ML20040F478
| ML20040F478 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 02/03/1982 |
| From: | Mattimoe J SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| To: | Stolz J Office of Nuclear Reactor Regulation |
| References | |
| ALAB-655, NUDOCS 8202090249 | |
| Download: ML20040F478 (58) | |
Text
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e suuo SACRAMENTO MUNICIPAL UTILITY DISTRICT C) 6201 S Street. Box 15830. Sacramento. California 95813; (916) 452-3211 February 3, 1982 9
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-9 DIRECTOR OF NUCLEAR REACTOR REGULATION
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L% /p ATTENTION JOHN F STOLZ CHIEF OPERATING REACTORS BRANCH 4' C
sh C U S NUCLEAR REGULATORY COMMISSION M.Jpg\\ ' #
WASHINGTON DC 20555 DOCKET 50-312 RANCHO SECO NUCLEAR GENERATING STATION UNIT NO 1 HIGH PRESSURE INJECTION N0ZZLE CYCLES On December 11, 1981, the District responded to the Atomic Safety and Licensing Appeal Board's Memorandum and Order, ALAB-655. Since that time, we have had numerous discussions with your staff concerning the allowable number of transient cycles for the high pressure injection nozzles. We have provided the Babcock & Wilcox Company Field Change Package, "HPI Nozzle",
Document No. 04-3370-00 FC-0174-00, for your review, and following additional questioning, B&W Calculation Nos. 32-1121811-00, "HPI Nozzle Usage Factor" and 32-1119809-01, "HPI Nozzle Usage Factor". These documents form the basis for including within the total number of allowable reactor trips (400) a specific provision for up to 70 actuations of high pressure injection following such a trip.
(These 70 cycles are in addition to the 80 cycles included in the original stress analysis.) As noted on page 4 of Document 32-1121811-00, the original 80 cycles resulted in a usage factor of only.44, which demonstrates the conservation in the analysis. These documents are attached to this letter for your reference.
As discussed in our response to ALAB-655, these analyses do not explicitly address the Appeal Board's question concerning the ultimate number of cycles these nozzles are physically capable of safely withstanding. Our response provided the justification for not responding to this question. We feel, however, that these additional cycles demonstrate that safety limits are not being approached, and that no safety concern exists due to cycle usage at Rancho Seco Unit No. 1.
Your staff has requested additional information for t
their use in reviewing these calculations. The workload within the Babcock &
Wilcox Company (due to such factors as the Thermal Shock issue) precludes this information being generated and submitted for your review prior to July 1, 1982.
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D JOHN F STOLZ CIIIEF Page 2 February 3, 1982 This information relates primarily to techniques used in the original Rancho Seco Stress Analysis and we feel this information is irrelevant to the Appeal Board's concern over high pressure injection nozzle usage, but should satisfy your needs in reviewing this change to the Rancho Seco Administrative Procedures.
We expect to receive a written request for information from you which will confirm the information to be supplied in July of this year.
L%n John J. Mattimoe Assistant General Manager and Chief Engineer Attachments cc:
Mr. Tom Baxter - Shaw, Pittman, Potts & Trowbridge Mr. Dave llolt - Babcock & Wilcox Company Mr. Mark Padovan - Nuclear Regulatory Commission
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SHEET 0F TYPE DOC Field Change Package Idr. R.J. Rodriguez TTN 2 pkgs.
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HPI No::le 50 001 001 DESCRIPTION 'io JU3TiftCATION OF Change:
It has been reported that the SMUD operators have been starting a second make-up pu=p following each reactor trip to obtain additional make-up flow.
This wts done to prevent less cf indicated prescuricer level during the transient.
All four of the high pressure injection no::les have been used for these occurrences cnd the three no::les which are not continuously adding make-up flow receive a thermal shock from the cold EWST water.
The contract Fun :ticnal Specification, CS (F) 3-92/USS-ll/0372, defines forty cycles of high pressure injection actuation and forty cycles of high pressure injection testing.
At SMUD's present usage rate, 31 cycles, the nc= les would excaed their allowable cycles before the end of their forty-year life.
B&W has perfor.ed a HPI no =le stress analysis evaluation of the actual experienced transients (actual transients are less severe than functional specificatien transients) to determine their actual fatique life expectancy.
In reviewing the SMUD HPI Design Analysis, it was discoveredNhat the original analysis was performed as a generic analysis and the faticue analysis was perfor.ed te a later, more restrictive design code.
In analyzing the HPI no::le to the contractual design code, 1968 draft USAS, B31.7 with June 1963 errata, the number of allcwable HPI actuatiens can be approximately doubled.
The re-analysis must conform to the requirements of ASME Section XI which now applies to the site cnd allows the use of earlier codes.
This Field Change Authorization is being written to document the increase in allowable HPI systen actuation cycles and therefore no hardware changes are required.
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REF DOCUMENT no.
REFERENCE DOCUMENT TITLE CS(F)3-37/NSS-ll/037 Specification for Reactor Coolant System Piping SNEET I
0F l DESCRIPTION OF CHANGE:
1.
Revise paragrapli 1.1 of Appendix 1 of the reference document to read as follows:
1.1 General Functional Specification for Reactor Coolant System Components, Specification No. CS(F)-3-92/NSS-11/0672 as modified by Field Change Authorization 04-3370-00.
2.
Add paragraph 2.2 in Appendix 1 of the reference document to read as follows:
2.2 ASME Eoiler and Pressure Vessel Code a.
Section XI, Inservice Inspection 1974 Edition with SunTner The reanalysis of the HPI nozzle 1975 Addenda is to be in accordance with the requirements of the code listed above.
The changes to the reference document above are certified to be correct and complete and in compliance with the requirements of ASME Boiler and Pressure Vessel Code,Section XI.
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Title 4 di 8-8 Surge Line Te=perature and Flov.
8-9 Makeup and Spray Flow Rate and Te=perature.
8-10 Reactor Trip, Cooldown and Nor=al Power Recovery - Primary Syste.
8-11 Reactor Trip, Cooldown a=d No:=al Recovery - Secondary System.
8-12 Reactor Trip frc= Full Power for Trans. 8[&.S - Stea= Gen. Conditions.
2 8-12A Feedvater Conditions Following Reactor Trip (8A&B).
8-13 Less or Feeevater Resetor Trir-Stes: Feedvater Srste-
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9-1 Reactor Coolant Conditions.
9-2 Surge Flov and Te=perature.
9-3 Makeup and Spray Flov and Te=perature.
9-h Feedvater and Reactor Coolant Flow Range.
9-5 Stea= Te=perature and Test.
9-6 Stea= Generator Pressure.
9-7 Feedvater Temperature Range.
Transient No. 10 - Chanre of Flov 10-1 Reactor Coolant Te=perature.
10-2 Surge Line Water Temperature and Flow Rate.
10-3 kakeup and Spray Line Flov and Water Te=perature.
10 h W -Steam System Para =eters for Stea= Generator with Two Reactor Coolant Pu=ps.
10-5 W-Stea= Syste= Para =eters for Stea: Generator with one Pu=p.
L Transient No. 11 - Rod Withdrawal Acciden:
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11-1 Reactor Coolant Te=perature and Pressurizer Liquid Level and Pressure.
i 11-2 Spray and Makeup Te=perature and Flow Rate.
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! ISSUE DATE: f 72 PAGE vii SPECIFIC ATION NO. cS'F)-3-92 NSS-11 m /, m Babcock & Wilcox
e OY-3370-co 34 in Transient 3 (Plant loading 8% to 100%). These return to 2
Power events are included in thelS000 cycles specified for Transient 3 2) 56 Reactor trip events, folleved by cooldovn as described in Transient 13 (Cooldown), folleved by heatup as described in
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Transient 1A (heatup) and Transient 3 (Plar.t Loading). These 56 heatup and cooldovn e7eles folleving reactor trip are in-eluded in the 2k0 cycles specified for Transien 1.
- %g, 6.9.3 Transient Data jff/
6.9.3.1 Reaeter Coolant system I'
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Reacter Coolant System conditions for the'reacter trip transients are presented in the following figures:
Tyne A Reactor Trie - Figures S-1, S-1A, 8-2.S-3 Tyre 9 aeacter Trt
- Figures 8 L, 6 LA, 5-5, 8-6, S-10
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h c.9.3.2 Steam Generaters
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are presented in Figures 3-11, S-12, 8-12A and 8-13.
6.10 Transient 0 - Rarid Derressuri:stier (Upset Condition) 2 6.10.1 General Descrirtion Rapid depretsurization is a short ter=, racid cocling cf the Reactor Ccolant Syste= by the steam generaters in order to re-duce the reactor coolant syste= pressure to a value less *.han the design pressure (1065 psia) of the steam generaters within 15 minutes. The objective of the rapid depressurizatien :s to isolate a tube leak.
The initial conditions at the start of the transient are assu=ed to be hot standby with decay heat re= oval by the stes= generaters dumping stea= to the condenser.
The turbine bypass 0:ntrol pres-sure is assc=ed to be 1050 psia. This gives an average reaeter 2
coolant syste= temperature of acout 553.7 7.
s When average reactor coolant te=perature has been reduced to 500*F, and reactor coolant pressure is equal tc or less than 1065 psia (15 =inutes), nor=al cocidovn as described in Transient 1B is initiated.
6.10.2 Cycles The number pf rapid depressurization events for design chall be 40.
A cc=plete cycle consists of a) power raduction fr:= 100" 3
power as descriced in Tran:1ent h (Plant Unicading); b) Rspin de-BUE D A TE: ' 7?
PAGE 16 SPECIFIC ATICM M0.
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Cycles Nos.
IA Heatup frc= 70'F to 8% Full Power (Nor_-al) 2h0 1A 1A-1 2
13 Cooldown fres 8% Pull Power (Norn')
2h0 13 -l 6 2
Power chary;e O to 15% and 15 to 0%
( Nor=al) 1440 2A 2A 5 2
23 23-I 3
Power Leading 8% to 1005 power
( Nor=al) 18,000 3 3-5 2
h Power Unloading 100% to 8% power (Normal) 18,000 L_1 _ L_3 5
10% Step Lead Increase (Normal) 8,000 5 5 4 6,
101 Step Load Decrease (Normal) 8,000 6 6-5 7
Step Load Reduction (100% to 8% Pover)(Upset)
Resulting frc= turbine trip 160 7 7-5 Resulting frc= electrie.a1 load rejection 150 Total 310 8
Reactor Trip (Upset)
Type A k0 8 8-13 Type B 160 Type c 88 Trips included in transient numbers 11, 15
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2 12 Hydrotests (Test) 35 13 SteadyM tate Power Variations (Normal) )
13-1 14 Control Rod Drop (Upset) k0 14-1 -lL L 15 Loss of Station Power (Upset) k0 15-1 7 16 Steam Line Failure (Faulted) 1 15-1 -lE-3 17A Loss of Feedvater to One Steam Generator (Upset) 20 17A 17Ac 17B Stuck open Turbine Sypass Valve (F.=ergency )
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NONE STRESS REPORT FOR PRIMARY PIPING (620-0011-50)
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DEsCRI3 TION OF CHANGE:
l'he following documents are added as addenda to the original stress report:
32-1119809-01 HPI Nozzle Usage Factor 32-1121811-00 HPI Nozzle Usage Factor I certify that the two (2) above mentioned analysis when taken as a whole justify the modifications to be made under the authority of FCA No. 04-3370-00 and that these two (2) analysis when considered as a whole meet the require-ments imposed by the applicable Ecuipment Specificatien, as modified by FCA No. 04-3370-00.
Attested to this date:
October 10, 1980 g.ALTH g qC p
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h O GEORGE J. VAMES
'2-U CERTIFICATE NO.
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' TITLE Senior Enqineer DATE 10/0/80 TITLE Supervisory Enoineer DATE,10/9/80 Pl l PURNSl;;
A To detemine if the Ho! nozzle can withstand 70 HP! actuations following a reacter trip condition. These will be in addition to the 40 rapid
[.a depressurization and 40 test transient cycles specified in the contract
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functional specification CS(F)3-32/NSS-ll/0372.
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The HPI Nozzle can withstand 40 test transient cycles, 40 rapid depressurization I
transients and 70 reactor trip transients as described in FCA 04-3370-00.
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Thepurposeofthiscakculation~istoamendCDS 32-1119809-00 Ref (1).
Ref (1) justified an additional 70 cycles of rapid depressurization
%j transient. This calculation will justify 70 cycles of HPI actuation f
fo'. lowing a reactor trip instead of 70 additional rapid depressurization h
transient cycles was done in Ref (1),
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CDS 32-1119809-00, Ref. (1), detennined the maximum allowable number of Q
rapid deoressurization transient cycles. This calculation was based f) upon the assurption that FCA 04-3370-00 would characterize the 70 additianal PPI transients as required by SPR #13-11-361-00, Ref. (5) as rapid depressurization transients.
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Khen FCA 04-3370-00 was issued the additional 70 HPI transients were E
assumed to follow a reactor trip transient. Thus, this calculation i
will show that the HP! can withstand 70 cycles of HP1 Actuation following a reactor trip transient instead of 70 additional rapid i
depressurization cycles as shown in Ref. (1).
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WNP-20210-1 (1-78) lIN4* fn %.{ CAI.Clfl.ATION DATA / TRANSMITTAL S11EET W 01 I[h cAtc. 32 - 1119809 DOCl*iFNr IDFNTIFIER ~ ({[ I 1 TRANS. 86 - l TYPE: _arsrecx a ervsmer _sArriv AsA:.vsts nrent _rx. sur. Ivrt r __etstcu m?n. Lei. stew vtttr. W[ y ~orag, h TITI.E HPI Nozzle Usage Factor [,[c,/.a REVIEVED BY i PREPARI;D 11Y i TITLE Senior Enoineer DATE 10/9/80 TITLE Supervisory Engineer ox7g 10/9/80 pd I PURPOSE: kI l This calculation is being revised to show that it is not the controlling D stress document for SPR #13-11-361-00 anymore. i This documnt is now amended by CDS #32-1121811 -00 J h li _ -.. = { I StM!ARY OF RESULTS (INCLUDE DOC. ID'S OF PREVIOUS TRANSMITTALS & SOURCE CALCUl.ATIONAL i f PACKAGES FOR TilIS TRANSMITTAL) ) k i See CDS-32-1121811-00 and CDS-32-1119809-00. [ T n Y' I 3 h e h f, DISTRIEUTION y Page _]_ of M
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