ML20217N856
| ML20217N856 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 04/27/1998 |
| From: | Gordon Peterson DUKE POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| TAC-M99629, TAC-M99630, NUDOCS 9805050455 | |
| Download: ML20217N856 (11) | |
Text
'
iq
. Duke Power Company
^ M % G"v*"I l
[(.)
e Catawba Nudear Station jy mm-w w 4800 Concord Road i
York, SC 29745 Gary R. Peterson (803) 831-4231 omCE Vice President (803) 831-3426ax
)
April 27, 1998 U.S. Nuclear Regulatory Commission Attention:
Document Control Desk Washington, D.C.
20555
Subject:
Duke Energy Corporation-Catawba Nuclear Station, Units 1 and 2 Docket Numbers 50-413 and 50-414 Reply to Staff Position and Request for Additional l
Information on Amendment Request Regarding Pressure Temperature Limits TAC Numbers M99629 and M99630
Reference:
Letter from Peter S.-Tam, NF.C, to Gary R.
Peterson, Duke, same subject, dated March 26, 1998.
Pursuant to 10 CFR 50.4 and 10 CFR 50.90, Duke Energy Corporation is hereby responding to Questions 3 and 4 of the NRC request for additional information contained in the-reference.
Our response to Questions 1 and 2 of'the' subject request will involve additional proposed changes to j
Catawba's Technical Specifications and will be provided under separate correspondence.
(
The enclosed information contains a restatement of Questions j
3 and 4, followed by our response.
Pursuant to~10 CFR 50'.91, a copy of this reply is being sent to the appropriate State of South Carolina official.
Inquiries on this matter should be directed to L.J. Rudy at (803) 831-3084.
Very-truly yours,
/
/
Gary %
R.
Peterson 9805050455 980427 P '
ADOCK 05000413 PDR p\\
O PDR c
j U.S. Nuclear Regulatory Commission Page 21 April 27,.1998 4
LJR/s Attachment xc (with attachment):
'L.A.
Reyes U.S. Nuclear Regulatory Commission Regional Administrator, Region II Atlanta Federal Center
-61 Forsyth St.,
SW, Suite 23T85 Atlanta, GA 30303 D.J-Roberts Senior Resident Inspector (CNS)
U.S. Nuclear Regulatory Commission j
Catawba Nuclear Station
'P.S.
Tam NRC Senior Project Manager (CNS)
U.S. Nuclear Regulatory Commission
' Mail Stop O-14H25-Washington, D.C.
20555-0001-M. Batavia, Chief Bureau of Radiological Health 2600 Bull St.
Columbia, SC 29207 I:
L
Attachment Staff Position and Request for Additional Information Catawba Nuclear Station, Units 1 and 2 Proposed Amendment on Pressure Versus' Temperature Limits Reply to Questions 3 and 4 3.
According to Duke's March 5, 1998, letter, a boltup temperature of 60 degrees Fahrenheit (60*F) was calculated.
However, Duke further stated that "with incorporation of the proposed P/T (pressure-temperature] curves, there will be a need to require that the minimum-temperature during LTOP mode be no lower than 66*F on Unit 1 and 65'F on Unit 2.
These numbers are based on the peak pressure from the limiting steady state condition mass addition transient from an NI pump, with no NC pumps running."
Do the 66*F and 658F temperatures discussed account for temperature instrument uncertainties?
DUKE RESPONSE
'Yes, these numbers have accounted for temperature
' uncertainty. The Duke uncertainty analysis actually
}
concluded with a 7.l'F uncertainty.
The LTOP evaluation conservatively used 10 F in the original development of these numbers.
Based on the peak pressure from a steady state condition l
mass addition event from a single safety injection pump with no reactor coolant pumps on, the minimum temperature during LTOP n.cde shall be no los.er than the following:
i l
Unit 1 peak pressure from safety injection pump with no reactor. coolant pumps on = 525.3 psig (assuming 3 second PORV stroke time which results in a 64.7 psi overshoot, 4.6 psi'for the difference in elevation Patween the beltline region and the-instrumentation locations, 55 psi for instrumentation uncertainty, 1 psi for flow dp offect, and a 400 psig'LTOP setpoint):
From WCAP - 13720 extrapolation (data presented in RAI
. question #4 response)
Appendix G Steady State pressure = 515.80 0 55 F Appendix G Steady State pressure = 560.16 0 60 F
I t
' Interpolation of 525;.3.psig-between'these above values
. yields a value'of 56.070F (rounded to 56.loF).
Adding the temperature uncertainty to this number yields:
56.10F + 100F = 66.loF (10oF-uncertainty) 56.1 F + 7.1 F = 63.20F (7.1 F uncertainty from Duke uncertainty analysis)
Unit 2 peak pressure'from> safety injection pump.with no reactor coolant pumps on = 535.3 psig (assuming 3 second j
PORV stroke time which results in a 74.7 psi overshoot, 4.6 i
psi for the differenceLin elevation between the beltline region and the instrumentation locations, 55 psi for instrumentation uncertainty, 1 psi for flow dp effect, and a 400 psig LTOP setpoint):
From WCAP - 13875 extrapolation'(data presented in RAI
- question'#4 response)
Appendix G Steady State pressure = 524.64 0 500F Appendix G Steady State pressure = 535.34 0 55 F Interpolation of 535.3 psig between these above values yiel'ds a value of 54.98oF'(rounded to 550F).
Adding the-temperature uncertainty to this numberlyields:
t SS F + 10oF = 65oF (100F uncertainty).
55 F + 7.10F = 62.10F (7.10F uncertainty from Duke uncertainty analysis) i Therefore, since the minimum allowed boltup temperature is actually slightly less than the above mentioned numbers and recognizing the fact that the-boltup temperature is the minimum temperature at the time of reactor vessel head detensioning/ tensioning and is not necessarily a minimum temperacure while in-all LTOP conditions, it is appropriate to impose a minimum temperature while in LTOP.
It was Duke's intention to impose this limit in operating procedures similar to the restrictions imposed for the number of reactor coolant pumps operating at a given temperature.
In keeping with NRC Staff's position that Catawba's Technical Specifications reflect the lowest temperature
- justified by LTOP, Duke proposes that both Catawba 1-and 2
~
Techni' cal Specifications be modified to show that 650F be the lowest temperature while in LTOP mode when the pressurizer PORVs are relied upon to mitigate an'LTOP evt at.
.The-65 F limit is slightly less than the number shown above for Unit 1 assuming a 10 F uncertainty, but it is still
r I
higher than the 63.20F value calculated above assuming 7.1 F j
from Duke's uncertainty analysis.
Using a 65oF limit for j
both units would provide the benefit of consistency between j
the units.
The proposed changes to the Technical Specifications will be submitted in the response to Questions 1 and 2.
1
)
4.
Provide a table with the extrapolated values of the pressure temperature limits used for the LTOP analyses and the value of each term (e.g., overshoot, instrument uncertainty, etc.) uned in the calculation of the 400 psig power-operated relief valve lift setpoint.
i DUKE RESPONSE Since LTOP transients are basically acknowledged as being a steady-state event, the LTOP setpoints should be based on i
steady-state pressure-temperature curves.
Starting at a high temperature and going down, a typical pressure-temperature curve has a decreasing slope as temperature decreases.
Therefore, data points below the 850F data points provided in WCAP-13720, Analysis of Capsule Y from the Duke Power Company Catawba Unit 1 Reactor Vessel Radiation Surveillance Program and WCAP-13875, Analysis of Capsule X from the Duke Power Company Catawba Unit 2 Reactor Vessel Radiation Surveillance Program are determined using a constant slope of a straight line based on the last two data points from the steady-state curve.
This method of extrapolatior. was recently reconfirmed by Westinghouse letter No. NSD-E-MSI-98-045, dated February 23, 1998.
Catawba Unit 1:
Following are the data points (from WCAP-13720 page B-18, Steady State) used to determine the equation of a straight line:
Temperature Pressure 116oF 1057.02 psig 120oF 1092.51 psig (Note: The flange limit data points are excluded or else the results would show a flat line extrapolation of 6210F which would be non-conservative.)
Following is the equation used to determine the pressure and temperature data points below 85oF:
4 Y = mx + b Y=
(8.8725)
(X) + 27.81
Contained in the following table are conservative allowable pressure. values for moderator temperatures below 85oF:
Temperature Pressure 80 F 621.00 psig 750F 621.00 psig 700F 621.00 psig 670F 621.00'psig 650F 604.52 psig 60 F-560.16 psig 55oF 515.80 psig 50 F 471.43 psig Catawba Unit 2:
Following are the data points (from WCAP-13875 page B-19, Steady State) used to determine the equation of a straight line:
Temperature Pressure 85oF' 599.54 psig I
90oF 610.24 psig l
(Note: The above mentioned data points are already below the
{
flange limit data point, thus excluding the flange limit data points.)
Following is the equation used to determine _the pressure and temperature data points below 85oF:
Y = nm + b Y=
(2.14) * (X) + 417.64 Contained in the following table are conservative allowable pressure values for moderator temperatures below 85 F:
Temperature Pressure 30 F 588.84 psig 75oF 578.14 psig 70oF 567.44 psig 65 F 556.74 psig 60 F 546.04 psig 55 F 535.34 psig 50 F 524.64 psig
l To consolidate the terms used in the'LTOP analysis (e.g.,
overshoot, instrument uncertainty, etc.), refer to the following tables and figures:
Mass Injection Unit 1 Unit 2 Comments Overshoot 64.7 74.7 3.0 sec PORV stroke time psi psi 1
Reactor beltline /
4.6 4.6 transmitter psi psi elevation difference i
Instrument 55 psi 55 psi
'l uncertainty 100F 10 F Delta P from 4 56.9 56.9
]
RCPs psi psi Delta P from 3 34.7 34.7 RCPs psi psi Delta P from 2 16.8 16.8 RCPs psi psi Delta P from 1 RCP 4.8 4.8 psi psi l
Delta P w/o RCPs 1 psi 1 psi Heat Injection Unit 1 Unit 2 Comments Overshoot 34 psi 20.3 3.0 sec PORV stroke time psi Reactor Coolant System temperature < 1000F Overshoot 77 psi 55.3 Reactor Coolant System psi temperature 100 to 1800F Overshoot 110 80.3 Reactor Coolant System psi psi temperature.180 to 2500F Reactor beltline /
4.6 4.6 transmitter psi psi elevation i
difference l
Instrument 55 psi 55 psi l
uncertainty 100F 10oF Delta P from 1 RCP 4.8 4.8 Worst case event being psi psi initiated by a'startup of a single RCP.
1 In addition, the NRC provided the following two additional questions verbally on April 23, 1998.
Below are these two questions restated and Duke's reply:
Question 1.
Do the dynamic pressure drops due to reactor coolant pump operation (with or without pumps running) include dynamic pressure drop due to two trains of residual heat removal also in service?
Answer 1. Yes.
For conservatism, the flow dp effects of two trains of residual heat removal operation have been factored into the pressure drop values associated with reactor coolant pump operation.
)
1 Question 2. For the 1977 Pressure Mitigating Systems Transient Analysis Results, Westinghouse Electric Corporation for Reactor Coolant Overpressurization, did you rely on the Westinghouse provided data and' perform a linear interpretation or did you re-run the results?
If you relied-on the Westinghouse provided data and performed a linear interpretation technique, what restrictions were used for j
the steam generator heat transfer area with regards to a heat input transient?
And what restriction on the reactor I
coolant system volume was used for a mass input transient?
]
Arswer 2. Catawba utilized the data provided in both the July 1977 and the September 1977 Pressure Mitigating Systems Transient Analysis Results, Westinghouse Electric Corporation for Reactor Coolant Overpressurization.
For the mass input transient case the following reactor coolant system volumes were used for a simple linear interpretation from Figure 4.2.2 from July 1977 Pressure L
Mitigating Systems Transient Analysis Results, Westinghouse l
Electric Corporation for Reactor Coolant Overpressurization:
Unit 1 Unit 2 13,000 cubic feet 12,000 cubic feet For the~ heat input transient case the following steam generator heat transfer surface areas were used (for I
calculation with a 3 second PORV opening time) from Figures 14 and 15 from September 1977 Pressure Mitigating Systems Transient Analysis Results, Westinghouse Electric Corporation for Reactor Coolant Overpressurization:
Unit 1 Unit 2 79,800 square feet 48,000 square feet i
4 NOTES:
- 1) Figures 14 and 15-are for a 58,000 square foot heat-transfer surface area steam generator.
This had to be adjusted for by a simple ratio of the actual heat transfer to 58,000 square feet.
- 2) Unit 2 also had to be adjusted to the reactor coolant system volume of 12,000 cubic feet by a simple interpretation of the dp results using the 13,000 cubic feet graph and the 6,000 cubic feet graph.
- 3) Unit 1 BWI steam generators which use alloy 690 actually have a different heat transfer conductivity which is lower than the original material.
A lower thermal conductivity would be beneficial from the standpoint of a heat input transient case, but no credit was taken for this benefit.
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