ML20083B100
| ML20083B100 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 05/05/1995 |
| From: | Saccomando D COMMONWEALTH EDISON CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 9505110276 | |
| Download: ML20083B100 (8) | |
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.e I 400 Opus PLu e Downers Gnn c, IL 60515 May 5, 1995 Office of Nuclear Regulation U.
S. Nuclear Reulatory Commission Washington, D.C.
20555 Attn:
Document Control Desk
Subject:
Response to Request for Additional Information (RAI) Pertaining to the Proposed Technical Specification Amendment for the Nominal PORV l
Pressure Relief Setpoint Versus RCS Temperature for the Cold Overpressure Protection System Braidwood Station Units 1 and 1 NRC Docket Number 50-456 and 50-456
References:
1.
R. Assa letter to D. Farrar dated April 20, 1995, transmitting Request for Additional Information 2.
D.
Saccomando letter to the Nuclear Regulatory Commission dated December 16, 1995, transmitting a Proposed Technical Specification Amendment for the Nominal PORV Pressure Relief Setpcint Versus RCS Temperature for the Cold Overpressure Protection System Reference letter 1 transmitted the Nuclear Regulatory Commission's Request for Additional Information (RAI) regarding the proposed Technical Specification amendment for the Nominal PORV Pressure Relief Setpoint Versus RCS Temperature for the Cold Overpressure Protection System which was transmitted in Reference 2.
Our response to the RAI is attached.
If you have any questions regarding this, please contact this office.
i'cerely,
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yu eniserM.) Saccomando Nucleab' licensing Administrator Attachment cc:
R. Assa, Braidwood Project Manager-NRR S. Dupont, Senior Resident Inspector-Braidwood J. Martin, Regional Administrator-RIII Office of Nuclear Safety-IDNS Kinla \\twd\\lt oprall 9505110276 950505 ADDCK05000g56 PDR 8fb\\',,j m__,,
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-l Attachment A Questions arad Responses j
1.
Provide the revised Appendix G curves based on 8.5 effective full power years (EFPY) for the reactor heatup and cooldown i
process.
As a result of a teleconference between Braidwood and the Nuclear Regulatory Commission (NRC) staff held on May 2, 1995, the basis for Braidwood's December 16, 1994 amendment request is under review.
A revised submittal will be made which will address.this question.
2.
Provide the curves of pressure / temperature (P/T) limits which are developed in accordeace with ASME Code Case N-514.
l This question will be addressed in the revised submittal referenced in the response to question 1.
3.
It is our understanding that the instrument uncertainties have not been incorporated in the proposed figure 3.4-da (power operated relief valve (PORV) setpoints for low-temperature overpressure protection (LTOP) applicable up to 8.5 EFPY) as well as in the proposed P/T limits in Item 2 above.
This design is not acceptable to the staff.
The margins in the Appendix G curves can not be used to justify the elimination of the instrument uncertainties in the LTOP setpoints.
Other wise the P/T limits will not be adequately i
protected by these setpoints.
Please provide a revised Figure 3.4-da for staff review.
This question will be addressed in the revised submittal referenced in the response to question 1.
4.
Provide a discussion on the change made for the administrative limit to protect the PORY discharge piping i
from water hammer effects.
The existing LTOP administrative limit at 638 pounds per square inch gauge (psig) for temperatures greater than 250 F was 0
' developed to address an NRC concern. This concern was raised in a teleconference between Braidwood and the NRC regarding Braidwood's July 21, 1994 supplement to a March 30, 1994 request to amend Technical Specifications (TS) 3.4.9.1 and 3.4.9.3.
This amendment request proposed a new LTOP curve that didn't include instrument uncertainties in the 800 psig PORV discharge piping limit.
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In the teleconference mentioned above, the NRC expressed concern that beyond the 638 psig point if all instruments were at the high end of their tolerance bands simultaneously, and these values were added together, it was possible to exceed the 800 psig PORV discharge piping limit.
Since the original LTOP curve expired July 30, 1994, the 638 psig interim administrative limit line was determined to be the most expeditious method of addressing the issue.
The limit value was arrived at based on the following:
796 psig -
Maximum pressure for LTOP event - less than 800 psig PORV discharge piping limit.
- 98 psig -
Greatest pressure overshoot for temperatures 2 250 F.
- 60 nsia -
Random instrument uncertainty.
638 psig -
Administrative limit.
Per standard Westinghouse methodology, the 800 psig PORV discharge piping limit is factored into the development of the LTOP setpoints.
A constant 800 psig Reactor Coolant System (RCS) pressure value is selected to control PORV piping loads due to water hammer effects from PORV actuation during water solid pressurizer conditions.
The pressure values on the revised 10 CFR 50 Appendix G pressure limit curve, or the 800 psig PORV piping water hammer load limit, whichever is lower at a given temperature is used to develop the LTOP PORV setpoint curve.
t As a result of a teleconference between Braidwood and the NRC staff held on May 2, 1995, the basis for Braidwood's December 16, 1994 amendment request is under review.
Any revised LTOP setpoint curve that will be submitted will address the 800 psig PORV discharge piping limit as described above with the appropriate instrument uncertainties applied.
Thus, since the 800 psig PORV discharge piping pressure limit will be factored into the development of the LTOP setpoints with the appropriate instrument uncertainties applied, the administrative limit line will not be needed.
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5.
Technical specification (TS) 3.4.9.3 is effective at 3 5 0'F (Mode 4 and below).
The proposed PORV setpoints cover 4 0 0'F.
What is the actual enable temperature for LTOP7 Discuss the basis for this LTOP enable temperature in light of Appendix G requirements.
Per Braidwood Station operating procedures and TS, the LTOP system is physically enabled at a RCS temperature of 350 F (entry into Mode 4).
The recommended enable temperature for the 8.5 EFPY Appendix G values for Braidwood Unit 1 is based on the following relationship in accordance with Branch Technical Position RSB 5-2 of Standard Review Plan 5.2.2:
RTmy + 9 0 F + max ( AT, 31)
Where:
RTmn is either the 1/4-t Adjusted Reference Temperature (ART)
(105.2 F) or 3/4-t ART (83.9 F)
AT is the temperature difference between RCS water and either ot 1 the 1/4-t (30.293 F) or 3/4-t (63.042 F) metal temperature at the controlling location.
Performing these calculations produces a bounding value of 240 F.
The actual enabling temperature for the LTOP system at Braidwood is then conservatively (LTOP enables sooner on plant cooldown) 0 set 110 F higher at 350 F.
6.
What provisions in the Ts would limit the start of an idle Reactor Coolant Pump (RCP) with a maximum temperature difference between the steam Generator (SG) and the RCs of no more than 50'F.
" Reactor Coolant System - Hot Shutdown,"
limits the start of an idle RCP when the maximum temperature difference between the SG and the RCS exceeds 50 F.
This specification refers to a footnote which reads:
"A reactor coolant pump shall not be started with one or more of the Reactor Coolant System cold leg temperatures less than or t
equal to 350 F unless the secondary water temperature of each steam generator is less than 50 F above each of the Reactor Coolant System cold leg temperatures."
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TS 3.4.1.4.1,
" Reactor Coolant System - Cold Shutdown - Loops Filled," limits the start of an idle RCP when the maximum temperature difference between the SG and the RCS exceeds 50 F.
This specification also refers to a footnote.which reads identically to the footnote for TS 3.4.1.3.
7.
Discuss the limitations in Emergency Core Cooling system (ECCS) TS which could support the assumption used in the mass addition transient for LTOP design.
The Braidwood Unit 1 LTOP nominal PORV setpoint values were based on a mass addition event involving a single centrifugal charging pump operating with the charging line flow control valve fully open and the letdown line flow control valve in the Chemical and Volume Control System failed closed.
Also, LTOP analysis assumes a water solid RCS.
To ensure the LTOP analysis and mass addition transient assumptions are met, Specification 3.5.3, "ECCS Subsystems -
T.y,
< 350 F, " limits operable ECCS subsystems to one subsystem consisting of an operable centrifugal charging pump, an operable Residual Heat Removal (RHR) System Heat Exchanger, an operable RHR pump and an operable water source for these pumps.
Further, TS 3.5.3 requires that:
" A maximum of one charging pump shall be OPERABLE and that pump shall be a centrifugal charging pump, whenever the temperature of one or more of the RCS cold legs is less than or equal to 330 F."
This requirement is contained in a footnote referenced in the body of the Specification.
Surveillance requirement 4.5.3.2 of Specification 3.5.3 states:
"All charging pumps and Safety Injection pumps, except the above allowed OPERABLE pumps shall be demonstrated inoperable by verifying that the motor circuit breakers are secured in the open position at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> whenever the temperature of one or more of the RCS cold legs is less than or equal to 330 F. "
Although Specification 3.5.3 leaves a RHR pump operable, the mass addition transient associated with one centrifugal charging pump remains the most limiting event due to the very low shut-off head (approximately 200 psig) of the RHR pumps as compared to the shut-off head of the centrifugal charging pumps (approximately 2600 psig).
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In Modes 5 and 6 with Pressurizer (PZR) level greater than 5%,
Specification 3.5.4.1, "ECCS Subsystems - T,y, Less than or. Equal to 200 F - Pressurizer Level Greater than 5%," requires that all Safety Injection Pumps be inoperable.
In addition to Specifications 3.5.3 and 3.5.4.1, Specifications 3.1.2.1, "Boration Systems Flow Path - Shutdown" and 3.1.2.3,
" Charging Pump - Shutdown," require that.
" A maximum of one charging pump shall be OPERABLE and that pump shall be a centrifugal charging pump, whenever the temperature of one or more of the RCS cold legs is less than or equal to 330 F."
This requirement is contained in a footnote referenced in the body of the Specifications.
B.
Discuss the TS lLmitations that could assure no more than one RCP operating when the RCS temperature is less than 12 0'F.
There are no TS limitations on starting more than one RCP when the RCS temperature is less than 120 F.
This limitation is contained in Braidwood Operating Procedure (BwOP)RC-1, "Startup of a Reactor Coolant Pump," Braidwood Unit 1 General Operating Procedure (1BwGP) 100-1, " Plant Heatup" and 1BwGP 100-5, " Plant Cooldown" These procedures contain Limitation and Action steps that state in part; "No more than 2ng RCP shall be running when the RCS temperature is s120 F and the PZR is solid. "
This particular step is identified as a commitment step in these procedures and thus cannot be removed from the procedure unless the commitment it is based on is cancelled.
Since Braidwood Station requires verbatim procedure compliance, these procedural steps are considered adequate to ensure no more than one RCP is running when RCS temperature is less than or equal to 120 F.
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9.
Confirm that the transient analysis performed in developing PORV setpoints assumes water solid conditions in RCS.
The transient analysis used for the determination of PORV setpoints for LTOP at Braidwood assumes that the pressurizer and the remainder of the RCS is in a water solid condition.
This is consistent with the design basis for LTOP as defined in WCAP 14040, " Methodology used to develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves."
10.
Discuss the reasons for the highest PORV setpoint at the RCS 8
temperature of 200 F.
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The normal PORV setpoint versus RCS temperature curve developed for the Braidwood Unit 1 LTOP system is an envelope of values calculated by performing an analysis of heat injection and mass injection events at various RCS temperatures such that the appropriate Appendix G plus 10% for Code Case N-514 pressure limits or the 800 psig PORV discharge piping pressure limits are not exceeded.
The peak at 200 F results from the transition from the Appendix G plus 10% pressure limit (which increases with RCS temperature) to a constant 800 psig PORV piping pressure limit which becomes more limiting than the Appendix G plus 10% pressure limit beyond 200 F.
As a result of a teleconference between Braidwood and the NRC staff held on May 2, 1995, the basis for Braidwood's December 16, 1994 amendment request is under review.
It is expected that for any revised LTOP setpoint curve that will be submitted the same peak will occur when the Appendix G P/T limit curve intersects the 800 psig PORV discharge piping limit; however, this point may not be at 200 F on the revised curve.
11.
Provide a tabulation of the data obtained in your analysis l
for the limiting pressure in Appendix G, the lLuiting pressure allowed by ASME Code Case N-514, the PORV setpoint with instrument uncertainties, the increment of pressure during the limiting transient, and the peak transient RCS pressure at 70'F, 15 0'F, 2 0 0'F, 3 0 0'F, and 3 5 0'F.
As a result of a teleconference between Braidwood and the NRC staff held on May 2, 1995, the basis for Braidwood's December 16, 1994 amendment request is under review.
A revised submittal will be made which will address this question.
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1 12.
Confirm that.the PORV setpoints will prevent overpressurization of the RHR System.
The-PORV setpoints selected as part of the LTOP analysis do not specifically consider protection of the RUR system from overpressurization.
The issue of potential RHR system overpressurization'during an LTOP event was addressed by Commonwealth Edison in 1990 as a response to a notification by Westinghouse of RHR system concerns.
The three concerns identified by Westinghouse in this notification are discussed' below.
1.
RCS LTOP events may exceed the capabilities of the RHR relief valves.
The most severe credible transient considered in the design of the Braidwood RHR relief valves is a mass input event in which one centrifugal charging pump is discharging into the RCS in an unthrottled condition while RCS letdown is isolated.
At Braidwood, the RHR inlet relief valve capacity was verified to be sufficient to provide overpressure protection for the RHR system for the mass input from two centrifugal charging pumps operating and discharging into the RCS in an unthrottled condition while letdown is isolated.
The LTOP design basis mass injection transient postulates one centrifugal charging pump injecting to the RCS in an unthrottled condition with RCS letdown isolated. Thus, the preceding discussion shows that, at Braidwood, the RHR inlet relief valve is capable of protecting the RHR system during the design basis LTOP mass injection event.
2.
RHR pressure relief valves may not be able to relieve their rated capacity due to a greater than estimated back-p pressure.
The Westinghouse valve design data used to procure the Braidwood relief valves assumes that the valve discharge pressure is 100 psig and that the discharge flow is two phase at the maximum temperature.
The Braidwood RHR relief valves discharge to the Recycle Holdup Tanks.
This provides a backpressure of only 12 psig versus 100 psig.
Thus, at Braidwood, the valve backpressure is significantly less than the estimated backpressure and issue 2 is not a concern.
3.
Inconsistencies may exist in RHR relief valve design basis documentation.
A review was conducted of the Westinghouse design information and the Braidwood Updated Final Safety Analysis Report and no inconsistencies were identified.
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