ML20024G644
| ML20024G644 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 12/05/1973 |
| From: | Mayer L NORTHERN STATES POWER CO. |
| To: | Skovholt D US ATOMIC ENERGY COMMISSION (AEC) |
| References | |
| NUDOCS 9102140354 | |
| Download: ML20024G644 (12) | |
Text
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Regulatory Go Cy.
NSED NORTHERN STATE 5 POWER COMPANY M I N N E A P O L.t e, M I N N E S OT A 945401 j' 'ML/
December 5, 1973 g
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4 Mr. D J Skovholt Oggg C
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Assistant Director for h
b Operating Reactors y
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Directorate of Licensing f
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S Office of Regulation 9
U S Atomic Energy Commission
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Washington, DC 20545
Dear Mr. Skovholt:
MONTICELIO NUCLEAR GENERATING PIANT Docket No. SC 263 License No. DPR-22 Supplemental Information on EOC Transient Analysis Requested in October 18, 1973 Letter Your letter of October 18, 1973, asked that certain supportive information concerning end of cycle 2 transient analyses te submitted prior to achieving an exposure increment of 2680 WD/STU. We expect tu reach that threshold about December 7, 1973. Your letter asks for the identification and justification i
of changes made in the calculational assumptions used in performing transient l
analyses. There have been four major transient analysis submittals to date.
(See References 1 through 4).
For convenience, the parameters and calcul-ational assumptions that have changed are presented in the attached tables with short explanations as appropriate.
Table I sucinarizes the Safety / Relief Valve Sizing Transient Analysis while Table II sunanarizes the Safety Valve j
Sizing Event Analysis (Including Failure of Direct Scram). A discussion of I
the various points in question follows.
1 Limiting Transient or Event The turbine trip without bypass (TT w/o BP) has been the limiting transient for relief valve sizing throughout the four major analyses.
l The TT w/o BP with failure of direct scram was the limiting event at the 10C-1 for safety valve sizing.
For subsequent analyses the limiting event became the main steamline isolation valve closure with failure of direct scram.
As discussed in Reference 2, this change resulted from the combined ef fects of the available steam space along with the change in rate of pressur-itation for modified scram react.4 vity curves.
9102140354 731205 PDR ADDCK 05000263 P
PDR 8752
NoHTHERN CTATED POWER COMPANY Mr. D J Skovholt Number of Valves Operable The BOC-1 case assumed three of the four relief valves were operable for the relief valve nizing transient.
For the EOC-1 analysis, the Technical Specifications were changed (Refe-;uce 5) requiring all four relief valves to be operable.
From that time on, all four valves were assumed operable during a transient.
Reference 3 discusses the history of vessel ever-pressure protection design, the number of valves installed in excess of those required and justification for taking credit for additional valves in the transient analysis.
The cycle 1 analyses of safety valve sizing events were based on three relief valves and two safety valves being operable. For the reasons discussed in Reference 3, subsequent analyses took credit for four safety valves and four relief valves for the majority of the analyses. Additional analyses for ASME code requirements were performed showing over-pressure protection for the various combinations of operable valves tabulated. A margin of 25 psi or more from the vessel design over-pressure limit was calculated in each case.
Safety Valve Setpoint Used in Analysis The acceptance criteria of the relief valve sizing transient analysis is that a 25 psi margin exists between peak reactor vessel pressure and the lowest safety valve setpoint.
The lowest safety valve setpoint of 1210 psi was used for cycle 1 analyses.
Reference 3 requested that the four safety valve setpoints allowed by the Technical Specifications be raised from 1210 and 1220 psib to 1240 psig; that value was therefore used for subsequent analyses.
For the safety valve sizing event, the valves are assumed to open to keep the vessel from exceeding its 1esign over-pressure Itmit.
The analysis for BOC-1 used the nominal setpoints.
Subsequent analyseo added a measure of conservatism by assuming a 1% deviation from the nominal setpoint.
Relief Valve Setpoint The same relief valve setpoints and relief capacity models apply to both the relief valve sizing transient and the safety valve sizing event. The Technical Specifications have always required that the setpoints of all valves be less than or equal to 1080 psig.
The reactor kinetics model used in calculating pressurization transients allows for a sLmulated spread in setpoints.
The BOC-1 analyses assumed three valves opened at 1080, 1085, and 1090 psig re-spectively. This allowed for a nominal sctpoint deviation in the undesired direction. (While the FSAR states a fourth setpoint modeled at 1095 psig, the analysis was done taking credit for opening of unly the first three valves.)
Reference 6 states that the relief valves are set at 1070 psig when cold to provide assurance of lifting at 1080 psig during hot operating conditions.
The model was nodified allowing a 1% deviation plus a nominal deviation for subsequent analyses. The model assumed one third of the relieving capacity l
at 1081 psig (1070 + 1%), another third at 1086 psig and the remaining third
NORTHERN OTATED POWER COMPANY Mr. D J Skovholt at 1091 psig.
This was a 1 psi shift in the conservative direction from those setpoints used in the FSAR analysis.
In assuming four valves open rather than three, the total valve capacity is still modeled in three segments from approximately the nominal setpoint to 1% above the nominal setpoint, each segment representing the capacity of 1 1/3 valves, in the course of the EOC-2 analysis, a conservative change was made in the nodel by representing all relief valve setpoints at 1091 psig (1080 + 1%).
Reference 4 shows that while this series of changes is in the conservative direction, the reported change in the relief valve setpoint model resulted in only a 3 psi change in peak transient vessel pressure.
Relief Valve Delay Time Earlier this year, it was observed that the delay in the initial opening of the relief valves was longer than initially assumed. The observed time was used in the subsequent analysis.
The cause of the longer than expected delcy has been identified and the valves have been modified accordingly.
Tests of modified valves show the delay in initial opening time to be within the 0.4 second in-terval used in the most recent analysis. This topic is thoroughly discussed in References 3, 4, 7, and 8.
Scram Times Reference 2 requested that the Technical Specifications be changed to require a faster scram time.
The change was subsequer.tly granted; at all times proposed or existing Technical Specification scram times were used 16 the analyses.
Scram Reactivity Curve The DOC-1 analysis was performed using what is termed the Generic A scram reactivity curve. When it was realized that exposure has a marked ef fect on the curve, the Generic B curve was developed. The.1 curve applied to the EOC-1 as well as a significant portion of cycle 2.
Reference 2 showed that transients are acceptable when the scram reactivity availab'e is greater than or equal to the B curve.
The question then became over W at portion of cyclo 2 the B curve was applicable.
This was done in three stages:
- 1) Reference 5 presented 2250 MWD /T as the exposure increment of cycle 2 in which the B curve would not bi exceeded.
This was a beginning of cycle estimate with the i'itention of being refined at a later date.
- 2) Reference 3 reported the threshold to which the E scram reactivity curve applied to be 2400 MWD /T.
This was based upon a generic scram reactivity curve / excess reactivity correlation study of another reactor which was applicabic for Monticello.
The result of the study was that scram reactivity degradation was primarily a function of excess reactivity, or control density requirei to compensate for excess reactivity. Excess reactivity calculations for Monticello were performed stad an exposure threshold was determined.
l
NORTHERN OTATES POWER COMPANY Mr. D J Skovholt 3)
Reference 4 states that the B curve corresponds to 2680 H4D/T in cycle 2.
This was determined by making actual scram reactivity calculations for the as-loaded Monticello core over the cycle an( finding the exposure point at which the Monticello scram and the B scram curves were equivalent, in terms of transient analysis.
While it may appear that the threshold to which the B curve applies is sensitive to exposure and therefore shif ting, the changes are the rest *'
of more accurate methods being used to home in on the exact threshold 6 is approached.
The C1 curve was originally used as the EOC-2 Monticello scram reactivity curve.
This curve was used for design purposes and was based on the " reference core" shown in Reference 9.
As reported in Reference 10, there were some slight changes made to the "ref erence core" to allow for greater cycle 2 exposure.
The af fect of thete changes on the C1 curve was known to be small and therefore not calculated exactly until doing the EOC-2 analysis.
(Reference 4.)
The calculation of s'eram reactivity at points within the cycle as well as end of cycle is based on an operating history consistent with the Haling power shape.
Because scram reactivity is somewhat power shape dependent, the C2 curve will be obtained only with a Haling power shape at all rods out, end of cycle.
If the target exposure shape is not met the actual core scram reactivity will differ from the C2 curse. To date most reactor experience has been that at EOC the axial power shape is peaked somewhat more strongly at the bottom of the core than the Haling power shape. This should enhance the actual scram reactivity response slightly.
If a core is operated in a anner such that the power peak is shif ted more to the top of the core, it is possible that the actual core scram reactivity will be slightly below the end of cycle curve.
As discussed below the transient analysis calculations apply a conservative multiplier to the scram reactivity curve. Also there is a minimum of 25 psi margin for the overall transient.
Coupled together there is ample conservatism to offset any conceptual loss in scram reactivity response below the design basis curve due to operating history. The power shape in the Monticello reactor has been maintained very near the Haling shape. This fact, along with the conservatisms in the calculations,tuake us confident that the scram reactivity curves used in the EOC-2 analysis are applicable.
Recirculation Pump Trip The recire pumps draw their power from the auxiliary transformer during normal operatina. The design of the Monticello plant includes a f ast transfer of cuxiliary loads to the reserve transformer on a turbine trip.
If the fast transfer fails, a backup transfer is initiated by low voltage relays on the
NORTHERN STATES POWER COMPANY Mr. D J Skovholt auxiliary bus.
If the recirc pumps were not tripped with a turbine trip, their momentum would generate voltage out of phase with the reserve trans-former, resulting in a large current surge on transfer to the reserve transformer and possibly a failure of equipment.
To prevent this, the recire pumps are tripped automatically with a turbine trip before attempting to make the fast transfer.
During an MSIV closure event, there Is no immediate turbine trip.
The generator remains connected to the grid and begins to motor at synchronous speed. This condition exists for approximately 20 seconds at which time protective devices initiate a turbine trip.
During the 8 to 10 seconds of interest in the MSIV closure event, the recire pumps receive a near-normal source of pvver from the auxiliary transformer.
The tripping of recire pumps increases the peak transient vessel pressure somewhat. The TT w/o BP reported in the DOC-1 analysis did not acknowledge the load-shedding feature of a turbine trip and therefore did not assume the recirc pump trip.
Likewise, the EOC-1 analysis was done without the pump trip; the core flow time response appears significantly different, though, because this analysis was done assuming automatic flow control. Since the load demand is set to zero on a turbine trip, a recirc pump runback occurs in this mode.
The flow is not affected by the runback scheme for the first four seconds the change in core flow due to the runback after that time has a negligible effect on the peak vessel pressure.
The recire pump trip feature has since been added to the model for all turbine trips.
(See References 3 and 4.)
As stated above, power is available to the recire pumps during the initial portion of the MSIV closure event. Table 11 shows that the pumps were correctly assumed not to trip during these events.
It should be noted that Figures 5, 6,10, and 15 of Reference 4, are labeled incorrectly, stating that the pumps were tripped.
In these cases, it can be seen that core mass flow actually increases by about 10% as the reactor pressure increases and voids collapse.
Use of Design and Onerational Conservatism Factors (DCF/OCF)
There are three parameters affecting the Monticello analyses to which con-servative multipliers were app *. ed as follows:
Parameter DCF OCF Scram Reactivity 0.8 0.95 Doppler Reactivity 0.9 0.9 Void Reactivity 1.25 1.15 i
l l
l l
I
NORTHERN STATES POWER COMPANY 6-Mr. D J Skovholt All transient analyses for Monticello have shown compliance to applicable s
limiting criteria using the DCr.
The only analyses reported in uhich OCF were used are in Reference 4.
The af fects of the two sets of conservative multipliers were presented to indicate the margin of safety they contribute to the calculation.
h.
We understand L.+t your representativec have recently met with Gent.ral Electric persor.nel to discuss this subject.
The philosophy on the use of DCF and OCF is as follows.
Analysis of a piant in the design phase using a mathematical model em Joying design data must a ntider uncertainties associated with the model
,d design data and contingencies associated with design characteristics and feat e es.
Evaluations of n operating plant need only consider uncertainties connects 2 ;uh the model and as-built plant data.
Consequently, the margin inplicit in DCF should logically be larger than in OCF.
By this scheme, Monticello could use the OCF whereas we presently show compliance to limits based on the more conservative. DCF.
Nature of Failure Assumed in MSIV Closure with Failu g of Direct Scram The assmaption in the safety valve sizing event is that there is a f ailure of the direct serem on MSIV pssition; an indirect scran. is assumed to be initiated by hi-hi neutron flux. The MSIV closure inputs to the reactor protection system are from valve stem position switches mounted on the eight MS1V's.
Each ot the switches is desigred to open before the valve is more than 107. closed. The logic is arran ed so that any two main steam lines can be isolated (eoth inboard and outboard valves), but when a third line is isolated, a scram occurs.
The reactor protection system is a one out of two taken twice logic. A scram occurs on de-energization of the (Al or A2) and (B1 or B2) subchannels.
Failure to scram therefore occurs when the (Al and A2) or (B1 and B2) channels remain energized. The attached Figure 1 shows in simplified form, the arrangement of MSIV position uvitches in the scrnm 1,gic.
The 2-80 A throug" D and 2-8.5 A through D switches are the inboaro and outboard MSIV Fp.
position rwitches, respectively, for the four main steam lines.
Through k
additional curcuitry, the Al subchannel is said to be energized when either the A or E relay is energized and so forth.
E Suppose the failure to scram involves ii.ilures causing the Al and A'. sub-channels to remain energized during an MEIV closure.
This would require the (A or E) and (C or G) relays to remain energized.
For the A relay to remain energized, both position switches 2-86A and 2-60A must fail to open as yr
'esigned.
Likewise, for the C relay to rema'u energized, switches 2-92 and 2-80:: must fail to open. Failure of the direct scram in the MSIV closure event can therefore occur only when specific combinations of four or more of the eight valve position swltches fail in either F A or B channel; a very unlikely situation.
5 P
e NORTNERN OTATEO POWER COMPANY Mr. D J Skovholt Fuel Response to pres orization Transients Fuci damage during transients is analyzed to assure that perforation of the cladding from overheating or excessive cladding strain is prevented.
The damage limit for the former is when MCHFR reaches 1.0 based on the Henchy-Levy correlation and for the latter is a MLHGR of approximstely 28 kv/ft.
(See Reference 9).
During pressurization transients there has always been a wide margin between these limits and the esiculated MCHFR and MLHGR values.
The peak fuel center temperatura change is plotted in Reference 2, 3, and 4.
The 0 and 1007. points correspond to saturation temperature and the steady state temperature for rated power operation at 17.5 kw/ft, respectively.
The Mtter is approximately 4300 F.
Reference 11 discusses the fuel thermal model.
(The peak fuel center temperature plots in References 2, 3, and 4 g
used an improper scaling factor; each curve should be increased by a 1.04 multiplier).
Entbnipy limits are only applied to prompt critical transients (i.e. Rod Drop Accident). Limits for abnormal transients are MCHFR greater than 1.0 and MLCHR such that the 1% plastic strain limit is not exceeded. If the transient analyses show that these are not exceeded, fuel damage is not expected to occur. In the transient r.nalyres discussed above, MCHFR and MLHCR are well wi;hin these limits.
Yours very truly, 9..
L 0 Mayer, PE Director of Nuclerl: Support Services IDM/MHV/1h sc:
J G Keppler G Charnoff MPCA - Attention K. Dzugan Telecopied to AEC-DL, December 5,1973 lb
RE_F.ERENCES 1.
tbnticello Nucicar Generating Plant, PS AR, Docket No. 50-263 2.
Supplemental Report of a Change in the Transient Analysis as Described in the FSAR, L 0 Mayer to A Ciambusso, February 13, 1973.
3.
Change Requent Dated September 13,1973, L 0 Mayer to J F O' Leary, September 13, 1973.
4 Response to October 2,1973 Letter Requesting EOC Transient Analysis, L 0 Mayer to D J Skovholt, October 10, 1973.
5.
Change Pequest Dated June 1,1973, L 0 Mayer to J F O' Leary, June 1,1973.
6.
Safety / Relief Valve Settings Exceeding 1080 Paig, R 0 Duncanson to P A Morris, April 30, 1971.
- i. Observed Relief Valve Opening Times Different Than Those Assumed in the Transient Analysis, L 0 Mayer no J F O' Leary, August 1,1973.
8.
Planned Reactor Operation from 2000 NWD/T to the End of Cycle 2, L 0 Mayer to J F O' Leary, August 21, 1972.
3 9.
Request for Authorization to Operste With Reload Fuel in the Core, L 0 Mayer to A Ciambusso, February 20, 1973.
10.
Supplementary Information Regarding the First Monticello Reload, L 0 Mayer to J F O' Leary, April 13, 1973.
11.
Analytical Methods of Plant Transient Evaluations for the GE/BRR, Topical Report NEDO-10802, February,1973.
j
6 TABLE II l
SAFETY VALVE SIZING EVENT ANALYSIS (INCLUDING FAILURE OF DIRECT SCRAM)
B Transient Analysis FSAR 2/13/73 9/13/73 10/10/73 Calculctional BOC-1 E00-1 S V Set Point Change EOC-2 IA :umption (Reference 1)
(Reference 2)
(Reference 3)
(Reference 4)
Limiting Event TT w/o BP MSIV Closure MSIV Closure MSI7 Closure I No. of Safety / Relief 3 RV 3 RV 4 RV 4 RV Vs1vac cnd Safety Valves 2 Sy 2 Sy 4 py 4 3y A;cumed Operable 25 psi margin exists 25 psi margin exists l Alternate Combinations of with RV/SV combinations:
with ff RV/SV l Scfety Relief and Safety t/1, 4/2, 3/3, 3/4 combinations: 4/0, 3/1.
, Valves Available 3/2. 2/3, 2/4 Safety Valve Set Point 2-1210 psig 2-1210 + 1% psig 40 1240 + 1% psig 4@ 1240 + 17. psig Uscd for Analysis 2-1220 psig 2-1220 + 1% psig (T S Change allowed raising set point)
Ralisf Valve Set Point 1080 psig 1080 psig 1080 psig 1080 psig 1080 psig (Nominal)
Ralief Valve Set Point 1/3 at 1080 p sig 1/3 at 1081 psig 1/3 at 1081 psig 1/3 at 1081 All 4 at 1/3 at 1085 psig 1/3 at 1086 psig 1/3 at 1086 psig 1/3 at 1086 1091 1/3 at 1090 psig 1/3 at 1091 psig 1/3 at 1091 psig 1/3 at 1091 psig Modal R311ef Valve Delay Time 0.2 sec 0.2 sec 0.8 see (round valves 0.4 sec (Modified valves) not to respond as predicted)
'65
'67A
'67A
'67A Scram Time (Changed T S requirements)
Scram R:activ8ty Curve A
B B to 2400 MWD /T; C1 B to 2680 MWD /T; C2 to (Acknowledged change to EOC (Identified EOC (Recalculated ceram due to exposure) limiting exposure reactivity curves) thresholds)
No No (Load shedding No No - Figures 5, 6, 10 Racirc Pump Trip done during TT w/o and 15 are labeled BP does not occur incorrectly on MSIV Closure)
Con 2ervstive Multipliers DCF DCF DCF DCF (OCF used to show comparison; no credit taken) v
I t
[
TABLE I SAFETY / RELIEF VALVE SIZING TRANSIENT A'IALYSIS 1
Transient nalysis FSAR 2/13/73 9/13/73 10/10/73 Calculational BOC-1 EOC-1 S V Set Point EOC-2 Assumption (Reference 1)
(Reference 2)
Change (Reference 4)
(Reference 3) l Limiting Transient TT w/o BP TT w/o BP TT w/o BP TT w/o BP l
1 3
4 l
No.~ cf Relief Valves Asrumed to be Operable 3
(T S Change re-4 4
j 1
quired the fourth l
j RV to be operable) j t
1240 psig Safety Valve Set Point Used for Analysis 1210 psig
.1110 psig (T S Change allowed 1240 psig raising set point)
Ralief Valve Set Point 1080 psig 1080 psig 1080 psig 1080 psig 1080 psig
]__Jposinal) l Relief Valve Set Point 1/3 at 1080 1/3 at 1081 1/3 at 1081 1/3 at 1081 all 4 f
Model 1/3 at 1085 1/3 at 1086 1/3 atq1086 1/3 at 1086 at 1091 i
1/3 at 1090 1/3 at 1091 1/3 at 1091 1/3 at 1091 l
Relief Valve Delay Time 0.2 sec 0.2 sec 0.8 sec (Found 0.4 see (Modified valves)
{
1 valves not to respond as pre-g dicted)
Scram Time
'65
'67A l
(Changed T S
'67A
'67A l
requirement) i l
Scram Reactivity Curve A
B e B to 2400 NWD/T; B to 2680; C2 to EOC j
(Acknowledged chang due to exposure)
C1 to EOC (Iden-(Recalculated scram tified.1Lmiting
' reactivity curves) l f
exposure thresholds) uo (no, assumed auto yes yes Racirc Puap Trip flow control with i
puinp runback) l Conservative Raltipliera DCF DCF DCF DCF (OCF used to show comparis6n; no credit taken) f n
rt v.r, e
m-s -
+ - - -,,
i Al A2 l
2-86A 2-86B 2-86C 2-86D 2-80A 2-80B 2-80C 2-80D i
l A
E C
G 1
B1 B2 2-86A 2-86C 2-86B 2-86D 2-80A 2 30C 2-80B 2-80D J
B F
D H
FIGURE I Simplified Arrangement of MSIV Position Switches in Reactor Protection System s
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AEC D7STRIBUTION FOR PART 50 DOCKET MPERI AL (TEMPORARY FORM)
CONTROL NO: 8752 FILE:
PROM:
DATE OF DOC DATE REC'D LTI-MEMO RPT OTHER Northern States Power Company a
Minneapolis, Minnesota 55401 L. O. Mayer 12-5-73 12-8-73 X
TO:
SENT LOCAL PLR X
Mr. h vholt No Oric CLASS UNCLASS PROP INFO INPUT NO CTS REC'D DOCKET NO:
\\;
XXX 39 50-263 DESCRIPTION:
ENCLOSURES:
Ltr re our 10-18-73 ltr..... furnishing requested Suppl info on EOC Transient Analysis.....
W/ Attached Tables I &II & Fig I.
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