ML19317H255

Requests Approval for Encl Tech Specs Change Re Reactor Vessel Overpressurization & Forwards Subj B&W Generic Analysis.Refs NRC 760813 Ltr

ML19317H255
Person / Time
Site:	Arkansas Nuclear
Issue date:	12/03/1976
From:	Phillips J ARKANSAS POWER & LIGHT CO.
To:	Ziemann D Office of Nuclear Reactor Regulation
Shared Package
ML19317H256	List: ML19317H255 ML19317H259
References
1-126-1, NUDOCS 8005050442
Download: ML19317H255 (11)

Text

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ugf[onoigg U.s. NUCLE AR CEGULATCRY C 8005050 W 2 7 HSsloN DOCKET NUMEE R 2 2. , 50- 313 NRC DISTRIBUTION roR PART 50 DOCKET MATERIAL FROM: Arkansas Pwr & Light Co DATE FDoCVME" TO: Mr Ziemann Little R ek, Ark 2_3-76 2 D Phillips oATE RECElvED 12-7-76 AE77E R O NoToml2 E D PROP INPUT roRM NUMBER oF COPIES RECElVED Eor '"W AL QUNcLAssiriE o Oco: - one signed OEsCRIPTioN ENCLOSURE Ler re our 8-13-76 ler...trans the followin z: Amdt to OL/ Change to Tech Specs: Consists of revisice with regard to susceptibility to reactor vessel overpressurization......

THIS DOCUMENT CONTAINS consisting or the rotiowing:

POOR QUAUTY PAGES

1. B&W Generic Analysis Concerning Reactor Vessel Overpressurization

. Proposed tech specs.....

REACTOR VESSEL OVERPRESSURIZATION DISTRIBUTION PER G. EECH 10-21-76 (4,0 sets enc 1 rec'd)

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H E L PI N G BUILO ARKANSA9 ARK ANS AS POWE A G LIGHT COMPANY

1. 0. sox 551 UTTLE AOCK. A AK ANSAS 72203.(5013379-4000 December 3, 1976 -

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/'. C Director of Nuclear Reactor Regulati YN ATTN: Mr. D. L. Ziemann, Chief 6s

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/, 4 U. S. Nuclear Regulatory Commission Washington, D. C.

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Subject:

Arkansas Nuclear One-Unit 1 Docket No. 50-313 License No. DPR-51 Reactor Vessel Overpressurization (File: 1510, 1511.1)

Gentlemen:

Your letter of August 13, 1976, requested that we perforn an analysis to determine our susceptibility to an overpressurization event at low system pressure and temperature which might cause us to exceed those pressure / temperature limitations as presented in the Technical Spec. fi-cations (Appendix A to License No. DPR-51). Our letter to you of November 15, 1976, presented justification as to why this event could not occur at Arkansas Nuclear One-Unit 1, but advised that final hardware modifications to ensure this would be submitted to you by December 3, 1976. This letter and the attached BSW generic analysis (Attachment I) '

serves to inform you of these modifications and their bases.

' We will be installing a dual setpoint feature on the pressurizer electro-matic relief valvo during our upcoming refuelf ng shutdown in January 1977.

This dual setpoint feature will enable the setpoint on the electromatic relief valve to be reduced to 550 psig upon reducing the reactor coolant system pressure to 525 psig and system temperature below 280F (reference Figure 3.1.2-2 of Appendix A to License No. DPR-51) . The lower setpoint will be removed when the system is heated up to 280F (reference Figure 3.1.2-1 of Appendix A to License No. DPR-51) . This dual setpoint feature will provide relief capability in the incredible event of overpressurization.

.I.b M 2 TAX A AvlNG. INVESTO A OWN E O MEMBE A MIDDLE SOUTM UTauTIES SYSTEM

4bn. D. L. Ziemann Dscemb9r 3, 1976 1-126-1 The final hardware change involves racking out the breakers to the motor operators of the high pressure injection valves below 280F to eliminate the possibility of initiating high pressure injection into the system at

- cold conditions, thereby causing the system to go solid. However, before this change can be implemented the Technical Specification change to Technical Specification 3.2.1.1 as shown in Attachment II must be approved by you'. This change is needed to eliminate our interpretation that two makeup flow paths be operable below 200F in conjunction with the two makeup

- pumps. As our interpretation now stands, we can not rack out t'he breakers to but three of the high pressure injection valves as a second operable flow path involves one of the high pressure injection legs. The Technical Specification change is consistent with the B6W-STS (3.1.2.1) and will alleviate our interpretation conflict with the final fix as proposed in this letter. It should be noted also that racking out of the breakers to these valve operators will be precluded when the surveillance required by Technical Specification 4.S is to be performed.

Your expeditious concurrence and approval of this Technical Specification change will allow implementation of the second part of this final fix in a timely manner. -

Very truly yours, i ' l l fit 1 i ,'$h J. D. Phillips Senior Vice President .

JDP:tw Attachments e

ATTACIDfENT I B6W Generic Analysis Concerning Reactor Vessel Overpressurization 0

1

l EVALUATION OF POTENTIAL REACTOR VESSEL OVERPRESSURIZATION

1. Purpose-The purpose of this evaluation is to examine the system design and operation ' for susceptability to overpressurization events during start-up and shutdown and to determine the pressure response of the Reactor Coolant System (RCS) to potential events which cause p7e _ure -

-increases.

2. Events Evaluated The events examined in this evaluation were:

a. Erroneous actuation of the High Pressure Injection (HPI) System.

b. Erroneous' opening of the core flood tank discharge valve.

c. Erroneous addition of nitrogen to the pressurizer.

d. Makeup control valve (makeup to the RCS) fails full open.

e. All pressurizer heaters erroneously energized.

f. Temporary loss of the Decay Heat Removal System's capability to remove decay heat from the RCS.

.g. Thermal expansion of RCS after starting an RC pump due to stored thermal energy in the steam generator. .

3. Results of Event Evaluation 3.1 General .

For events which cause the RCS pressure to increase, the pressure will increase significantly faster in a " solid water" system than it will in a system with a steam or gas space. The RCS always operates with a' steam or gas space in the pressurizer; no operations involve '

a " solid water" condition, other than system hydrotest.

Considering the modest rate of pressure rise (because of non-solid pressurizer) from the events and tne high level alarms in the pressurizer that would normally alert the operator, it is reasonable to expect the operator to terminate the' event prior to reaching an overpressurization condition. However, without operator action, the pilot actuated relie. valve located on the pressurizer will terminate any pressure increase, thus preventing an overpressur-

.i:ation condition. A dual setpoint is utilized for this valve to

, provide overpressure protection during startup and shutdown con-ditions. The lower setpoint is enabled by actuation of a switch in the control room during the plant cooldown prior to startup of the Decay Heat Removal System at 2SOF RCS temperature. Character-istics of this valve at the lower setpoint are:

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- $ Open Setpoint 550 psig Close Setpoint 500'psig

3.1 General. -

continued-Steam capacity at 550 psig 25,985 lb/hr Equivalent liquid insurge volume rate into pressur-1:er 2,650 gpm

, Liquid capacity 9 550 psig 500 gpm Nitrogen capacity 0 550 psig 32,420 lb/hr Equivalent liquid insurge voluine rate into pressur-izer 2,350 gpm All-events involving insurge to the pressurizer were evaluated with the pressurizer and makeup tank water levels initially at .high levels. For the pressurizer, a water level at the high high level alarm setpoint was used. The relationship of this level to the other pressurizer water Ivvel setpoints is:

0"-320" Level Indicating range 275" High High level alarm 220" High level alarm 180" Normal level 160" Low level alarm 40" Low level interlock (heater cut-out) _

and alarm -

For ~ the makeuj tank, which is the normal suction source for the makeup /

HPI pump, a water level at the high level alarm setpoint was used. The relationship of this level to the other makeup tank level setpoiats is: ~

0-100" Level Indicating range 86" High level alarm 73" Normal level .

55" Low level alarm The initial pressurizer level used for the event does not affect the peak pressure reached; it only affects the rate of pressure increase.

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3.2' Erroneous Actuation' of the HPI System This event is not credible because the circuit breakers for the closed HP _ injection motor operated valves are " racked out" during the plant cooldown prior to startup of the Decay Heat Removal System.

These valves are 'shown on SAR Figure 9-3. Startup of the Decay Heat Removal . System occurs at an RCS temperature of 280F.

3.3 Erroneous .0pening of the Core Flood Tank. Discharge Valve This event is not credible because this valve is closed and the circuit

, breaker, for the motor operator is " racked out" during the plant cool-down before the RCS pressure _ is decreased to 600 psig.

3.4 Erroneous Addition of Nitrogen to the Pressurizer It is not credible that this event can overpressuize the RCS. Nitrogen

-is added to the . pressurizer during plant cooldown at an RCS pressure of 50 psig or less. Nitrogen addition is controlled by a S0 psig' regulator.

A relief valve (75 psig associated with the regulator provides protec-

, tion in the event of regulator. failure.

3.5- Ibkeup Control Valve (makeup to the RCS) Fails Full Op.:

This valve is on SAR Figure 9-3 and is automatically controlled by the pressurizer _ level controller. The pressu're response of the RCS to_this event is shown on Figure 1. If it is assumed that the operator ~

does not take action to terminate the event during the pressure increase, the peak RCS pressure is limited to 550 psig by the pressurizer pilot actuated relief. valve. Initial conditions used for the analysis were: ,

- a. 275" pressurizer water level (high high alarm setpoint)

b. _86" makeup tank-water level '(high level alarm) c., 32" GPM total; seal injection flow to RC pumps (automatically controlled)

'~d.. _45 GPM letdown flow from RCS to makeup tank i

e. no spray into pressurizer l(normally there would be)

Figure 1 depicts two pressure response curves. One is' for an initial RCS pressure 'of -250 psig ~This is the RCS pressure at which'the Decay Heat . Removal System.is started up during plant.cooldown or at which .the RC pumps;are started during plant heatup. Pressurizer water level would-normally. be about 180" instead of the 275", used in the analysis. The-higher level used in the' analysis increases the' rat of pressure rise, lute other pressure response ~ curve on. Figure 1 is for an . initial RCS pressure _of-100 psig. This is about-.the-lowest RCS pressure at which

.the Makeup System would be:in operation.

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Relief through the pressurizer relief valve will be terminated by operator action (stop makeup pump or close makeup line isolation valve) or without operator action when the makeup tank water volume is exhausted.

Peak insurge rate into the pressurizer is 260 gpm. in addition to the

- alarms shown on Figure 1, other alarms and indications which would alert and aid the operator in evaluating the event are:

a. Pressuri:er high level alarms (s)

(with initial level below high high setpoint which would be normal).

b. liigher than normal makeup line flow rate indication

c. Lower than normal makeup pump discharge pressure- .

d. Full open indicating light for makeup valve

e. liigh temperature alarm for relief valve discharge line (after relief valve relieves)

f. liigher than normal RCS pressure indication

g. liigher than normal pressuri:er level indication 3.6 All Pressurizer lleaters Erroneously Energized The pressure response of the RCS to this event is shown on Figure 2.

If it is assumed that the operator does not take action to terminate the event during the pressure increase, the peak RCS pressure is limited to 550 psig by the pressurizer pilot actuated relief valve.

An initial pressurizer water level of 5, inches (10 inches above low level heater cut-out interlock) was used because the lower water level results in the fastest pressure increase. Even with the low level, the pressure increase is very slow. The pressurizer water level will .

not change during this event as it is being automatically controlled.

The heaters are generating 1625 lbs of steam per hour in the 500 to 550 psig range. In addition to the alarms shown on Figure 2, other alarms and indications which would alert and aid the operator in eval-uating the event are:

a.  !!igher than normal RCS pressure indication

b. liigher than normal letdown flow rate indication to makeup tank (due to increasing RCS pressurizer). '

c. liigher than normal makeup line flow rate indication odue to increas-ing letdown flow rate.

d. lugh temperature alarm for relief valve discharge line (after valve relieves).

e. The "On" indicating lights " lit" for all pressurizer heater banks.

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Relief througit the pressurizer relief valve will be terminated by operator action (de-energize heaters) . Without operator action, the heaters will be de-energized when the pressurizer water level drops to the heater cut-out interlock set point. Since pressurizer water level is on automatic control, water is transferred auto-matica11y from the makeup tank to the RCS to replace that which is lost through the relief valve. For an initial makeup tank level at the high alarm setpoint, it would take six (6) hours to empty the makeup tank and thus result in pressuri er '.ater level decreasing to the heater " cut-out" setpoint.

3.7 Temporary Loss of Decay Heat Removal Systems Capability to Remove Decay Heat From the RCS The pressure response of the RCS to this event is shown on Figure 3.

If it is assumed that they perator does not take action to terminate

'the event during the pressure increase, the peak RCS pressure is limited to 550 psig by the pressurizer pilot actuated relief valve.

Loss of decay heat removal capability could only be caused by loss of flow in the Decay Heat Removal System or in the cooling water system serving the Decay Heat Removal System. Loss of flow in either system would'immediately actuate low flow alarm (s), thus alerting the operator. Relief through the pressurizer 7elief valve will be terminated by operator action restoring the decay heat removal function. Insurge rate into the pressurizer is 98 gpm in the 550 psig pressure range. Conditions used in this pressure response analysis were: ,

a. Event occurs during cooldown after startup of Decay Heat Removal System and shutdown of steam generators.

b. Pressurizer level at 275 inches, normally it would be near 180 inches

c. Cooldown to the Decay Heat Removal System " cut-in" temperature at 100 F/Hr, this produces maximum decay heat generation rate.

d. All decay heat absorbed by reactor coolant, no heat absorbed by the metal components or by the steam generators. Actually, these are heat -

absorbing sinks.

e. 32 gpm total seal injection flow to RC pumps (automatically controlled).  !

f. 45 gpm initial letdown from RCS to makeup tank

g. No spray into pressurizer.

3.8 Start of an RC Pump with Stored aermal Energy in UTSG Secondary Several postulated situations have been examined which may lead to primary fluid expansion due to energy _ absorption from hot OTSG l secondary water after start of an RC pump. The two types of situations l which lead to possible RCS pressurization have been identified as i follows:

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' Type A. Filling of OTSG secondary side _with hot water with subsequent start of an RC pump, and Type B.: Restart 'of an RC pump Under Type A Condition

' Figure number 4 presents results of RCS pressure ~versus time for the worst case Type A (see above) condition. Initial conditions for this transient are a result of filling of the steam generators with feedwater at 420F, This temperature is 'a result of the failure of the feedwater heating controls causing auxiliary steam flow 'to 'the heaters to produce a feedwater temperature -

in excess of the allowable value of 22SF for OTSG fill operations.

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.The temperature of the' feedwater in the OTSG secondary side -

p following. the filling operation reaches a temperature of 240F as does the primary water contained in the RCS at elevations greater than the lower OTSG tubesheet. This is 8 result of the heating of OTSG ~ tubes and primary water during OTSG filling

where. heated primary water circulates to a limited extent i: .through the RCS. At the end of the filling operation, the RCS water located below the OTSG lower tubesheet remains at the initial value of 140F, T -

The primary system pressure versus time as shown in Figure 4 is based. on an initial- pressurizer level at the maximum value of '

the high-high level alarm for a 177 FA plant. The initial

[ pressurizer level'is normally kept much lower to minimize the heating requirements for raising the pressurizer temperature -

,j and pressure in preparation for starting an RC pump. The

-initial' pressure is 300 psig, the normal pressure required

! prior to starting an RC pump. No credit has been taken.for

-pressurizerilevel-control. The pressurizer level increased l . during the transient by 30 inches;-the level would have to ~

i rise an Ladditional "710 inches before entering the" upper head.

Other conditions of primary and secondary temperatures which ,

may exist prior. to starting of. an RC pump' have been evaluated and are bounded by the results- of Figure 4. These conditions include the. situation where the feedwater temperature entering

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the OTSG's during fills the steam generators beyond the ~ maximum

~ allowable level and ' completely. fills the . steam generators.' In addition, the -results jresented here bound the case where the initial RCS temperature i' 50F before filling the steam _ generators.

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3.8.2 Start of an RC Pump Under Type B Condition Figure number 5 presents results of RCS pressure versus time for the Type B conditions (see above). Initial conditions for this transient are a result of the accumulation.of pump seal injection and makeup injection water in the RC cold leg piping during stagnant (no flow) conditions. Although the operator is required to ini-tiate a cooldown of the RCS if RC pumps are inoperable and RC temperature >250F (Plant Limit and Precautions), the assumption is made that the operator fails to do so while allowing makeup and seal injection water temperature to drop to 50F, which is below the minimum value of RC temperature less 120F. The cold water is assumed to accumulate in the RC cold leg piping without mixing with hot RC water. The RC pump is started following a period of one hour of stagnant (no flow) conditions in the RC System.

The primary system pressure versus time as shown in Figure 5 is based on a initial pressurizer level at the maximum value of the high-high level alarm for a 177 FA plant. The initial pressure is 450 psig which is approximately midway between the Tech. -Spec.

and RC pump NPSH pressure limits at 275F. No credit has been taken i

for pressurizer level control. The decrease in pressure at approxi-mately 2 minutes is a result of hot RC primary fluid entering a steam generator which has been cooled by the passage of the slug of low temperature RC fluid (the mixing of RC fluid and heat transfer i

through the OTSG tubing brings -the RC fluid to a constant temperature and produces a net contraction of the fluid and a decrease in system j pressure at final equilibrium conditions) . The pressurizer level increases during the transient by 13 inches; the level would have to rise an additional 87 inches before entering the upper head.

4. Conclusions '

I The preceding evaluation and analysis demonstrates that the reactor vessel is protected from overpressurization during events which cause increasing ,

pressure.

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