
4 4
4 4
4 Pressure Sensors TB 4
4 4
2 4
4 2
4 4
4 4
4 Control System CR 4
4 4
4 4
4 4
4 4
4 4
4 Neutron tionitoring System:
LPRtt's & Cables DW/RB 2
2 2
4 2
2 4
2 2
2 4
4 APlut's & Cables DW/RB 2
2 2
4 2
2 4
2 2
2 4
4 ItPIS/ Rod Block Hon.
DW/RB 2
2 2
4 2
2 4
2 2
2 4
4 TIP DW/RB 2
2 2
4 2
2 4
2 2
2 4
4 Reactor Protection System:
.s 4
2 4
4 2
4 4
4 4
4 MG Set CB 4
.i 4
4 4
4 4
4 4
4 4
4 Reactor Manual
_Cout rol Svst em Ril/CR 4
4 4
4 4
4 4
4 4
4 4
3 3
4 3
3 4
3 3
4 4
4 Rl3CCW System RB 4
4 2
4 4
2 4
4 4
2 4
4 RWCU DW/RB 3
3 2
4 3
2 4
3 3
2 2
2 6
a Page 3 of 3 Ln O
N COOPER NUCLEAR STATION O
ENVIRONMENTAL INTERACTION TABLE CO HA1N STEAM FEEDWATER LOCA RUCU RCIC llPCI INSIDE INSIDE REACTOR TURBINE INSIDE REACTOR TURBINE REACTOR REACTOR REACTOR SYSTEMS 1.OCATICN SHALL 1.ARGE BLDG.
BLDC.
BLDG.
BLDG.
SHALI. LARGE BUILDING BUII. DING BUILDING Suppression Pool:
Temperature Monitoring RB/ Torus
?
2 2
4 2
2 4
2 2
4 a
4 4
Level Monitoring
!8/ Torus 4
4 2
4 4
2 4
4 4
4 2
2 Circulating Water System (Non-Safety)
Intake /TB 4
4 4
2 4
4 2
4 4
4 4
4 IIVAC System All 2
2 2
2 2
2 2
2 2
2 2
2 Non IE 11attery System CB 4
4 4
4 4
4 4
4 4
4 4
4 A. C. Auxiliary Electric RB/TB 4
4 4
4 4
4 4
4 4
4 4
4 Condensate Transfer and Storane T!!
4 4
4 2
4 4
2 4
4 4
4 4
Main Tutbine & Controls
'i.
4 4
4 2
4 4
2 4
4 4
4 4
Main Condenser & Control TB 4
4 4
2 4
4 2
4 4
4 4
4 Instrument (Control)
Air System:
Compressor, Piping and Controls TB/RB/DW 2
2 2
2 2
2 2
2 2
2 2
2 Fire Protection System R3/TB 4
4 2
2 4
2 2
4 4
2 2
2 CRD llydraulic System 4
2 4
4 2
4 4
4 4
4 4
_INan-' cram)
RV ilead Vent DW 2
2 4
4 2
4 4
2 2
4 4
Standby Liquid Control DW/RB 3
3 4
4 3
4 4
3 3
4 4
4 i
The following non-sau.;y grade syste=s/ equipment cannot conceivably fail so as to impact the safety analysis and the adequacy of the protective functions performed by safety grade equipment:
Station Lighting Communications Service Air Equipment Drain Piping Dryvell Temperature Monitoring Under Vessel Maintenance Equipment Process Computer Area Radiation Monitoring Process Radiation Monitoring (Non-Safety Part)
Plant Process Sampling Maintenance Monorails & Hoists Pocable Water Screen Wash Turbine Building Closed Cooling Water Generator Cooling (Generator)
Offgas Radwaste Fuel Pool Cooling and Cleanup Generator Seal Oil Plant Heating Make Up Water Treatment Tarbine Oil Purification and Transfer 1150 209 L