ML20128K985
| ML20128K985 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 07/18/1985 |
| From: | Opeka J NORTHEAST NUCLEAR ENERGY CO., NORTHEAST UTILITIES |
| To: | Youngblood B Office of Nuclear Reactor Regulation |
| References | |
| A04958, A4958, NUDOCS 8507240263 | |
| Download: ML20128K985 (7) | |
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(203) 66s-s000 July 18,1985 Docket No. 50-423 A04958 Director of Nuclear Reactor Regulation Mr. B. 3. Youngblood, Chief Licensing Branch No. I Division of Licensing U. S. Nuclear Regulatory Commission Washington, D. C. 20555
Reference:
(1)
B. 3. Youngblood to 3. F. Opeka, Request for Additional Information, Questions 440.72 through 440.77.
Dear Mr. Youngblood:
Millstone Nuclear Power Station, Unit No. 3 Transmittal of Responses to Questions 440.72 through 440.77 Enclosed are Northeast Nuclear Energy Company's responses to NRC Reactor Systems Branch questions concerning three-loop operation of Millstone Unit No. 3 contained in Reference (1). These responses should fully resolve the Staff's concern regarding these questions.
If there are any questions, please contact our licensing representative directly.
Very truly yours, NORTHEAST NUCLEAR ENERGY COMPANY et. al.
By NORTHEAST NUCLEAR ENERGY COMPANY Their Agent 0 N-
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Senior Vice President 8507240263 850718 PDR ADOCK 05000423 A
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.Then personally appeared before me 3. F. Opeka who being duly sworn, did state that he is Senior Vice President of Northeast Nuclear Energy Company, an Applicant herein, that he is authorized to execute and file the foregoing -
information in the name and on behalf of the Applicants herein and that the statements contained in said-information are true and correct to the best of his knowledge and belief.
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' RESPONSES TO THE NRC (RSB) QUESTIONS
-CONCERNING N-1 (3-LOOP) OPERATION
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440.72' In.'Section' 15.1.5 you state that' the main steam line rupture t
. analysis has been performed by assuming the most limiting single failure of a stuck RCCA with or. without offsite power, and
- assuming the most limiting single failure in the engineered safety.
features. Identify the single failure and explain ~ why it is the most limiting single failure.
. Response:
o The limiting single failure assumed in the Engineered Safety Feature systems is the failure of one ECCS train. This is the limiting single failure because it reduces ECCS flow and delays -
the injection of boron 'into the core. This allows for a higher power level af ter the return to criticality.
440.73 You state in Section 15.1.5.2 that the steam line rupture analysis assumes the boric acid is mixed with, and diluted by, the water flowing in the RCS prior to entering the reactor core. Discuss the design.or procedural provisions incorporated to preclude boron stratification when ' boron' solution is injected into the inactive 4
loop.
Response
If safety injection (SI) is actuated during the course of N-1 Loop
. operation due to a postulated small break LOCA or a main steam line break,it is expected that boron in the Si flow supplied to the isolated loop cold leg will be well mixed with ambient primary coolant by the time the mixture enters the core region. The isoitted loop in the N-1 configuration behaves hydraulically like the stagnant loop that was studied for Pressurized Thermal Shock (PTS) concerns. Experimental thermal fluid mixing data obtained
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. from the testing of a one-fif th scale model of a typical PWR cold leg and downcomer, Reference 1, clearly demonstrates that good mLing' of simulated SI and ambient fluid within the model is obtained at the midplane of the model's downcomer. Further, analytical studies performed using advanced hydrodynamic codes that have been benchmarked against the one-fifth scale model
_. test' data have shown that the mixing process continues as the cooler mixture of the 51 and ambier}t primary coolant flows into the reactor vessellower plenum,' Reference 2. These data support
- the conclusion that,. should 51 be actuated during N-1 Loop -
operation, the SI' flow supplied to the isolated loop will become well mixed with ambient primary coolant prior to that flow entering the core region. If the SI flow mixes well with the -
ambient primary coolant, it follows that the boron carried with the 51 will also become well mixed with the primary coolant prior
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to that flow entering the core region.
The experimental and analytical. results noted in the preceding paragraph assumed no contributiori to the mixing process due to the flow from adjacent active loops, and therefore display a minimum amount of mixing that may be expected due to SI in an isolated loop. The action of flow from the active Icaps, even at natural circulation. flow rates, will promote and further improve the mixing of SI and ambient primary coolant prior to that flow's entry into the core region.
I References.
Lt.
- Report NP-2035, P. H. Rothe and M. W. Fanning, " Thermal Mixing in a
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Model Cold ' Leg and Downcomer at Low Flow Rates, Electric Power
. Research Institute, March,1983.
2.'
Presentation to NRC, L. L. Eyler and D. S. Trent, " PTS Mixing Simulations
' Using the TEMPEST Code," Bethesda, Maryland, December 2,1982, p
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440.74 You state in.Section 15.2.7.2 that the method of analysis for loss of normal feedwater flow assumes the most limiting single failure in the auxiliary feedwater system, and that only one active steam generator would receive feedwater supply for N-1 Loop operation.
It ' appears that operator's action would be required to divert the feedwater flu t'o the active SG. Discuss the design or procedural provisions incorporated to preclude feedwater being introduced inadvertently to the inactive SG and overfilling it to the main steam line.
Response
The N-1 Loss of Normal Feedwater (LONF) analysis assumes that the inactive steam generator is isolated from the main feedwater and auxiliary feedwater systems. The one motor driven AFW pump that is assumed to start is the pump that would feed the isolated steam generator and one other steam generator. Millstone Unit No. 3 three Loop operating procedure requires isolation of main feedwater and auxiliary feeds ater to the inactive steam generator.
Because the one steam generator is isolated, the operating motor driven AFW pump feeds only one steam generator.
440.75 You indicate in Section 15.2.8.2 that the auxiliary feedwater system is assumed to supply 375 gpm to two intact steam generators for N-1 Loop operation, while Table 15.2.1 suggests that auxiliary feedwater would be delivered to three intact steam gererators for both N and N-1 Loop operations. Please clarify these statements.
I
Response
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For the 3-loop (N-1) case, the auxiliary feedwater system is assumed to supply 375 gpm to two intact steam generators.
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I 440.76 You state in Section 15.2.8.2 that the reactor returns to L-criticality briefly following a feedline break with offsite power available because of the cooldown caused by steam generator blowdown.
It appears that the reactor would also return to criticality following a main steam line break with offsite power available.
Discuss the design or procedural precautions incorporated to preclude overcooling the reactor by introducing excessive auxiliary feedwater to the steam generators.
Response
The procedural precautions are incorporated in the EOP to preclude overcooling the reactor by introducing exessive auxiliary feedwater to the steam generator. The reactor does return to criticality following the main steam line break, (MSLB) but the peak power is less than nominal and the minimum DNBR is greater than 1.30. The MSLB cooldown ( 1000F) is more severe than the Main Feedline Break cooldown, ( 500F) as can be seen in the coolant temperatures shown on Figures 15.1-16A and 15.2-15A.
440.77 Westinghouse has recently informed the staff that the rod withdrawal event during Mode 3 operations (hot shutdown), when analyzed with only one RCP operating consistent with current technical specifications for Westinghouse plants, may violate the DNB criterion and thereby not conform to CDC-10. We have also determined that the technical specifications for mode 4 operation, which allows only one RCP to be in operation may also not supported by the existing analysis.
With regard to your request for approval for N-1 Loop operation, please confirm that the Millstone 3 technical specifications ensure operation of the plant in modes 3 and 4 consistent with the safety analyses which support operation in these modes,in particular with respect to the number of RCP's required to be operational. Address both the current technical specifications as well as any impact N-1 Loop operation might have.
.i
Response
p Millstone 3 Technical Specification 3.4.1.2 is consistent with the guidance ' provided in NUREG-0452 Rev. 5 (Draft).
' Action j
statement 3.4.1.2A requires that the reactor system trip breakers be opened within I hour if less than two reactor coolant pumps are in operation. Opening the trip breakers will remove power from' the control rod drive mechanisms and is sufficient to prevent a - single failure induced control rod drive withdrawal event under these conditions.
The rod withdrawal event analyzed in Mode 3 assumed that two reactor coolant pumps were operating.
The analysis was performed for the N Loop configuration: all loops unisolated, 2 L
loops with forward flow, and 2 inactive loops with reverse flow.
A separate analysis was not performed for the N-1 Loop i
configuration: 1 Isolated loop, 2 active loops with forward flow, I
and 1 inactive Loop with reverse flow. The N Loop analysis is h-bounding since the N Loop analysis core flow is less than the N-1 l
Loop analysis core flow.
Westinghouse does not analyze the rod withdrawal from a l
subcritical event in mode 4. Millstone 3 Technical Specifications (3.4.1.3) will be revised to incorporate the requirement to have two reactor coolant pumps operating in mode 4 when the. trip I
breakers are closed.
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