ML20042A149
| ML20042A149 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 03/17/1982 |
| From: | Withers B PORTLAND GENERAL ELECTRIC CO. |
| To: | Clark R Office of Nuclear Reactor Regulation |
| References | |
| REF-SSINS-6820, REF-SSINS-SSINS-6 IEB-80-04, IEB-80-4, TAC-46865, NUDOCS 8203230143 | |
| Download: ML20042A149 (5) | |
Text
.
Portland General ElectricCompany r
-- El:
m u =, -..
s..
.a -
March 17, 1982 Trojan Nuclear Plant Docket 50-344 License NPF-1 Director of Nuclear Reactor Regulation gin k/3, ATTN:
Mr. Robert A. Clark, Chief
/s '
V7 Operating Reactors Branch No. 3
/
g Division of Licensing
[
v.3 7
U. S. Nuclear Regulatory Commission
),9 wI~/327s [2 Washington, DC 20555 1
- .L f
Dear Mr. Clark:
\\
' 'd A' -
8 Response to Request for Additiona Q[
Information, IE Bulletin 80-04 Your letter dated January 27, 1982 requested additional information relating to our response to 18 Bulletin 80-04, concerning a PWR main steam line break witt continued feedwater addition, for the Trojan Nuclear Plant.
Subsequently, a conversation with your staff has signi-ficantly modified the scope of this request. Attached is our response to the amended request for information.
Please contact me if you have any questions concerning our response.
Sincerely, G_: z '
Bart D. Withers Vice President Nuclear 00 Attachment I[
l c:
Mr. Lyr.a Frank, Director State of Oregon Department of Energy 8203230140 820317 F'DR ADOCK 05000344 G
PDR 121 S W Sa mon Street. Pomard, Cregon 97204
s Trojan Nuclesr Plant Robert A. Clerk Docket 50-344 March 17, 1982 License NPF-1 Attachment 4
Page 1 of 4
- i.
Response to IE Bulletin 80-04 Request for Additional Information NRC Request 1 Provide a determination of runout auxiliary feedwater (AFW) flow to the affected steam generator during a main steam line break (MSLB). This should be determined from the manufacturer's pump curves at zero back-pressure, unless the system ccatains reliable' anti-runout provisions or an actual backpressure value during the MSLB has been conservatively calculated.
PGE Response l
An evaluation of runout AFW flow to the affected steam generator'during a MSLB has been conducted. This evaluation is based on flow-calculations that first assume a maximum discharge pressure of one AFW pump correspond-ing to delivery of AFW to the intact loop steam generators while reliev-ing through the steam generator safety valves. Maximum flows were then calculated using valve and piping flow resistances and assuming that the broken loop steam generator had completely depressurized to atmospheric pressure, taking no credit for the increase in Containment pressure due to the MSLB. The results show that the maximum single-train AFW flow l
to a broken loop steam generator is 780 gpm.
NRC Request 2 i
Provide an evaluation of the potential for exceeding Containment desiga j'
pressure using the feedwater runout flow rate determined in Request'l.
PGE Response As discussed in our original response to NRC Position 1 of IE Bulle-tin 80-04, the Containment response due to a MSLB with continued AFW flow i
has been addressed in the Trojan analyses. Trojan FSAR Section 6.2.1.1.2.4 l
discusses the Containnent response due to main steam and feedwater line l
breaks inside Containment. The assumptions used for the MSLB evaluation with continued AFW addition, Case E, include an AFW runout flow to the broken loop steam generator of 880 gpm.
Thus, this assumed AFW flow used in the Containment analysis is conservative with respect to the calculated marimum AFW flow to a broken loop steam generator of 780 gpm, as discussed in our response to Request 1.
A graphic representation of the energy release versus time for secondary Plant ruptures is shown in I
FSAR Figure 6.2-57b, and tabulated mass-energy release data used for the Containment response analyses is shown in FSAR Table 6.2-2b.
l A summary discussion of the conclusions reached in the MSLB analyses is presented in FSAR Section'6.2.1.1.2.4.
The Containment pressure response
[-
-for'all the MSLB events that were evaluated result in predicted peak j
pressures that are less than those calculated for the design basis l
loss-of-coolant accident, and thus, Containment design pressure is not
[
exceeded.
i l
i
Trojtn Nuclear Plant Robert A. Clark Docket 50-344 March 17, 1982 License NPF-1 Attachment Page 2 of 4 NRC Request 3 4
Provide the time af ter the start of a MSLB when Containment design pressure will be exceeded if no operator action is taken to terminate the accident. Provide the magnitude of the peak pressure and the time at which the peak occurs.
PCE Response As stated in our previous response, the Containment design pressure is not exceeded in any of the four MSLB events analyzed. The worst case MSLB, Case B, has a pressure peak of 52.9 psig at approximately 140 sec.
The pressure in each case drops as Containment heat removal systems overcome the energy addition rate and and remove more energy than is being added. This is true for Case E where continued energy addition is censidered. Containment energy removal mechanisms described in Trojan FSAR Section 6.2.2 have a capacity significantly greater than the energy addition rate due to AFW runout flow. Even if-operator action at 10 min to terminate AFW flow is not taken credit for, the mass-energy release rate, shown in FSAR Table 6.2-2b at 10 min, will not increase and repres-surization cannot occur. Thus, the Containment design pressure is not exceeded at any time during the transient.
NRC Request 4 Provide an evaluation of the ability of the AFW pumps to continue at runout flow during the accident.
PGE Response The safety-grade AFW pumps for the Trojan Nuclear Plant are described in FSAR Section 6.6.
These pumps are equipped with separate automatic speed controllers that maintain a preset differential pressure between the steam generators and pump discharges. A reduction in steam generator pressure will result in a decrease in pump speed. The turbine-driven AFW pump is equipped with an electronic overspeed limiter that prevents the turbine.from exceeding a speed of 5480 rpm. The diesel-driven. APW pump has a mechanical speed limiter that prevents motor speeds from exceeding 1200 rpm.
In the event that either of these speed limiters should fail to perform its design function, each pump is also equipped with an over-I speed trip mechanism that will trip the pump before damage due to pump overspeed can occur.
The Trojan AFW system has a high flow isolation system, as described in FSAR Section 6.6.3, for each train of AFW. These systems will automati-cally isolate the individual AFW discharge line to the affected steam generator during a secondary line break when the AFW flow exceeds 500 gpm.
4 They are designed to allow only one isolation valve to close in each AFW j
train, thus ensuring that AFW flow will continue to be delivered to the intact loops. The analyzed MSLB event with continued AFW flow to the af fected steam generator takes no credit for functioning of one isolation i
system. Thus, the conclusion is that should either AFW isolation system
Trojan Nuclect Plent RobIrt A. Clark Docket 50-344 March 17, 1982 License NPF-1 Attachment Page 3 of 4 1
fail to properly function during an MSLB, the automatic speed control system and the installed speed limiter would prevent the AFW pump from reaching an overspeed condition.
In addition, the occurrence of a runout flow condition to the broken loop steam generator is limited by the physical constraints of piping and valves in the discharge line, as stated in our response to Request 1.
NRC Request 5 Provide the actions to be taken by the operator to identify the affected steam generator and then to isolate the AFW system. Provide justification for operator action in 10 minutes.
PGE Response For the Trojan Naclear Plant, operator action to isolate AFW flow to the affected steam generator would be necessary only if the AFW automatic isolation system fails to perform its design function, as discussed in our response to Request 4.
Should the AFW isolation system fail to function properly, the following key indications are available to the operator to diagnose a MSLB:
1.
High Containment pressure and humidity.
2.
Low pressure in one steam line.
3.
High steam flow.
4.
Steam flow / feed flow mismatch.
i The Trojan Emergency Instructions are based on Westinghouse generic emergency guidelines. Trojan Emergency Instruction EI-0 (Safety Injec-tion and Diagnosis) is designed to guide the operator through immediate actions and accident diagnosis. After verification of automatic func-tions, accident diagnosis takes place.
One of the primary diagnostic indications that is examined is steam line pressure. This will give positive indication that the accident involves a loss of secondary coolant.
l Emergency Instruction EI-2 addresses events that involve a loss of i
secondary coolant. After verifying that main steam line isolation has occurred and initiating the emergency plan, subsequent actions are to verify that steam dump valves.and atmospheric release valves are not inadvertently open. The operator is then instructed to identify the affected loop by monitoring steam line pressures and to isolate AFW flow to the affected steam generator if the automatic AFW isolation system has failed to do so.
The diagnosis of an in-Containment MSLB is not deemed to be a difficult task. Positive indication of a secondary lite break is given by a significantly reduced pressure in one steam line and supported by rapidly
Trojan Nuc1scr Plant Rob 2rt A. Clark Docket 50-344 March 17, 1982 License NPF-1 Attachment Page 4 of 4 increasing Containment pressure and humidity.
Instructions to isolate AFW flow for both main steam and feedwater line breaks are explicit and occur immediately following diagnosis of the event.
In addition, Trojan operators are required to monitor AFW flow and steam generator levels to ensure adequate heat removal capability. The likelihood of an operator not recognizing a high AFW flow condition is remote. Thus, we conclude that operator action within 10 minutes is fully justifiable.
i l
I l
l GMY/4sh6B4
_ - _ _ _ _ _ _