ML18046A655
| ML18046A655 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 05/18/1981 |
| From: | Vincent R CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | Crutchfield D Office of Nuclear Reactor Regulation |
| References | |
| TASK-15-01, TASK-15-02, TASK-15-03, TASK-15-04, TASK-15-05, TASK-15-06, TASK-15-07, TASK-15-08, TASK-15-09, TASK-15-1, TASK-15-10, TASK-15-12, TASK-15-14, TASK-15-15, TASK-15-17, TASK-15-19, TASK-15-2, TASK-15-3, TASK-15-4, TASK-15-5, TASK-15-6, TASK-15-7, TASK-15-8, TASK-15-9, TASK-RR NUDOCS 8105220064 | |
| Download: ML18046A655 (6) | |
Text
General Offices: 212 West Michigan Avenue, Jackion, Ml.49201 * (517) 788-0550 May 18, 1981 Direc_tor, Nuclear Reactor Regulation Att Mr Dennis M Crutchfield, Chief Operating Projects*Branch No 5 US Nuclear Regulatory Commission Washington, DC 20555 DOCKET 50-255 -
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- --ISSION PALISADES PLANT - SEP TOPICS XV-1, xV-2, XV-3, XV-4, XV-5, xv~6, XV-7, XV-8, XV-9, xv... 10, XV-12, XV-14, XV-15, :XV...;17 AND XV DESIGN BASIS EVENTS By letter dated March 23, 1981, the NRC transmitted for comment a. draft evaluation of. the. Design Basis Events for the Pali sades Plant. "'-We have completed our review of the plant responses discussed in this document and have developed a number of commentsfor your consideration.
These comments.are provided as an attachment to this letter.
Although we have not yet completed our reviews of the radiological con-sequence portions of Topics XV-12, 16, 17, 18, 19 and 20, we d.o not
- expect that our results will be significantly different from the staff's.
It is expected.that these reviews will be complete within approximately another week.
Any comments which are developed from this radiological consequence review.wil.l be transmitted separately at that time.
Robert A Vincent Staff Licensing Engineer CC:
Director, Region III, USNRC NRC Resident Inspector-Palisades
- Attachment ( 5 pages) 81 0 5 2 ~ 0 1bt 'I
PALISADES PLANT-SYSTEMATIC EVALUATION PROGRAM COMMENTS CONCERNING NRCs EVALUATION OF DESIGN
---~-
BASIS EVENTS:
ACCIDENTS AND TRANSIENTS The following are comments to the above report:
- 1.
Table 2 page III-6:
Low steam generator setpoint has a +22 psia uncertainty.
- 2.
Table 2 page III-6:
Auxiliary feedwater is actuated on closure of main feed pump turbine stop valve.
- 3.
Table 3 page III-8:
Assumed power level for steam generator tube failure DBE is 2650 Mwt not 2560.
- 4.
Table 3A page III-9: It should be clarified that the tech spec limit is at HZP which when corrected to HFP is the value assumed in analysis.
- 5.
Sec III.1.2 page III-12 para. 7:
The small-break a,nalysis referenced, CEN-114P (ref 29), is a best estimate analysis.
A more appropriate refer-ence is CEN-137P which is an Appendix K analysis.
Since this.report is not on reference list, add a new reference, reference 45:
"45 Combustion Engineering, *Inc., "Calculative Methods for the CE Small Break LOCA Evaluation Model:, CENPD-137P, August 1974, and CENPD-137P, Supplement 1-P, January, 1977.
- 6.
Sec III. l. 4 page III-15 para 5: It is not evident what the initial increase in primary pressure is due to. It is very likely just an instability in the code.
It has no effect on the final results but may add to confusion.
- 7.
Page III-16 Table:
The reactor trips at 124.7 sec which rounds off to 125 se~ vice 124 ~ec.
- 8.
Page III-17 1st complete paragraph:
In the last sentence of this paragraph the words "III.1.4 of this report" are repeated twice.
Delete these words where they appear first.
- 9.
Page iII-17 Table:
- At 8+ sec the wording is confused: It should read:
- "Steam-generator level increases; power increases to 59%."
At 60 sec power is also asymptotic as well as temperature and flow per XN-77-18 fig 3.45.
- 10.
Page III... 17 Second Last Paragraph:.* Feedwater flow increase event from full power is being further addressed in response to NRC.letter dated July 1, 1980 (RE. SEP Design Basis Events Request for Additional Information).
- 11.
Page III-18 Fourth Full Paragraph:
In the third last sentence specify that HPSI flow reaches core 80 seconds after dump or bypass valves open.
By not specifying the valves it may be misconstrued that the flow reaches core 80 sec after HPSI valves open.
- 12.
- 13.
- 14.
- 15.
- 16.
- 17.
- 18.
- 19.
Page III-18 Fourth Full Paragraph:
In the second last sentence add that the boric ac_i_d _re_i:i-c_hes the core 128 sec after inij!_i~~io_n 9f ___ ev~.Qt _to J;ie __
consistent with table on -pS:ge -rir-18 ~--- -
Page III-19 Last Full Paragraph:
The last sentence is not correct.. The dump valves cannot be hand-jacked closed locally but rather the air supply to the valve operator _can be turned off causi~g the valve to close.
There is, however,.a m~uaj... valve _up.steam of each dump valve that* can be closed.
Page III-20 Second Paragraph of sec*l.5:
The third sentence is incorrect.
According to Tech Spec 3.1. lc, the plant must be in hot standby within 24 hrs. if there are less than four primary coolant pumps operating.
Page III-23 Last Full Paragraph:. In the last sentence the acceptance criteria for pressure is 2750 psia, not 2500 psia (per XN-77-18 sec 3.6 and Tech Spec sec 2.2)
Page III-25 Taqle:
The atmospheric dump valves close 38.3 sec following reactor trip.
Thus, the time from event initiation is 38.3 + 26.7 - 65 sec.
Page III-27 Fourth Paragraph from the Bottom:
The main steam and feedwater piping near the containment penetrations were thickened during original plant construction, not as a result of SR6.
Page III-32 Third Last Paragraph:
The statement made in the second sentence is only true for the 'LPSI system.
For the HPSI system there are two parallel.
valves for. each injection line with each valve on a separate diesel.
- Thus, on loss of any diesel there are always 4 injection points available.
Page III-33 First Full Paragraph:
LPSI valve interlocks are being addressed as part of Topic V-11.B.
The conclusion that interlocks will be required is premature and shouid not be-stated here.
In addition, this concern (WASH-1400 Event V) has been addressed in an Order for Modification of Palisades'_ license dated April 20, 1981.
- 20. *Page III-35 Fifth Paragraph:
In the last sentence it is not clear what MSS system failures are to be considered.
- 21.
Page III-39 Fourth Paragraph of sec 4.1:
Although turbine generator coast-down was designed to provide power to PCP's for 30 seconds, startup testing results showed that the turbine generator coastdown to 80% speed took 80 seconds.
- 22.
Page III-43 Second Paragraph of sec G;l: *In the fourth sentence the comma after "pressurizer" should be deleted to avoid confusion.
- 23.
Page III-43 Second Paragraph of sec G.l:
In the last sentence change the minimum DNB ratio to 1.30 vice 1.45.
- 24.
Page III-48 First Paragraph of sec 7.1: It is not true that a loss of coolant in general leads to a core heat up - only if the core uncovers.
Thus, change "core heatup" to "possible core heatup".
- 25.
Page III-48 Fourth Paragraph of sec 7.1:
Correct the last sentence to
-- -- ---- -_ _ _ __ shoK that tb.e_ prgss\\Jr~ _§.v~_il~"Qle tQ __ QP~I!- ~h~ _c_h~~~ \\l"alve is the stnn of the nitrogen overpressure and the el~vation head of-the-ifr t~k~-:----- ---- *-
- 26.
Page III-49 Fourth Paragraph:
There was no credit taken for charging pumps in LOCA analysis.
- 27.
Page III-49 Fifth Paragraph:
Clarify that without diesel/generator fail-ure that in fact all containment-pressure-suppression systems would function.
- 28.
Page III-50 Second Paragraph:
LPSI flow is through only one shutdown heat exchanger and through only one contairirnent spray line.
- 29.
Page III-50 Second Paragraph:
For PCS pressure qelow 20 psig the LPSI pumps discharge into the cold leg injection points as well as through the containment spray line.
- 30.
Page III-50 Second Paragraph:
In light of the preceding comment, the last paragraph should clarify that only hot-leg suction to LPSI cold-leg injec-tion is used during normal cooldown.
- 31.
Page III-50 Third Paragraph:
The method described in this paragraph is not an "alternate" method but is the preferential method according to plant procedure EOP-8.
Hot leg suction as described in paragraph 2 is the alternate method.
32.* Page III-51 First Paragraph:
A later analysis, CENPD-137P*supp 1-P (ref 45; see comment #5) found the 0.1 ft2 break to be most l~miting with a PCT of 1855°F.
The second most limiting break was the 1. ft-break with a PCT of 1697°F.
- 33.
Page III-51 Second Last Paragraph:
The melting temperature of the fuel is much higher than the melting temperature of the clad.
Also, following a LOCA the temperature profile in the fuel is relatively flat since the core is shutdown.. For this reason the last paragraph should read:
"Clad tern-*
peratures remain below the.acceptance criteria of 10CFR 50.46 during core uncovery.
- 34.
Page III-52 First Full Paragraph:
CEN-114 considere_d opening of PORV' s for LOFW evept only. -.that is, this corrective action w:as not analyzed for breaks.
- 35.
Page III-52 Third Full Paragraph:
The effect of RCP operation on small-break LOCA's was assessed in CEN-115~P. This should be referenced here and therefore must be added to the references in the back:.
"46.
Combustion Engineering, Inc., "Response to NRC IE Bulletin 76-o6c*
Items 2 and 3 for Combustion Engineering* Steam Supply Systems; CEN-115-P, August 1979.
- 36.
Page III.;..52 Third Paragraph:
In order to clarify the third last sentence it should be rewritten to:
"A best estimate analysis was also performed.
In Whl-ch*. -OffSi-:C-e--power w8._s __ a-ssumed--ava.11able---a.na---chUS--CWo-- HP-SIP..,-5 --wer-e------- ----*--- ~----
- 37.
- 38.
- 39.
- 40.
operating.
No operator action was assumed so the RCP's remained on.
Also, no detrimental single failure was assumed.
The results of the**
analysis show that core uncovery does not take place.
Page III-59 First Paragraph of sec 9.1:
The worst transient withregard to primary pressure is a loss of load as analyzed in sec 3.6 of Ref 5 (XN-77-18).
The analysis shows that the peak pressure achieved is 2394 psia wh:l.ch is less than the minimum pressurizer safety valve setting of 2485 psia.
This analysis* conservatively assumed no PORV relief and no pressurizer spray.
Thus, it doesn't seem to follow that PORV isolation during normal operation "could*result in more frequent challenges to the safety valves."
Page III-57 Last Paragraph: *The throat area of a* PORV is 1. 485 in2
(.01 ft 2) and that of a safety valve is 2.545 in2 (.018 ft2).
Thus, 2 the "0.0005 ft2 break" in the parenthesis should be.more like 0.01 ft Page III-63 Top of Page:
Auxiliary pressurizer spray can also be used to reduce PCS pressure for tube rupture event.
Page*III-64 Fourth Paragraph:
Although there have been four instances in inadvertent MSIV closure, only two of these occurred since 1973 * ( 8/31/76,
- 5/22/18).
- 41. ** Page III-65 Third Paragraph:
Due to the size. of.a dump or safety valve
- it is very* unlikely that failure of one of these valves is. worse froni the point of view of.core consequences than a failure of a train of safety injection due to a loss of a diesel/generator.
This latter event is being analyzed in response.to NRC l~tter dated July 1, 1980.
- 42.
Page III-65 Fourth Paragraph:
Explicit analyses of feedwater-line breaks has never. been performed because theywere not part of design basis ci.t the time Palisades was designed.and built.
- 43.
Page III-:65 Fourth Paragraph:
A loss of AC to station auxiliaries is
. less limiting than a 4.;.pwn.p loss of flow event* from the* point' of view of
.the core because the reactor trips immediately r.ather than on low flow.
which would occur later.
- 44.
Page III-66 Second Paragraph:
Should a reanalysis be required, CPC in conjunction with CE.Owners Groups may choose to provide the analysis on a generic. bas.is for the 2560 Mwt class CE plant, vice a plant-specific analysis.. If this is done, suitable justification: of the approach will be provided.
45~
Page III-17 First Paragraph:
Our analysis assume*s a 50% increase in feed water flow as an upper bound. It is believed that the increase in flow which would actually occur with no change in feed water regulating valve position would be less than 50%.
This should be clarified.
""'F'- *- **°;'.=
- 46.
Page III-21 Fi~h Paragraph:
There is no alarm associated with 2% shutdown margin.
- 47.
Page III-22 Last Paragraph:
On a turbine trip signal the feedwater regula-ting valves "fail as is" and the feedwater pump drive turbines ramp.down.
Although this ramp down was designed to go to a speed which would produce 5% full power feedwater flow, in actual practice the flow is much greater.
Following a reactor trip, to prevent a high steam generator level from causing the feedwater regulating valves to close at the high level override setpoint and to prevent unnecessary PCS cooldown, plant operating procedures require that one feedwater pump be manually tripped immediately and the second pump be tripped if PCS temperature approaches 525°F.
Even if the pumps are not tripped, however, the excess flow does not affect core thermal margins, and the overcooling transient can be no more severe than analyzed for the reduction in feedwater enthalpy and excessive feedwater flow events.
In the case of a steam line break, feed pump flow does not have a significant effect because a steam generator low pressure signal causes the feedwater regulating vhlves to be overridden closed and causes the*MSIVs to close which removes steam from the feed pump turbines.
- 48.
Page III-30 Third Paragraph:
Containment.air coolers are slipped from service water, not component cooling.
In addition, CHP signal trips the low horsepower fans leaving the high power fans on rather than changing fan speeds as suggested in the parenthetical expression.
- 49.
PageIII-30 Fourth Paragraph:
See comment 47.
- 50.
Page III-32 First Paragraph:
The modification to provide automatic. feed water regulating valve and bypass valve closures on low steam generator pressure has been completed.
- 51.
Page III-42 Last Paragraph:
Each primary coolant pump has an anti-rotation device which prevents reverse rotation.*.
- 52.
Page III~49 Fourth Paragraph:
According to the FSAR Sec. 8 9nly 2 HPSI pumps are loaded on the diesels (one per each).
Thus, for a loss.of offsitepower, 1 of.only 2 HPSI pumps are available not 1 of 3 as stated.