ML20098H069
| ML20098H069 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 10/02/1984 |
| From: | Edelman M CLEVELAND ELECTRIC ILLUMINATING CO. |
| To: | Youngblood B Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8410090335 | |
| Download: ML20098H069 (5) | |
Text
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T H E C L E V E L A N D E L E C T R I C I L L U M ii1 AT I N G C 0 in P A il Y P.O. BOX S000 - CLEVELAND, OHIO 44101 TELEPHONE (216) 622-9800 - ILLUMINATING BLDG. - f 5 PUBLICSOUARE Serving 7he Best Location in the Nation MURRAY R. EDELMAN vice mSIDENT October 2, 1984 NUCu^a PY-CEI/NRR-0139 L Mr. B. J. Youngblood, Chief Licensing Branch No. 1 Division of Licensing U.S. Nuclear Regulatory Commission Washington D.C.
20555 Perry Nuclear Power Plant Docket Nos. 50-440; 50-441 Response to NRC Question No. 420.03 through 420.07-Effects of High Energy Line Breaks on Control Systems (0 pen Item No.14)
Dear Mr. Youngblood:
This letter and its attachments are provided in response to your request for additional information (dated January 10,1984), regarding effects of high energy line breaks on control systems (0 pen Item No.14) for the Perry Nuclear Power Plant (Units 1 and 2).
The information provided in the attachments will be incorporated into a future FSAR Amendment.
If you have any questions, please feel free to call me.
Very truly yours, f8 N
Murray R. Edelman Vice President Nuclear Group MRE:njc Attachments cc: Jay Silberg, Esq.
John Stefano Jack Grobe 8410090335 84100g pj@\\
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Q420.03 Provide an identification of.the locations (elevations / areas) which contain high energy piping systems and in which E
components for the nonsafety related control systems are located.. Relate these to the adverse conditions discussed in o
your letter dated March 14, 1983.
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" Response The following information identifies the locations
~ (panel, building, elevation and coordinates) of.
non-safety control systems and high energy lines for adverse conditions #1-#5 discussed in our March 14, 1983 letter. We have referenced layout drawings contained in the FSAR for your convenience in identifying locations.
ADVERSE CONDITION #1 INSTRUMENT PANEL BUILDING ELEVATION COORDINATES u
1833N001A 1H22-P025 Containment 620 C1/12 IB33N0013 1H22-P041 Containment 620 C1/17 1C34N003B 1H22-P025 Containment 620 C1/12 1C34N0030 1H22-P041 Containment 620 C1/17 HIGH ENERGY LINE BUILDING ELEVATION C0 ORDINATES C11-Control Rod Hydraulic Containment 630 C1/11 l
System C41-Standby Liquid Control Containment 602 C1/17 All of the abo *;e instruments and high energy lines are located between the drywell and the reactor building wall. Exact locations may be identified by referencing figures 3.6-4, 3.6-6, 3.6-12 and 3.6-14 of the FSAR.
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ADVERSE CONDITION #2 INSTRUMENT PANEL BUILDING ELEVATION COORDINATES IN27N156A 1H51P1147 Heater Bay 600 B/01 1N27N156B 1H51P098 Heater Bay 620 C/02 IN27N087A 1H51P098 Heater Bay 620 C/02 IN27N087B 1H51P098 Heater Bay 620 C/02 1N25N263A Heater 6A Heater Bay 600 D/03 IN25N263B Heater 6B Heater Bay 600 D/03 1N25N303A Heater SA Heater Bay 600 D/02 IN25N303B Heater 5B Heater Bay 600 D/02 1N36N030A 1H51P1305
_ Heater Bay 620 B/01 IN36N030B 1H51P1330 Heater Bay 620 B/01 IN36N030C 1H51P1330 Heater Bay 620 B/01 IN26N153A Heater 3A Heater Bay 647 D/03 1N23N153B Heater 3B Heater Bay 647 D/04 1N21N339 Heater 4 Heater Bay 580 C/02
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.3 11 eater Bay High Energy Lines Elevation
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N22 Main, reheat extraction &
580, 600, 620, 647 I
-misc. drains N25 H.P. Htr. drain & vent 580, 600, 620
. 'N27 Feedwater system 580, 600, 620, 647 s
N33 Steam seal system 600, 620, 647 4
N36 Extraction steam system 600. 620, 647 N11 Main steam system 620, 647 P61 Auxiliary steam system 580, 600, 620, 647 N21 Condensate system 600, 620, 647 Refer to -plant' layout drawings, figures 1.2-4 and 1.2-5 of the FSAR, for the Heater. Bay layout.
Note: A break in any of the above lines would not change the conclusions
-described in our. March 14, 1983 letter (Q420.05/#2), which only addressed a feedwater line break in the Heater Bay.
ADVERSE CONDITION #3 i
INSTRUMENT PANEL BUILDING ELEVATION C0 ORDINATES 1R25S002 TP 620 A/04 1R22S003 TP 620 A/04 4
1R23S003 TP 647 3/05 1R23S004 TP
.647 B/06 1R23S006 TP 647 B/06 1R24S034 TP 620 B/06 1R425017 TP 620-C/06 1R42S021 TP 620 C/06 1R42S022 TP 620 C/06 i.
1R42S023 TP 620 C/07 1R14S008
.TP 620 B/06 IC85N001A TB 593 E/16 1C85N001B TB 593 E/1G iC85N002A' 1H51P187 TB 647 G/08
- IC85N0028-1H51P187 TB 647 G/08
~1C85N002C 1H51P773 TB 647 G/08
'TP - Turbine Power Complex TB - Turbine Building
- High Energy Lines - No high energy lines are located within the 620 and 642 levels of the Turbine Power Complex. No further review was necessary
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-to determine high energy lines in the vicinity of the C85 components.
Refer to plant layout drawings, figures 1.2-5 and 1.2-6 of the FSAR.
ADVERSE CONDITION #4 Refer'to' Adverse Condition # 1.
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,b ADVERSE CONDITION #5 Ihi_STRUMENT PANEL ButLDING ELEVATION C00RIDNATES 1R11S004 TP 620 B/04 1R115005 TP 620 B/02 i
1R14S008 TP-620 B/06 1R225002 TP 620 A/04 1R22S003 TP 620 A/04 1R22S004 TP 647 A/05 1R223005 TP 647 A/04 1R23S004 TP 647-B/06 1R23S006 TP 647 B/06 1R22S016 TP 620 A/03 1R235017 TP 647 A/05 1R23S003 TP 647 B/05 1R245034 TP 620 B/06 1R42S017 TP 620 C/06 1R42S021 TP
'620 C/06 1R425023 TP.
620 C/07 IC85N001A TB 593 E/16
-1C85N001B-TB 593 E/16 1C85N002A 1H51P187 TB 647 G/08 1C85N0028 1H51P187 TB 647 G/08
'1C85N002C 1H51P187 TB 647 G/08
- TP - Turbine Power Complex "03 - Turbine Building High Energy Lines-No high energy' lines are located within the 620 and 642 levels of the Turbine Power Coraplex. No further review was necessary to determine high energy lines in the vicinity of the C85 components. Refer C
to plant layout drawings, figures.l.2-5 and 1.2-6 of -The FSAR.
NOT'E: An error was made in the feedwater pump section of Adverse Condition #5, "1R42S002" should be "1R225002."
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Q420.04 Provide a detailed analysis for the turbine trip without bypass event (FSAR Section 15.2.3) in conjunction with a high j
energy line break that causes a loss of feedwater heating (and subsequent increase in reactor power level). Without operator action, the staff.is concerned that this event could lead to a turbine trip without bypass event from a higher power level than previously analyzed.
Response
A detailed review and walkdown of-steam bypass and feedwater heating has determined that there is no single high energy line break (HELB) which could fail both systems.
A high energy line break that would result -
in a loss of feedwater heating is an accident' condition and is evaluated under 10 CFR 100 guidelines. Any HELB that could affect feedwater heating would be bounded by the Chapter 15 analysis for " Steam System Piping Break -
Outside Containment" (FSAR Section 15.6.4). The radiological dose calculation for the event shows that its consequences are bounded by 10 CFR 100 limits (FSAR
-table 15.6-8).
4 The turbine trip without bypass transient has been analyzed unde worst expected transient conditions (FSAR 15.2.3). This analysis is consistent with the requirenents of SRP 15.2.2, including the requirements for initial reactor power level, n
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3 420.05 If used, provide the results of a zone analysis and a plant walkdown. If zone analysis was not used, describe the procedure by which the locations of non-3efety related control system coraponents affected by liELB's were determined.
Response
As stated.iri our March 14, 1983 letter (Question 420.05-approach), CEI identified all major components of
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systems which could significantly. impact reactivity, reactor vessel level, reactor system pressure and decay-1 heat removal.. Once these non-safety related control systen components were identified, locations were l determined from plant layout drawings. Components were then assigned to their respective environmental zones.
All equipment within a single zone would be assumed failed and the impact on system alignments was determined. 'The process was repeated for each i
environmental zone.
Patential system interactions were identified for further evaluatien. A plant walkdown was then performed to determine the location of high energy lines with respect to potentially affected components.
The results of the plant walkdown plus engineering analysis then determined the consequences of each adverse condition. The results of the analysis are contained in the evaluation section, for each adverse condition, in our March 14, 1983 letter. Supplemental information is provided in our September 6,1983 letter and also contained in this letter.
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Q420.06
- Verify that no credit was taken in the analysis for non-safety related equipment. (e.g., feedwater trip on level 8) to L
mitigate the effects.of the HELB and consequential control system failuras.
Response
According to 15.0.1, "only safety grade equipment can be assumed-to mitigate an acciuent and safely shut down the reactor." Our response to 420.04.is bounded by FSAR i
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Chapter 15 accident analyses, which do not take credit for non-safety related equipment.
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-Q4?0.07; Verify that the consequences of the worst case event combinations considered in your analysis are bounded by a small fraction (<10%) of 10 CFR 100 guideline doses.
Response
Consequences of the scenario postulated in 420.04 are bounded by 10 CFR 100 limits. This is consistent with the SRP requirements for analysis of HELBs as defined in
- Branch Technical Position ASB 3-1 Section B.3.a.
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