ML20044G140

Safety Evaluation of MSIV Low Turbine Inlet Pressure Isolation Setpoint Change for Pilgrim Nuclear Power Plant.

ML20044G140
Person / Time
Site:	Pilgrim
Issue date:	05/31/1986
From:	Alvarez E, Cornwell K, Sozzi G GENERAL ELECTRIC CO.
To:
Shared Package
ML20044G126	List: ML20044G125 ML20044G133 ML20044G135 ML20044G140
References
NEDO-31296, NUDOCS 9306020074
Download: ML20044G140 (23)

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SAFETY EVALUATION OF.MSIV  !

.. LOW TURBINE INLET PRESSURE  !

!! ISOLATION SETPOINT CHANGE FOR

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SAFETY EVALUATION OF MSIV LOW TURBINE INLET  ;

PRESSURE ISOLATION SETPOINT CHANGE FOR PIIERIM NUCLEAR POWER STATION  !

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I E. M. Alvarez l K. F. Cornwell ,

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G/L.Sozhi,Pyt%g'er . Artigas, Manager Application Engineering Licensing Services  ;

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NUCLEAR ENERGY BUSINESS OF'E. RATIONS

• GENERAL ELECTRIC COMPANY SAN JOSE, CALIFORNIA 95125 GENERAL $ ELECTRIC  ;

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i IMPORTANT NOTICE REGARDING l

CONTENTS OF THIS REPORT  !

F PLEASE READ CAREFULLY  ;

i This report was prepared by General Electric solely for the use of the l Boston Edison Company. The information contained in this report is believed ,

by General Electric to be an accurate and true representation of the facts-  ;

known, obtained or provided to General Electric at the time this report was prepared.

j Boston Edison Co. is advised that the information contained in this [

report will not be incorporated in the Pilgrim Nuclear Power Station Reload f

License Submittals or other associated GE documentation. Further, the i specific operating conditions, features or results pertaining to the analyses l reported herein will not be incorporated in other services provided to Boston f

Edison Co. without Boston Edison Company's direction to do so. In the event -!

. the implementation of these results requires technical specification changes f i

J or changes in other services provided to Boston Edison Co., General Electric l j should be notified immediately in order to assure proper consideration of  !

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, these changes. l b

The only undertakings of the General Electric Company respecting infor-

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mation in this document are contained in the contract between Boston Edison -

Company and the General Electric Company governing Purchase Order No. 63430, l dated February 27, 1986, and nothing contained in this document shall be con-strued as changing that contract. The use of this information except as  !

defined by said contract, or for any purpose other than that-for which it is intended, is not authorized; and with respect.to any unauthorized use, neither General Electric Company nor any of the contributors to this document makes any representation or warranty (express or implied) as to the completeness,-  !

accuracy or usefulness of the information contained in this document or that such use of such information may not infringe privately owned rights; nor do they assume any responsibility for liability or damage of any kind which may l c

result from such use of such information. l

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CONTENTS-  ;

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1. INTRODUCTION 1-1

2.

SUMMARY

AND CONCLUSIONS 2-1  !

3. PILGRIM NUCLEAR POWER STATION MSIV LOW PRESSURE ISOLATION SETPOINT CHANGE ANALYSIS 3-1 _l 3.1 Description of MSIV Low Pressure Isolation Setpoint  !

Regulator Failure (Open) Event 3-1  ;

3.2 Methods of Analysis - Simulation of the Pressure -:

Regulator Failure (0 pen) Event 3-2  ;

3.3 Results of Pressure Regulator Failure (0 pen) Calculation 3-3 3.4 Effect on Extended Operatlig Domain 3-6

4. PILGRIM NUCLEAR POWER STATION FATIGUE ASSESSMENT FOR 750 PSIG MSIV ISOLATION ANALYTICAL LIMIT 4 I 4.1 Peak Stress for Pressure Regulator Failure Transient  !

with 750 psig MSIV Isolation Analytical Limit 1-1 j 4.2 Fatigue Usage Under Pressure Regulator Failure Transient 4-2 ,

5. OTHER SAFETY SAFETY CONSIDERATIONS FOR LOWERING MSIV ISOLATION SETPOINT 5-1 -l 5.1 Eadiological Consideration 5-1 .

5.2 MCPR and Future Fuel Cycle Considerations 5-1 j 5-3 5.3 Impact on Other Systems

6. REFERENCES 6-1 l.

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i ILLUSTRATIONS l

Figure Title Page 3-1 Pilgrim Nuclear Power Station System Response ,

to Pressure Regulator Failure (0 pen) Transient 3-4 3 3-2 Pressure Regulator Failure (0 pen) Transient Pilgrim  ;

Nuclear Power Station 3-5 i, 4 ,

5-1 Initial and Final Conditions-for Pressure Regulator l Failure Event 5-1 i i

5-2 Core Power Response to Pressure Regulator Failure 5-1 f k

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1. INTRODUCTION i The purpose of this analysis is to provide the technical basis for chang-  ;

ing the main steam isolation valve (MSIV) low pressure isolation analytical limit

• from the present value (880 psig) to 750 psig at the Pilgrim Nuclear Power Station (pNPS). As part of the Reactor Vessel Isolation Control System, the MSIV low pressure isolation setpoint is normally intended to effectively j prevent excessive vessel depressurization under high steam flow demand con- j ditions and ensure that excessive thermal gradients will not be imposed on vessel hardware. However, unwanted isolations and scrams can occur because of the limited margin between the current setpoint and the operating pressure at ,

the turbine inlet. Reducing the MSIV setpoint can decrease such occurrences f

and consequently improve plant availability and safety. I I

The pressure regulator failure (open) transient is evaluated, assuming an MSIV isolation analytical limit of 750 psig, to show that the thermal stress .l t

resulting from vessel cooldown is within allowable values and does not impact  ;

the reactor vessel lifetime fatigue usage. Core thermal margins and impact on l other systems associated with the 750 psig setpoint are also examined to dem-onstrate acceptability. ,

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• The term " analytical limit" is defined as the value of the sensed process i variable established as part of the safety analysis, prior to which a j desired action is to be initiated to prevent the process variable from  !

reaching the associated design safety limit. j 1-1/1-2 .

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2.

SUMMARY

AND CONCLUSIONS l I

A Pressure Regulator Failure (open) transient was simulated for PNPS with the MSIV low pressure isolation analytical limit set at 750 psig. The ,

calculated vessel cooldown results were analyzed and compared with the reactor ,

pressure vessel thermal cycle fatigue evaluation for PNPS. It was determined

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that lowering the MSIV isolation analytical limit to 750 psig7will not cause a ,j significant impact on the reactor vessel's lifetime fatique usage. Further, j there are no adverse effects on core thermal margins or other systems w tich utilize the isolation setpoint.

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It is concluded that the MSIV low pressure isolation analytical limit can  !

be lowered to 750 psig, and an overall safety benefit will be realized by a

, f reduction in unwanted isolations and attendant safety relief valve (SRV) l 1

actuations.  ;

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3. PILGRIM NUCLEAR POWER STATION MSIV LOW PRESSURE  !

ISOLATION SETPOINT CHANGE ANALYSIS  !

3.1 DESCRIPTION

OF MSIV LOW PRESSURE ISOLATION SETPOINT FUNCTION '

ii i- l As part of the Primary Containment and Reactor Vessel Isolation Control l System, PNPS has been equipped with an MSIV low pressure isolation function with a fixed pressure serpoint. The MSIVs automatically close when the pres-

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sure at the turbine inlet decreases below this preset pressure setpoint. The reactor scrams shortly af ter because of MSIV closure.. Thus, the reactor  !

I vessel low pressure isolation prevents excessive vessel depressurization in  !

the event of a malfunction of the nuclear system pressure regulator. As a  !

t result of the isolation, no significant thermal stresses are imposed on the j nuclear system process barrier. The isolation transforms the consequences of  ;

a vessel depressurization event, such as a pressure-regulator failure (open),

into a controlled, isolated shutdown event. This isolation function is not. i required to satisfy any of the safety design bases of this system. f i

The MSIV low pressure isolation setpoint was originally chosen at  ;]

approximately 100 psi less than the turbine inlet pressure. With this 100 psi  ![

or less operating margin, unwanted isolations and scrams could occur since h t

pressure regulators with built-in control time constants may not be able to  ;

limit the pressure drop before the isolation setpoint is reached.- The reduc- -l tion of the current limit to some lower setpoint value could potentially l 1

reduce the number of spurious scrams and unwanted MSIV isolations, conse-  !

quently improving plant availability.

To reduce the low pressure setpoint to some lower value, it should be  !

demonstrated through conservative engineering analysis that further depres-  ;

surization and cooldown of the vessel to below the current turbine inlet-l prescure setpoint limit would not cause any excessive thermal stress on the -l nuclear system components.

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NED0-31296 Based on the request of Boston Edison Co., an engineering analysis was l performed to justify the reduction of the MSIV low pressure isolation analyti-  !

t 1 cal limit from 880 psig to 750 psig. The approach was to evaluate the Pres- l t

sure Regulator Failure (open) transient, assuming an MSIV low pressure isola- i

tion analytical limit of 750 psig, to show that the thermal stresses are '

within allowable values and do not increase the reactor vessel component's ,

lifetime fatigue usage factor. The Pressure Regulator Failure (open) - f transient provides the vessel response to a fast vessel depressurzaztion which  !

is terminated by MSIV closure on low turbine inlet pressure. From the f

response, the vessel cooldown rate and resulting thermal stresses are  ;

i evaluated.

l 4 i 3.2 METHODS OF ANALYSIS - SIMULATION OF THE PRESSURE REGULATOR FAILURE (OPEN) .l EVENT l 4  :

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The turbine at PNPS is provided with two pressure regulators. These two l

regulctors have slightly different pressure setpoints such that one functions j j as a controlling pressure regulator and the other as a back-up pressure regu-l lator. The controlling pressure regulator is used to control both the turbine 'I

[ control valves and the turbine bypass valves to maintain constant turbine inlet pressure. If either the controlling pressure regulator or the back-up pressure regulator fails in an open direction, it will cause the main turbine r

control valves to respond by opening further, thus increasing steam flow and l dropping turbine inlet pressure until the MSIV closure (MSIVC) is reached and a reactor scram is initiated.

1 The depressurization of the reactor can also cause an increase of the 1

l bulk fluid void volume which can produce a level swell. If the resulting depressurization is rapid, the rising vessel water level may reach the high  ;

trip level (Level 8) before the turbine pressure drops to the MSIV low  ;

pressure setpoint. In this case, the high level trip initiates a main turbine  ;

stop valve closure (MTSVC) and possibly a feedwater pump trip. The MTSVC in j

turn initiates a reactor scram. With MSIVC or MTSVC (whichever occurs first) l the vessel depressurization will ultimately be terminated by the MSIV low  !

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pressure isolation. At this point, the reactor will be scrammed and iso- i lated. It can then be brought to a standard controlled shutdown with a con-l trolled vessel cooldown rate. l t

This analysis simulates a pressure regulator failure (open) event with. -t the MSIV low pressure isolation setpoint at the proposed 750 psig analytical limit. The GE thermal-hydraulic and nuclear kinetics coupled transient code i 1

REDY is used to evaluate the dynamic system response of the pressure regu-lator failure (open) event previously discussed. The calculations'are per-formed with the following basic assumptions and initial conditions.

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a. The initial reactor power is at 1998 MWt, corresponding to 100% of .

rated steam flow.

b initial dome pressure is 1050 psia.

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c. Initial core flow is at 100% of rated core flow.

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d Conservative end-of-cycle (EOC) scram, void and doppler reactivity .

curves are used, based on PNPS, Cycle 7, fuel loading conditions.

e. The MSIV closure time is assumed to be 5 seconds.

f. Ths turbine bypass valves with 25% capacity are open for faster i depressurization, providing conservative results.

g. The pressure regulator upper limit is set at 125% of initial steam flow demand.

3.3 RESULTS OF PRESSURE REGULATOR FAILURE (OPEN) CALCUIATION F

The results of the pressure regulator failure (open) transient'calcu- f lation are presented in Figures 3-1 and 3-2. Included in Figure 3-2 are the time histories of vessel steam flow, vessel dome pressure response and the I corresponding saturated steam temperature. Initially, the vessel steam flow j i

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Figure 3-1. Pilgrim Nuclear Power Station - System Response to Pressure Regulator Failure (0 pen) Transient i

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NED0-31296 increases rapidly as the turbine control valves open because of the pressure  ;

regulator failing in the open direction. With the high steam outflow, the  !

vessel pressure decreases, which results in vessel water level swell as the bulk fluid void volume is increased. At 5.0 seconds, the water level reaches the high water Level 8 setpoint. This initiates a closure of main turbine ,

t stop valve (MTSV) and a scram on MTSV position. A feedwater pump trip would also occur, but credit for it was not taken in this analysis because the ,

feedwater pump trip at PNPS is not qualified as safety grade, i Af ter the MTSV closure there is full steam discharge through the bypass system; however, the vessel begins to repressurize because the bypass capacity represents only 25% of the normal turbine steam flow. The pressure contin 2es j to rise until the rate of steam generation is reduced below the bypass flow .

capacity. At this point the vessel begins to depressurize; this depressur-ization continues until the low turbine inlet pressure setpoint is reached, -

initiating an MS1V closure at approximately 70 seconds l l

The initial reactor vessel temperature is at the saturated steam tempera- l ture of 550*F (Figure 3-2). The transient produces a maximum temperature in the steam dome of 558'F which occurs as the steam generation rate drops below f 1

the bypass flow capacity. The steam temperature in the dome then drops, '

l because of system depressurization, to a minimum value of 512*F at approxi-mately 70 seconds. The maximum change of 46*F in the steam dome saturation temperature provides the basis for the vessel thermal stress evaluation.

Following isolation, the system depressurization is terminated and the  !

vessel water level begins to fall as the system pressure increases. The j 4

reactor, being scrammed and isolated, can now be brought to a controlled. J shutdown.

3.4 EFFECT ON EXTENDED OPERATING DOMAIN j In general, transients initiated from operating conditions inside the j standard power flow map tend to be bounded by transients initiated at 100%  ;

i rated flow.

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. s NEDO-31296 PNPS is currently licensed with an extended operating domain which allows operation with increased core flow (up to 107.5% of rated core flow) and an extended load line limit (up to the 108% average power range monitor rod block line). This extended operating domain introduces two potentially limiting operating points outside of the standard power flow map. The limiting increased core flow point corresponds to 100% of rated power and 107.5% of rated core flow. The extended load line limit condition corresponds to 100%

of rated power and 87% of rated core flow. Note that both of these conditions only involve a deviation in the initial core flow.

The pressure regulator failure open transient is most severe when initi-ated from a high reactor power because of the increased steam generation rate. The steam generation rate governs the peak pressure reached following the turbine trip. The saturation temperature corresponding to this peak pressure sets the upper temperature limit for the thermal stress evaluation.

The lower limit is set by the saturation temperature corresponding to the pressure reached at the time of MSIV closure. This lower limit is not expected to change significantly because the MSIV closure is initiated from the same low turbine inlet pressure setpoint, regardless of the conditions  !

i f rom which the tranaient was initiated.

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It was anticipated that a change in the initial core flow would not significantly affect the peak pressure reached during the pressure regulator l I

event. However, to confirm this a sensitivity study was performed, initiating  ;

the transient from the increased core flow condition of 100% rated power and j 107.5% of rated core flow. The results indicated that the peak pressure was  !

impacted by less than 3 psi. This amounts to a temperature change of less than l'F. Therefore, it is concluded that the 100% power and flow conditions l t

previously presented are representative, in terms of AT, of all power flow  :

conditions including increased core flow and extended load line limit opera- j i

tion. Additionally, this conclusion is also true for final feedwater tempera- l 6

ture reductions performed at the EOC either with or without increased _ core j flow operation. l 3-7/3-8  !

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NED0-31296

4. PILGRIM NUCLEAR POWER STATION FATIGUE ASSESSMENT FOR 750 PSIG MSIV ISOLATION ANALYTICAL LIMIT 4.1 PEAK STRESS FOR PRESSURE REGULATOR FAILURE TRANSIENT WITH 750 PSIG MSIV ISOLATION ANALYTICAL LIMIT Lowering the MSIV low pressure isolation analytical limit to 750 psig can .

allow the reactor vessel to depressurize further, imposing a larger vessel temperature gradient during the transient. The impact of this temperature change on the vessel fatigue life can be evaluated based on conservative ther-mal stress calculations.

The most severe thermal stress experienced by any component material, which is undergoing a' temperature change within a finite amount of time, is bounded by the peak skin stress experienced by the material under a sudden temperature change of the same amount. This skin surface peak stress resulting from the. instantaneous temperature change can be expressed as:

3g , ATuE (1-v) where Ao =

peak stress induced at surface (psi) eT =

temperature change from thermal shock (*F) a = coefficient'of thermal expansion (in./in./*F)

E =

modulus of elasticity (psi) y =

Poisson's ratio 4-1 f

NED0-31296 Based on the maximum temperature change of 46*F, which the vessel com-ponent would experience under the pressure regulator failure (open) transient (Figure 3-2), the bounding thermal stress for stainless or carbon steel is calculated to be A# = 17800 psi.

750 psig MSIVC Analytical Limit This is a very conservative and bounding thermal stress value for the pressure regulator failure transient. This peak stress is classified in the ASME Pressure Vessel Code (Section III) as the stress for fatigue life evalu-ation, where an allowable stress limit is imposed for fatigue cycling consideration.

4.2 FATIGUE USAGE UNDER PRESSURE REGULATOR FAILURE TRANSIENT Fatigue cycling is classically described by a " usage factor," which is defined as I (o /Ng ) using a linear damage relationship; assuming that, if N gcycles would produce failure at a stress level Sg , then ng cycles at the same stress level would use up the fraction ng /N g of the total life. Failure is postulated to occur when the cumulative usage factor becomes 1.0.

Fatigue assessments '

of the PNPS pressure vessel components have been performed using postulated duty cycles based on a 40 yr plant operating -life.

The components analyzed which have the highest cumulative fatigue usage factor include the feedwater nozzle, recirculation inlet nozzle and recirculation outlet nozzle. The results of the aforementioned analyses provide a conve-nient basis for a conservative calculation of the fatigue usage because of the 750 psig isolation analytical limit. Incremental fatigue usage.for a particu-lar component is conservatively determined utilizing existing calculations which evolve from a transient which is more severe and, therefore, bounds the pressure regulator failure (open) transient. When this technique is not directly applicable, scaling techniques are applied to the nominal stresses and the 17,800 psi " skin" stress range is added conservatively as a peak 4-2

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stress. Both techniques assure a conservative calculation for component l fatigue usage. l

The PNPS reactor vessel component found to have the highest reported fatigue usage is the recirculation inlet nozzle. The dominating fatigue usage for the l component is due to an improper pump startup transient when the reactor is operating at a temperature of 522*F. '

The temperature differences and thermal

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l- stresses associated with this transient are considerable larger tb a those  !

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j caused by the pressure regulator failure (open) event. Therefore, a conserv-4 ative fatigue assessment is assured by enveloping the pressure regulator failure (open) transient, as previously described, with eight additional ,

i-cycles. (The pressure regulator failure event is limited to eight cycles in h 40 years.5) The incremental fatigue usage for the eight additional cycles is calculated to be less than 0.0003 for the recirculation inlet nozzle and ,

is, therefore, considered to be insignificant. The effects of increased core  !

flow combined with this fatigue analysis have also been evaluated and  !

determined to be insignificant. I i A similar result is obtained for the other reactor vessel components j analyzed. It was found that their lifetime fatigue usage: could not increase ,

appreciably, and all would remain at less than unity.  !

1 It is concluded, based on the preceding assessment, that lowering the i MSIV isolation analytical limit to 750 psig will not 'have a significant effect l 1 on the reactor vessel's lifetime fatigue usage.  ;

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5. OTHER SAFETY CONSIDERATIONS FOR LOWERING MSIV ISOLATION SETPOINT i

.i 5.1 RADIOLOGICAL CONSIDERATION f

The impact on the radiological release due to a reduction in the low tur-  ;

bine inlet pressure isolation setpoint has been considered. The radiological  ;

e release caused by the design basis steamline break outside the containment i i

will not be affected, because for this large a break, MSIV isolation is j i

assumed to occur as a result of high steamline flow rate, not turbine inlet .

pressure. Steamline breaks of a size small enough not to be detected by the I high steam flow signal are isolated either by temperature sensors in the steam tunnel or area radiation monitors in the turbine building. Therefore, it is l l

concluded that the radiological release will not be affected by reducing the low turbine inlet pressure isolation analytical limit to 750 psig.

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5.2 MCPR AND FUTURE FUEL CYCLE CONSIDERATIONS

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l The reduction of the MSIV low pressure isolation analytical limit from its current value to 750 psig will have no impact upon the minimum critical power ratio (MCPR) or fuel thermal margins. This can be demonstrated by com-- {

paring the scenario of two pressure regulator failure (open) transients,.one utilizing the current low pressure HSIV isolation analytical limit and the f other with the proposed value of 750 psig. In both cases the turbine control j

~f valves open, increasing vessel steam flow and decreasing vessel pressure. The.

'f vessel depressurization causes the water level to swell because of voiding. -

i When the water level reaches the high water level setpoint, a turbine trip

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occurs which initiates a reactor scram on turbine stop valve position.- Since j l

the bypass capacity represents only about 25% of normal turbine steam flow, i the vessel pressurizes after turbine stop valve closure until the steam gen-eration rate is reduced to below the bypass flow capacity. The vessel then depressurizes until the low pressure MSIV isolation setpoint is reached. By i f

the time of MSIV closure, the reactor is essentially shut down with a neutron j flux of less than 1% of rated. Up until the time that the current MSIV' iso-l' lation analytical limit is reached, the two transient responses are identical i

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I and, therefore, no differences in the MCPR response or fuel thermal margins-I are observed. After this time, the differences in the transient responses l t

will have no impact on MCPR or thermal margins because of the very low reactor i power. t t

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In the preceding scenario, the pressure regulator was assumed to fail j open to its maximum position instantaneously. Additional analyses have been. '

performed to evaluate a slow opening failure of the pressure regulator. The slow failure is postulated such that the water level does not rise high enough j to reach the high water level trip setpoint. Consequently, the reactor scram f I

will be delayed until the turbine inlet pressure drops to the isolation pressure setpoint and the MSIV closure initiates a reactor scram. It should i a

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be noted that this failure results in less severe vessel thermal duty because of the slower depressurization-rate and the lack of vessel repressurization previously initiated by the turbine stop valve closure. I Shown in Figure 5-1 are the initial conditions and'the approximate con-ditions ct the time the 750 psig isolation pressure setpoint is reached. The inherent decrease in core peak heat flux as the vessel pressure decreases is I shown in Figure 5-2. From these two figures, it is clear that reactor power

, decreases significantly, while the core flow remains near initial rated  ;

conditions.

The events previously postulated' inherently lead to an. increase in the fuel thermal margin because of the following: l

.h j a. Core thermal power decreases because of core voiding and eventually j as the result of scram, l

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b. Core inlet flow does not change significantly before scram.

]j c .- Critical power increases significantly with decreasing pressure over j the range of the transient. .!

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. e NEDO-31296 r As a result of the preceding conditions the MCPR or margin to the safety I limits actually increases substantially during tue vent. Therefore, it is f

concluded that reducing the MSIV low pressure isolation setpoint has no impact upon MPCR or fuel thermal limits.

f The present analysis is applicable to future fuel cycles since the MCPR 5 and fuel thermal limits are not a concern and the reactor response to this f

event is not sensitive to the cycle dependent core nuclear characteristics. i 5.3 IMPACT ON OTHER SYSTEMS i The isolation signal from the steam turbine inlet pressure will initiate ;

isolation of the following valves: j

a. Main steam isolation valve (MSIV) t

b. Main steam drain valve (MSDV) f i

c. Reactor water ss=ple valve (RWSV) {

i The isolation of the MSIV has been discussed earlier and is the main ,

objective of the safety evaluation. The' isolation of the RWSV at a lower i

steam turbine inlet pressure is acceptable because the RWSVs and MSDVs can be i opened from the control room. The setpoint for MSDV and RWSV isolation will not affect any of the design bases. Therefore, it is concluded that the 750 psig isolation setpoint is acceptable for PNPS. ,

i t

i 1

5-3 .

4 l

NEDO-31296 l P2 (steam dome)  !

L ji i

B P3 (turbine inlet) ,

P3 (core) f t

I APPROXIMATE INITIAL APPROXIMATE CONDmONS CONDITIONS AT TIME OF ISOLATION CORE THERMAL POWER 100% 69%

CORE FLOW 100 % 98%

Pj (psis) 1057 820 P2 (psia) 1050 812 P3 (psia) 994 765 l i.

Figure 5-1. Initial and Final Conditions for Pressure Regulator Failure Event s

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i NEDO-31296 i

6. REFERENCES f i

i

1. " Analytical Methods of Plant Transient Evaluations for the General .l r

Electric Boiling Water Reactor," General Electric Company, February 1983 j (NED0-10802) .  :

2. Formulas for Stress and Strain, Roark and Young, 5th Edition,.1975, I

'McG raw-Hill. f I

i

3. " Reactor Vessel Recirculation inlet and Outlet Nozzle Stress Report",

General Electric Company, December, 1984, (23A4084, Rev.0)  ;

i

4. " Reactor Vessel Feedwater Notzle Stress Report," General Electric j Company, December 1984, (22A5692, Rev.0)  !

5. " Data for National Reliability Evaluation Program" (NREP), A. J. Oswald i et al., EG&G Report, June, 1982 (EG&G-EA-5887)

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