ML19337A425

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Overpressure Protection for B&W Pwr. Related Correspondence
ML19337A425
Person / Time
Site: Three Mile Island Constellation icon.png
Issue date: 05/31/1972
From:
BABCOCK & WILCOX CO.
To:
Shared Package
ML19332B231 List:
References
BAW-10043, ISSUANCES-SP, NUDOCS 8009260402
Download: ML19337A425 (17)
v • d • e


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OUTLINE Topical Report BAW-10043, " Overpressure Protection for Babcock & Wilcox Pressurized Water Reactors" This report provides information on the degree of overpressure protection in Babcock & Wilcox pressurized water reactor nuclear steam systems and meets the requirements of the ASME Boiler and Pressure Vessel Code,Section III, Article 9, N-910.2 (1968).

The report demonstrates that the combination of the Reactor Protection System, the pressurizer safety valves, and the steam system safety valves provides overpressure protection for the most severe transients, which include control rod withdrawal at low power and turbine trip from full power.

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INTRODUCTION ............................................. 1 4

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.i DESIGN CRITERIA .......................................... 3 l

i METHOD OF ANALYSIS ....................................... 5 i

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RESULTS AND CONCLUSIONS .................................. 6 i

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BAW-10043 Topical Report May 1972 4

OVERPRESSURE PROTECTION FOR BABCOCK & WILCOX PRESSURIZED WATER REACTORS by J. D. Carlton l .

BABCOCK & WILCOX Power Generation Group Nuclear Power Generation Division P. O. Box 1260 Lynchburg, Virginia 24505 Babcock & Wilcox

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Babcock & Wilcox Power Generation Group i Nuclear Power Generation Division Lynchburg, Virginia Report BAW-10043 I

May 1972 i Overpressure Protection for Babcock & Wilcox Pressurized Water Reactors l J. D. Carlton Key Words: PWR, Overpressure Protection i

ABSTRAC T I

This report is submitted to meet the requirements of the ASME i

Boiler and Pressure Vessel Code,Section III, Article 9, N-910.2 (1968) ,

and to provide information on the degree of overpressure protection in I

i Babcock & Wilcox Pressurized Water Reactor Nuclear Steam Systems.

The report demonstrates that the combination of the Reactor Pro-tection System, the Pressurizer Safety Valves, and the Steam System Safety Valves provides overpressure protection for the most severe transients, which include control rod withdrawal at low power and tur bine trip from full power. -

- ii - Babcock & Wilcox

CONTENTS Page l.

IN TRO D UC TION . . . . . . . . . . . . . . . . . . . . . . . . . . . . I 2 DESIGN CRITERIA .......................... 3 2.1. From ASME Code,Section III, Article 9 . . . . . . . . 3

, 2. 2 Additional Criteria ......................

. 3

3. METHOD OF ANALYSIS ....................... 5 3.1. Analytical Models ..................... . 5
3. 2 Analytical Proc edure . . . . . . . . . . . . . . . . . . . . 5
4. RESULTS AND CONCLUSIONS .................. 6 4.1.
4. 2. G e n e ra l . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 4.3. Control Rod Withdrawal From Low Power . . . . . . .

Turbine Trip From Overpower Conditions . . . . . . .

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4. 4. Conclusions

........................... 6 List of Figures i

Figure

1. Schematic Diagram of NSS Pressure- -

Relieving Devices ........................ 2 2 Maximum Reactor Coolant System Pressure Following Control Rod Withdrawal Transient ...... 7 3

Pressurizer Safety Valve Capacity Required Following Control Rod Withdrawal Transient ...... 8

4. Maximum Steam Generator Pressure Following Turbine Trip From Overpower Conditions . . . . . . . . 9
5. Maximum Reactor Coolant System Pressure Following Turbine Trip From Overpower Conditions ............................. 10
6. Steam System Safety Valve Flow Following 7

Turbine Trip From Overpower Conditions . . . . . . , . 11 Pressurizer Safety Valve Capacity Required Following Turbine Trip From Overpower Conditions .............................

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- 111 - Babcock & Wilcox

1. INTRODUCTION Babcock & Wilcox Pressurized Water Reactor Nuclear Steam Systems are equipped with pressure relief equipment to provide over-pressure protection in accordance with the ASME Boiler and Pressure Vessel Code,Section III, Article 9. As shown in the schematic sketch,

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, Figure 1. the equipment consists of spring-loaded, ASME-rated safety valves on the pressurizer (pressurizer safety valves) and on the main steam lines (steam safety valves) outside the reactor building.

Two pressurizer safety valves are mounted on nozzles on the pre s surizer. The pressurizer safety valves discharge through mani-folding to the reactor building drain tank (quench tank). Each of these valves is rated to carry half of the total rated capacity.

Each steam generator has 11 steam safety valves, giving a total of 22 The total steam system relief capacity is divided equally between the two steam generators.

"'he pressurizer safety valves and the steam system safety valves ac: with the Reactor Protection System (RPS) to provide the required -

overpressure protection. The RPS includes reactor trip functions for the following signals should they reach preset limits: -

1. High reactor coolant system pressure.

2 High reactor outlet temperature. ,

3. Low or loss of reactor coolant flow. j
4. KIgh reactcr power.

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m-= meer name emme = mame m an mas Figure 1. . Schematic, Diagram of NSS Pressure-Relieving Devices

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NAIN STE AN LINE d i PRESSURilER SAFETT l VALVES j STEAM SYSTEM e

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STEAM URGE GENERATOR LINE STEAR REACTOR GENERATOR am FEEDWATER gypp y 1 P

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2 DESIGN CRITERIA f The design criteria for the pressurizer and the steam system safety valves are as follows:

l 2. 1. From ASME Code,Section III, Article 9 1

The rated capacity of the pressure-relieving devices, includ-ing any limitation imposed by the systems connected to the discharge side, shall be sufficient to prevent a rise in pressure within the vessels

{ which they protect of more than 10% above the design pressure at the design temperature when operating under the conditions summarized in i

the technical report of N-910.2 2

' The nominal pressure setting of at least one safety or relief

} valve connected to any vessel or system shall not be greater than the design pressure of the vessel (at design temperature) which it protects.

Additional valves required may have higher nominal settings, but in no case shall these settings exceed 105% of the design pressure (at design te mperature).

2. 2. Additional Criteria .

The pressurizer safety valves and the steam system safety valves are sized on the basis of the maximum pressure transient imposed on the reactor coolant system (pcwer imbalance caused by reactor upset) or imposed on the main steam system (power imbalance caused by tu'r-bine upset conditions). The following upsets are considered:

Control rod withdrawal.

Turbine trip.

Complete loss of power.

I. ors of feedwater flow.

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For the turbine trip condition, the turbine and reactor system are assumed to be operatir.g at a power level just under the high flux Babcock .Wilcox

lI reactor trip setpoint, which is slightly greater than the turbine's capa-bility.

2 The turbine bypass system is assumed not to actuate.

3. Control system runback of reactor power and turbine power I is assumed not to initiate.

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4. No direct reactor trip is assumed to result from turbine trip.
5. The pressurizer spray and the power-operated pressurizer j relief valva are assumed not to actuate.
6. The moderator and Doppler reactivity feedback is assumed to be zero.

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7. All safety valves are assumed to have 3% accumulation; i. e. ,

100% capacity is obtained at 103% set pressure.

8. The reactor coolant system and steam system temperature 1 l

and pressure are assumed to be at nominal values with all trip setpoints at maximum tolerances. The maximum tolerances are as follows:

f High flux = nominal trip point + 6.5%

High temperature = nomirni trip point + 3F 1

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3. METHOD OF ANALYSIS

. 3.1. Analytical Models The transients are analyzed on an analog / hybrid and digital sim-

. ulation of the unit. The models contain point kinetics, core thermody-namics, reactor coolant system hot leg and cold leg delays, steam generator primary and secondary, feedwater system, steam system including safety valves, and pressurizer including safety valves.

3. 2 Analytical Procedure The pressurizer safety valve capacity is analyzed on the basis of nuclear system upsets including (1) control rod withdrawal from low and high power, (2) control rod ejection from low and high power, (3) turbine trip conditions, and (4) complete loss of power transient. In-cluded in the pressure analysis are the pressure drop from the reactor coolant pump discharge nozzles to the connection with the surge line, the surge line pressure drop, the safety valve nozzle and inlet pressure drop, and the discharge manifold characteristics.

The steam system safety valve analysis includes the steam line

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l 4. RESULTS AND CONCLUSIONS l 4. 1. General

, The analyses of safety valve capacity have shown that the upsets that produce the largest pressure transients are the control rod with-drawal from low power and the turbine trip from the overpower condi-tion.

4. 2 Control Rod Withdrawal From Low Power

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The control rod withdrawal transient is more severe at zero or Iow power conditions because the steam generator secondary inventory is minimal and provides for minimum heat removal. (The steam gen-erator inventory is constant from 0 to 15% power and increases to a maximum at full power. )

I The control rod withdrawal transient produces the pressure trace g

shown in Figure 2 The pressurizer safety valve flow rate in percent of available relieving capacity is shown in Figure 3.

4. 3. Turbine Trip From Overpower Conditions t

The turbine trip from maximum overpower conditions produces the most severe combined pressure transient on both the steam sys -

tem and the reactor coolant system. The pressure responses for the steam system and the reactor coolant system are shown in Figures 4 and5.

Figures 6 and 7 illustrate the steam system safety valve and pressurizer safety valve flow rates in percent of i . stalled capacity.

4. 4. Conclusions The pressurizer safety valves and the steam system safety valves provide sufficient overpressure protection. For the pressurizer safety

. valves, the margin on capacity is a factor of 2 The corresponding capacity margin for the steam safety valves is 65 Babcock s.Wilcox l

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I Figure 2 Maximum Reactor Coolant System Pressure Following Control Rod Withdrawal Transient i 2800 1

1105 Design Pressure 1 2700 1

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I Figure 3 Pressurizer Safety Valve Capacity Required Following Control Rod Withdrawal Transient i

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Figure 4. Maximum Steam Generator Pressure Following Turbine Trip From Overpower Conditions I200 I .

1105 Design Pressure 1

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I Figure 5. Maximum Reactor Coolant System Pressure Following Turbine Trip From Overpower Conditions ,

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i Figure 7. Pressurizer Safety Valve Capacity Required Following Turbine Trip From Overpower Conditions l

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