ML20038B523
| ML20038B523 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 11/30/1981 |
| From: | Walbridge W SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| To: | Stolz J Office of Nuclear Reactor Regulation |
| References | |
| TASK-2.E.1.1, TASK-TM TAC-44673, NUDOCS 8112080370 | |
| Download: ML20038B523 (6) | |
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4 e suuo SACRAMENTO MUNICIPAL UTILITY DISTRICT C) 6201 S Street. Box 15830. Sacramento. California 95813; (916) 452 3211 November 30, 1981 p
A bEO7 DIRECTOR OF NUCLEAR REACTOR REGULATION b87am, t.g l
]7,j eMCC ATTENTION JOHN F STOLZ CHIEF u, s, I
OPERATING REACTORS BRANCH 4
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i U S NUCLEAR REGULATORY COMMISSION WASHINGTON D C 20555 s
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a rv i DOCKET 50-312 RANCHO SECO NUCLEAR GENERATING STATION UNIT NO 1 On November 16, 1981, representatives of the Sacramento Municipal Utility District met with NRC staff concerning responses to the Atomic Licensing Appeal Board's memorandum and order of October 7, 1981, (ALAB-655). It became apparent during this meeting that NRC staff required information and verification from the District on several items to reference in its responses to the Appeal Board's questions. This letter is intended to l
provide that information.
By letter dated January 21, 1980, the District described possible improve-ments to the reactor coolant flow signal from the Nuclear Instrumentation /
Reactor Protection System to the Integrated Control System. At the present time, we do not intend to implement these improvements. A major factor in this decision was our determination to install a safety -grade Emergency Feedwater Initiation and Control System (EFIC) at Rancho Seco. On the other hand, we also described an improvement to a main feedwater pump turbine drive minimum speed control in the January 21, 1980 letter. A new Feedwater Pump Control System has been purchased and installed at Rancho Seco.
We plan to submit abnormal transient operator guidelines for NRC review by July 1, 1982.
It is expected that these guidelines will be converted into Rancho Seco Operating Procedures during the outage for refueling and comple-I tion of TMI modifications presently scheduled to begin September 1, 1982.
The reliability analysis of the Rancho Seco Unit No. 1 Auxiliary Feedwater System inadvertently assumed the existence of auxiliary feedwater pump suction and discharge pressure indication in the Control Room. This instrumentation does not exist at Rancho Seco Unit No. 1, and the error was corrected in our letter of April 14, 1980.
It should be noted at this time, that none of our operating procedures rely on this instrumentation.
We would like to verify at this time that the auxiliary.feedwater flow control valves (FV-20527 & FV-20528) were tested and that periodic testing procedures have been implemented for functional tests of the auxiliary feedwater automatic initiation circuitry as requested in your letter of February 26, 1980. We would f[g( p:fSM8 A
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- o JOHN F STOLZ Page 2 November 30, 1981 also like to verify that procedures have been implemented for performing channel functional tests of the auxiliary feedwater flow indication and channel checks of the steam generator level indication as requested in the same letter. Also, as requested in the same letter, we have revised procedures to describe methods for obtaining water from the Folsom South Canal or the Plant Reservoir for the Auxiliary Feedwater System should the water from the condensate water storage tank be unavailable.
In particular these procedures address operation of the canal pumping station pumps.
The flow requirements for the Auxiliary Feedwater Syetem at Rancho Seco Unit No. I have been evaluated to determine that the existing flow capacity is adequate to meet the system requirements. This evaluation is attached for your reference.
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Wm. C. Walbridge General Manager Attachment
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- I RANCHO SECO AUXILIARY FEEDWATER FLOW EVALUATION The design basis event for sizing the Auxiliary Feedwater System (AFWS) is Loss of Feedwater (LOFW) with a concurrent Loss of Offsite Power (LOOP), and subsequently loss of reactor coolant pumps. The pertinenc parameters for this accident relative to the'AFWS are design flowrate and required time'to full AFWS flow. These parameters reflect the functional requirements of the AFWS to a) remove decay heat, and b) provide a smooth reactor coolant flow transition i
1 from RC pump operation to natural circulation. The design values which resulted from this analysis are 780 gpm deliverable to the steam generators within 40 seconds of the initiation signal. The 40 second time was chosen to allow the AFWS to inject feedwater and begin increasing SG level to the 50% operating range level, required for natural circulation, prior to completion of the RC pump coast-down. At that time, the design flowrate was selected to be equal to or greater l
than the decay heat generation rate.
Since decay heat rate changes with time, other values than 40 seconds and 780 gpm could have been used and been acceptable.
All other transients which either regnfre er assume the availability of AFW in the Safety Analysis use the design values derivcd from the LOFW analysis. The results of these other analysis are acceptable and are referenced in Table 1, attached.
4 Subsequent to this original analysis, additional analysis was done indicating that 4
j a required AFW flowrate of 760 gpm was sufficient to meet the decay heat generating at time of AFW initiation. These results were described in the B&W, May 16, 1979 letter to the NRC,'following the Three Mile Island accident.
Accidents 1, 2 and 3 of Table 1, which specifically require AFW for mitigation, were analyzed using the original AFWS performance criteria established by the LOFW accident. The results of these analyses were acceptable and are described in the FSAR sections noted in Table 1.
The other accidents listed in Table 1 (4-12) do not require AFW for mitigation though the availability of the AFWS, as defined by the performance criteria established by the LOFW accident, is assumed.
The results of those analyses were acceptable and are described in the FSAR sections noted in Table 1.
The accidents listed in Table 1 have not been reanalyzed using the revised AFW flow requirement (760 gpm). However, more recent analysis on identical plants indicate a significantly lower flowrate is adequate for all accidents addressed in the Rancho Seco FSAR.
Addressing the events included in the NRC letter of February 26, 1980, which have not been included in Table 1, we have the following comments:
OHFW w/ Loss of Onsite and Offsite AC Power - This event was not a design basis of the plant and subsequently is not included in Chapter 14 of FSAR. The B&W Report, " Auxiliary Feedwater Systems Reliability Analyses" (BAW-1504) indicates however, that the SMUD AFW System will provide injection under these conditions.
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Plant Cooldown - Plant cooldown with AFW is a new issue as stated in Reg. Guide 1.139 and not a design basis for this plant. The NRC has not indicated how Reg.
Guide 1.139 is to be applied to operating plants. The extent of plant cooldown for which-the AFWS is designed is discussed in FSAR Section 14.1.2.8.4D.
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Turbine Trip with and without Bypass - This event does not affect the AFWS unless MFW fails, in which case the loss of MFW event previously ~ addressed would bound the AFWS design.
Main Steam Isolation Valve Closure - Again, this event does not directly affect the AFWS unless MFW is lost as discussed above.
Main Feed Line Break - This event was not a require analysis for this plant and is not included in FSAR Section 14.
Main Fecdline Break is a.more. abrupt case of LOFW and results of an analysis would be approximately the same.
Small Break LOCA - The AFW criteria assured for this event is described in Topical Report BAW-10052 updated by letter report, J. H. Taylor.(B&W) to S. A. Varga (NRC), 7/18/78, and B&W report entitled, " Evaluation of Transient Behavior and Small Reactor Coolant System Breaks in the 177 FA Plant", 5/7/77.
I The RCS cooling rate is not a limit relative to accident acceptance criteria.
j The safety limit for all transients which use AFW for mitigation is that the core remain cooled with ultimate acceptance criteria being those addressed in Table 1.
For transients which result in draining the pressurizer or for which natural circulation is slowed or interrupted, restoration of pressurizer level and subcooling is accomplished by swelling due to core heat input and inventory restoration by HPI.
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Steam Generator level is not based on decay heat.remt. val rate or cooldown capa-bility. SG level is set low for decay heat removal and high for natural circu-lation. It is also set high for a small LOCA as described in Topical Report BAW-10052, and in the B&W report, " Evaluation of Transient Behavior and Small Reactor Coolant System Breaks".
As discussed above, the design basis event regarding AFWS design requirements is loss of main feedwater with concurrent loss of RC pumps; the analysis assump-tions for.this event are listed below. Corresponding technical justification, where not specifically listed below, is based on licensing requirements and prudent engineering judgement at the time of the analysis, a) Maximum Rx Power - 100%
b) Time Delay Initiating Event to Rx Trip - The reactor will trip on high RCS pressure approximately 5-10 seconds after a LOFW event. The initiation signal for AFW is loss of main feedwater.
c) AFWS Initiation Signal and Time Delay - The AFW initiation signal for the LOFW event is loss of both main feed pumps as sensed by steam inlet valve positions on the two main feed pump turbines. The design basis time delay from initiation event to full flow of AFW flow into SG is 40 seconds.
d) SG Level at Initiation Event - Steam Generator Inventory is dependent on power level.
In all' cases, AFW flow within 40 seconds will avoid steam generator dryout.
l e) SG Inventory and Decay Heat - For discussion of water inventory see d) 1 above. Reactor decay heat rate is shown in FSAR Table 14, 1-14.
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f) Maximum SG Pressure - 1103 psig.
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Page 3 g) Minimum Number of SG - The number'of generators was not specified in the analysis, heat removal capability is the pertinent parameter and can be accommodated by one SG.
h) RC Flow Condition - Both natural circulation and RC pump operation were analyzed.
- 1) Maximum AW Inlet Temperature - The maximum AFW inlet temperature assumed was 900F.
j) Steam, Feedline Break Time Delay - The feedwater line break was not a required analysis for this plant. Refer to FSAR Section 14.2.2.1 for steam line break analytical information, k) Main Feedline Volume and Temperature Between SG and AFWS - N/A - There is no piping connection between the MFWS and AFWS.
m) Water and Metal Sensible Heat Used - Plant Cooldown was not considered in the design basis analysis.
lx106 BTU /0F was used for the water and metal sensible heat from normal full power Tave to the post-trip Tave setpoint.
n) Time at Hot Standby, etc. Relative to AFW Inventory - The AFW inventory was sized for decay heat removal for the day after Rx trip as discussed in FSAR Section 14.1.2.8.4D.
The design basis for AFWS is not plant cooldown; the NRC Reg. Guide 1.139 requirements for operating plants have not yet been established.
s' TABLE 1 A_CCEPTANCE CRITERIA (
ACCIDENT DESCRIPTION FSAR SECTION
- 1) Loss of Coolant Flow 14.1.2.6 A, B
- 2) Loss of Electric Power 14.1.2.8 & 14.3.2 A,B,D
- 3) Steam Line Break 14.2.2.1 & 14.3.3 D
- 4) Uncompensated Operating Reactivity Changes 14.1.2.1 A, B
- 5) Startup Accident 14.1.2.2 A, B
- 6) Rod Withdrawal Accident at Rated Power Operation 14.1.2.3 A, B
- 7) Moderator Dilution Accident 14.1.2.4 A, B
- 8) Cold Water Accident 14.1.2.5 A, B 9)
Stuck-Out, Stuck-In, or Dropped Control Rod Accident 14.1.2.7 A, B
- 10) Steam Generator Tube Failure 14.2.2.2 & 14.3.4 B, D
- 11) Rod Ejection Accident 14.2.2.4 & 14.3.7 C, D
- 12) Loss of Coolant Accident 14.2.2.5 & 14.3.8 D, E NOTE:
(1)
KEY ACCEPTANCE CRITERIA TECHNICAL BASIS A
Max. RCS Pressure - 110% Design ASME Code B
DNB > 1. 3 with BAW-2 SRP 4.4 C
280 Cal./ Gram Fuel Limit Reg. Guide 1.77 D
Acceptable Doses 10CFR100 E
Fuel Cladding < 2200 F 10CFR50.46