ML20106C475

Application for Amend to Licenses NPF-4 & NPF-7,revising Tech Specs to Allow Positive Moderator Temp Coefficient at Reduced Power Levels in Reactivity Control Sys.Safety Evaluation Encl.Fee Paid

ML20106C475
Person / Time
Site:	North Anna
Issue date:	02/07/1985
From:	Stewart W VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	Harold Denton, John Miller Office of Nuclear Reactor Regulation
Shared Package
ML20106C480	List: ML20106C475 ML20106C499
References
666, NUDOCS 8502120279
Download: ML20106C475 (53)

Text

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y VIRGINIA ELECTRIC AND l'OWER COMPANY RIcnwown,VINGINIA 20261 D,",,,,Z February 7, 1985 Woctsam Ormaartone Mr. Harold R. Denton, Director Serial No. 666 Office of Nuclear Reactor Regulation PSE/J0E/mjp/2000N Attn: Mr. James R.' Miller, Chief Docket Nos. 50-338 Operating' Reactors Branch No. 3 50-339 Division of Licensing License Nos.: NPF-4 U. S.iNuclear Regulatory 'Comission NPF-7 Washington, D.C. 20555 Gentlemen:

AMENDMENT TO OPERATING LICENSES NPF-4 AND NPF-7 NORTH ANNA POWER STATION UNIT N05. 1 AND 2 PROPOSED TECHNICAL SPECIFICATION CHANGE

' Pursuant to-10CFR50.90, the Virginia Electric and Power Company requests an amendment, in the form of changes to the-Technical Specifications, to Operating License Nos. NPF-4 and NPF-7 for the North Anna Power Station Unit Nos. -1 and 2..

In our letter of' July -17,1984 (Serial No. 224A), we submitted the reload

!. information description for the fifth cycle core of North Anna Unit 1. In that submittal, we indicated that the moderator temperature coefficient for the hot zero power,.all-rods-out, beginning-of-life condition was calculated to be positive and .therefore initial escalation to power was.to be made with control rods inserted ;in the core in' order to maintain a non-positive moderator temperature coefficient. A non-positive moderator. temperature coefficient

_during normal' operation'is a basic assumption for the current UFSAR accident analyses for both the North Anna 1 and 2 cores and is therefore a current i Technical Specification requirement (TS 3.1.1.4).

Since our July 17, 1984 submittal, we have reanalyzed the relevant UFSAR-accidents for North Anna 1 and 2 in support of a positive moderator temperature coefficient at reduced power levels. By allowing a positive moderator temperature _ coefficient, the necessity of having the control rods significantly inserted in the core during initial startup and the potential ~for operating restrictions due to the delta flux limits associated with constant axial offset control are minimized. This'would allow greater flexibility in core designs for both North Anna units in future cycles. Enclosure 1 provides-the Safety

-Evaluation for-the proposed changes. The resulting specific Technical (Specification changes are given in Enclosure 2.

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(Mr.' Harold R. Denton.

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.. This Trequesti hasib'een reviewed 'and approved by the Station Nuclear Safety and Operating Conunittee and the Safety Evaluation and Control staff. It has m,

been determined.that .this request does. not involve any unreviewed safety

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equestions =asl defined"in 10CFR50.59 or a significant hazards consideration as

- defined in :10CFR50.92. if ',

We have evaluated this requestiin accordance with the criteria in

?10CFR170.12. L A check in the amount of $150 is enclosed as an application fee.

Very truly yours,

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-W. L. Stewart

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ures:

- (1)kSafety Evaluation for' Proposed ,Po'sitive F " Moderator Temperature Coefficient-(2); Proposed Technical Specification Changes (3)) Voucher Check for $150

. . .cc: Mr.? James' P. 0'Reilly

, ' Regional Administrator:

Region II -

Mr. Leon B. Engle.

NRC-Project Manageri .. North Anna

- Operating ' Reactors Branch No. 3 :

' Division of Licensing

- - Mr. M. W. ' Branch . .

.NRC' Resident. Inspector- ,

. North Anna Power Station

Mr." Charles Price

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Department of Health

-109 Governor Street-

Richmond, Virginia, 23219

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1-C9900NWEALTH OF VIRGINIA )

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CITY OF RICHMOND )

The foregoing document was acknowledged before me, in and for the City _ and Cosmonwealth aforesaid. . today by W. . L. Stewart who is Vice President -

Nuclear Operations, of the Virginia Electric and Power Company. He is duly authorized to execute ~and file the foregoing document in behalf of that j Company : and the statements in the document are true to the best of _ his knowledge and belief.

Acknowledged before me this 7 day of 19 II . ,

i Ny Commission expires: 4 - AS 19 II .

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Enclosure 1

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Safety Evaluation for A Positive Moderator Temperature Coefficient f

North Anna Power Station Unit Mos. 1 and 2  !

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PAGE 2 SECTION I INTRODUCTION I

iI. Introduction and purpose i

This- safety analysis has been performed to address the safety considerations :in allowing the North Anna Unit Mos. 1 and 2 to Loperate below . 70% power with a small, positive moderator temperature coefficient ~tMTC). The results of this study show that power.' operation with a positive moderator temperature coefficient, as alleued. by the attached proposed Technical Specifications changes, provides margin to UFSAR and other applicable safety.

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North Anna' Technical Specificationsodo not allow the units to- be brought' critical unless the moderator coefficient is negative, except during physics tests. This requirement is overly

, restrictive,- since allowance .of a.small' positive coefficient at

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reduced power levels would prcvide'significantly' increased fuel cycle flexibility, while Lonly causing a minor effect on safety

-analysis results presented in Ithe UFSAR. Nuclear design Ecalculations_ for recent -North . Anna cycles have indicated.that a

. positive -moderator -temperature coefficient may potentially be

< measured .at~ beg' inning of cycle, hot zero power. conditions with all

' control.' rods removed.from the core. Control rod insertion may be.

used- to make 'the_ coefficient negative, although plant startup is t

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m lengthened and made more complex by restrictions on boron ,

i concentration and centrol rod movement. However, to facilitate

. future- plant. startups, it is highly desirable to allow a slightly positive moderator ~ temperature coefficient at lower core power levels. As the. pouer : level is raised, the average core water temperature becomes- higher as allowed by the programmed average

. temperature for the plant, tending to bring the moderator

temperature coefficient more negative. Also, the boron concentration- can be reduced as xenon builds into the core. Thus,

.there is less need to allow a positive coefficient asifull power is approached. As fuel.burnup-is achieved, boron is further reduced

.andtthe moderator temperature coefficient will become negative over the entire operating power range. t The proposed Technical Specifications change, given in Enclosure 2, allows a +6 pcm/*F* MTC belou 70 percent of rated power, changing i

to. a '0 pcm/*F MTC at. 70 percent power and above. This MTC is depicted in Figure 1. A power-dependent MTC was chosen to minimize the effect of the MTC upon accidents initiated from high power

-lev 61s. Also, normal core. physical phenomena described above result in MTC becoming more negative as power level increases. This Technical Specifications change is expectedato provide a reasonable

' degree of flexibility in' core' design and plant operation for future 3 . cycles of North Anna Units 1 and- 2. The. proposed changes are similar to those which have been. approved for the Trojan and Turkey b ,

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Point plants. In addition, the Surry Unit Mos. 1 and 2 Technical Specifications allow a positive moderator temperature coefficient.

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.f SECTION II ACCIDENT ANALYSIS i

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~A. Introduction The impact of a positive moderator temperature coefficient for North ' Anna Units 1 and 2 on the accident analyses presented in i

Chapter' 15- of the- UFSAR(1I has been assessed. Those incidents which. were foundLto be sensitive to. minimum or near-zero moderator

-temperature coefficients were reanalyzed.. In general, these

' incidents. are limited -to transients which cause reactor coolant

. . tempera'ure-t to increase. The analyses presented herein were based on a +6 pcm/'F, moderator temperature coefficient, which was assumed.

.to remain- constant for variations in temperature. The assumption of -a - -positive moderator temperature coefficient existing at full l power His conservative since the' proposed Technical Specifications require .that ; the reactor .not be operated at full power if the 7

temperaturafcoefficient-is positive.

In gaaeral, Ethe : reanalysis was , based on 'the : tssumptions and methods

' employed :in :ther UFSAR; exceptions are noted in the discussion of eachi-incident. -The UFSAR basis referred to in this evaluation is

-for-; plant- operation at: 2775 MWt core. power with an RCS average

~. temperature .of 587.8 'F. This is-the. expected operating condition of both- North Anna units upon- implementation of the changes L::

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- supported by this report. Vapco computer codes were employed in the analysis of all accidents which were reanalyzed. The RETRAM code- (2,3,5) 'was used to obtain overall RCS parameter responses to the positive MTC. Core DNB analyses were performed with the COBRA code (4). Accidents not remnalyzed included those resulting in excessive heat removal from the reactor coolant system (for wh:.ch a large negative moderator temperature coefficient is conservative),

and those which experience heatup following a reactor trip (which are not sensitive to the moderator temperature coefficient). Table 1 . presents a list of accidents discussed in the North Anna UFSAR, and denotes those events reanalyzed for a positive coefficient.

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PAGE 7 B. . Transients Not Affected By a positive Moderator Temperature coefficient The following transients were not reanalyzed since they result in a reduction in reactor coolant system temperature, and are therefore not affected by a positive moderator temperature coefficient.

1. Rod-Cluster Control Assembly Misalignment The peak heat flux-following the drop of a control rod assembly is produced by action of the rod control system in response to the coolant average temperature decrease caused by an imbalance between core power and secondary system load. The existing analyses employed a-0.0 pcm/*F MTC, which maximizes the effect of the power overshoot for negative flux rate trip plants.

Since the limiting conditions for this accident are at or'near.

100X power- and- the proposed change. requires that MTC be less than 0.0 pcm/*F when above 70X power, this. accident is-not affected by the . proposed Technical Specification and th'a analysis was not repeated.

2.- Startup.of an Inactive Reactor. coolant Loop An- inadvertent startup of_ an idle reactor coolant pump.With loop. stop valves open results in the injection of cold water into 'the core. As the -most negative values .of' moderator

,- reactivity' . coefficient- . produce the greatest. reactivity addition,- the analysis reported in the UFSAR, Section 15.2.6, represents the limiting case. Startup of an inactive loop with loop. stop.-valves closed is effectively a boron dilution

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PAGE 8 accident, and will be discussed in Section II.C.

3. Excessive Heat Removal Due to Feedwater System Malfunctions The addition of excessive feedwater and inadvertent opening of the feedwater bypass valve are excessive heat removal incidents, and are consequently most sensitive to negative moderator temperature coefficients. Results presented in Section 15.2.10 of the UFSAR indicate'that the end of life case with a conservatively large negative moderator temperature coefficient results in the. minimum margin to DNB. Therefore, this incident was not zeanalyzed.

4. Excessive Load Increase An excessive load increase ' event, in which.the steam load exceeds .the core power, results in a decrease in reactor coolant system temperature. With the reactor in manual v control, the analysis presented in Section 15.2.11 of the UFSAR g

shows that the-limiting case is with a large negative moderator-temperature coefficient. 'If the reactor is in automatic

l. . control,' ~the control rods.are withdrawn to increase power and

'r'estore the average temperature to the programmed value. The

_ UFSAR analysis of this case'shows that the minimum DNBR is~not sensitive to 'the moderator temperature coefficient. The -UFSAR fanalysislcases are_therefore.still applicable to this incident.

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5. Loss of Mormal Feedwater, Loss of Offsite power to Station Auxiliaries The loss of normal feedwater and loss of offsite power accidents (Sections 15.2.8 and 15.2.9 of the UFSAR) are characterized by a gradual temperature rise due to decay heat e production and subsequent temperature reduction to the no load average value. A positive moderator temperature coefficient

.will not' affect these transients, since reactor trip occurs at the beginning of the transient, and the moderator reactivity coefficient -will become negative following control rod insertion. Therefore, there is no reduction in shutdown margin due to the heatup of the reactor coolont system.

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6. Accidental Depressurization of the Reactor Coolant System An' accidental _ depressurization of the reactor coolant system results .frome an inadvertent opening.of a_ pressurizer safety l valve (UFSAR Section 15.2.12). 'The most: limiting case assumes

.the reactor is- in automatic control, where the rod control system functions. 'to keep the power and- average coolant

temperature ~ essentially constunt until the reactor trip. This

!- portion of the transient is not sensitive to a positive t-

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moderator : temperature coefficient. Following-the reactor trip,

,, the _ average. coolant temperature decreases slowly. Thus, .the results presented in -the UFSAR represent _the most limiting conditions. Therefore, this~ transient was not reanalyzed with-

, :a-positive moderator temperature coefficient.

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7. _ Rupture of'a Main Steam pipe / Accidental Depressurization of the Main Steam System Since' the- rupture of a main steam pipe is a temperature red'ction transient, minimum core shutdown margin is associated

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' worst conditions for- a steamline break are therefore those analyzed in -UTSAR Section 15.4.2. Similarly, the accidental depressurization 'of the main steam system is a temperature

-reduction transient. A strong negative moderator temperature coefficient results in the minimum core shutdown margin. Thus, the. uor=t . condition =.for thi= transient are tho=c annly=ed in

UFSAR Section 15.2.13.

8. Spurious Operation of Safety Injection 4

Analysis ofL a spurious-operation of safety injection at power is presented in Section 15,2.14 of the UFSAR. This transient results 'n i a decrease ' in average coolant temperature and is g 'mostisensitive to a negative. moderator temperature coefficient.

Therefore,i this_ incident was not remnalyzed with a positive moderator' temperature _ coefficient.

~9. Rupture of~a' Main Feeduator pipe

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a mainifeedwater pipe accident (UFSAR Section

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The- rupture .of-15.4.2) is ~ analyzed to confirm the ability of the secondary system to -remove decay heat. This event.is not' sensitive to a'

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positive ~ moderator coefficient since the reactor trip occurs early in the transient before the reactor coolant system

temperature increases significantly. Thezefore, this event was

-not reanalyzed with a positive moderator temperature coefficient. -

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10. Loss of Coolant Accident (LOCA)

The loss of coolant accident (UFSAR Sections 15.3.1 and 15.4.1)

.is analyzedito determine the core heatup consequences caused by

a ! rupture of the reactor coolant system boundary. The event

-results in a depressurization of the RCS and a reactor shutdown

,4: _ at -the beginning .of the transient. This accident was not Jreanalyzed since the Technical specification requirement that

the moderator temperature coefficient be zero or negative at 70

, percent. power or above ensures that the previous analysis basis for th's'avant i is not affected.

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C. Transients Sensitive to a Positive Moderator coefficient The following incidents have been identified as being potentially sensitive to a positive moderator temperature coefficient, and the consequences of these incidents were reassessed.

1. Uncontrolled Boron Dilution

-For a boron dilution incident during refueling or startup, while the reactor is subcritical, Section 15.2.4 of the UFSAR shows that the operator has sufficient time to identify the problem and terminate the dilution before the reacter becomes critical. The UFSAR (Section 15.2.6) also shows that the operator has sufficient time to terminate a boron dilution during startup of an inactive loop. This incident is caused by violation of administrative procedures which require that boron concentration in the inactive loop be checked prior to opening the ' loop stop valve. These incidents- are therefore not affected'by the value-of the moderator' temperature coefficient.

'The reactivity addition due to a boron dilution at power-however. will cause ~ an increase in power and reactor coolant

. system temperature if the reactor is in manual control. Due to

~ the temperature' increase, a positive moderator temperature coefficient would add' additional' reactivity, and increase'the severity of the transients. However, this incident =is no more

, severe Ithan.'a- rod withdrawal at power, which is analyzed in this- section, and 'was therefore not specifically reanalyzed.

Following' reactor trip, the . amount of time available before shutdown margin Lis lost is- not- affected by the moderator b- _ _ _ . _ _ . . _ , . - _ . . - _ . . . .-.._.._ _ ._._--

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temperature coefficient.

2. Control Rod Withdrawal From a Subcritical Condition Introduction A control rod assembly withdrawal incident when the reactor is

.subcritical results in an uncontrolled addition of reactivity leading to a power excursion (Section 15.2.1 of the UFSAR). The nuclear power response is characterized by a very fast rise terminated by the reactivity feedback of the negative fuel temperature coefficient. .The power excursion causes a heatup of the moderator. However, since the power rise is rapid and is followed by an immediate reactor trip, the modarator

, temperature rise is small. Therefore, the transient is only moderately sensitive to the moderator temperature coefficient.

Method of Analysis The 'UFSAR states that for this transient, the. highest value of peak heat flux is produced for the highest rate of. reactivity insertion and lowest initial power. The analysis was reanalyzed. with a +6 pcm/'T MTC and the insertion rate of_75 x l10-5 dk/k / sec assumed in the UFSAR. The initial. power level,

, -reactor trip instrument delays and setpoint errors used in the

g . analysis. were. consistent with the UFSAR. As a result of recent Westinghouse concerns .related to the number of RC

Pumps allowed to be operating per the Technical Specifications, this analysis was performed for the more limiting case with one RC pump

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PAGE 14 operating. In addition, a more appropriate Doppler temperature coefficient which conservatively bounds current reload cycle values was used.

Results and~ Conclusions The nuclear power, coolant temperature, heat flux, fuel average temperature and clad temperature versus time for a 75 x 10-5 dk/k/sec insertion rate are shown in Figures 2-through 4.

Although the nuclear power exceeds the full power nominal value for a very short period of time, the peak heat flux, peak coolant . temperature and thermal power do not exceed nominal full power values. Since the heat flux does not exceed the

' nominal full power. value and remains bounded by the UFSAP, i: results, .the conclusions presented in the UFSAR are still applicable.- In addition, a detailed thermal hydraulic analysis 1has shown that thermal margin limits are met.

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3. Uncontrolled Control-Rod Assembly Withdrawal at power Introduction An uncontrolled control rod assembly withdrawal at- power

. produces a mismatch in steam flow and core power, resulting in an' ' increase in reactor coolant temperature. A positive moderator temperature. coefficient would augment the power

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i mismatch and could reduce the margin to DNB. A discussion of this incident is presented in section 15.2.2 of the UFSAR.

Method oflAnalysis The transient was reanalyzed employing the same assumptions regarding- initial conditions and instrumentation and setpoint errors used in the UFSAR. The analyses were only performed for a power- level of -102% of 2775 MWt, since this is the most limiting- case presented in the' existing plant analyses. A

-constant moderator temperature coefficient of +6 pcm/'F was used in'the analysis. The assumption that a positive moderator t

temperature coefficient -exists at full power is conservative.

'- since moderator the temperature coefficient will actually be I

zero;or negative at full power.

Results l.s Figure 5 shows the -minimum DNBR as a function of' reactivity l: -

! insertion rate for the full- power cases reanalyzed. The Elimiting case for DNB' margin is a reactivity insertion rate of ir l

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4.0 X 10-5 dk/k/sec, Using the conservative +6 pcm/*F moderator temperature coefficient, instead of the 0 pcm/'F limit allowed by the Technical Specification change, results in a minimum DNBR greater than the 1.30 limit value. This positive moderator temperature coefficient will therefore not lower the DNBR associated with a control rod assembly withdrawal at power below the design limit.

Conclusions These'results demonstrate that the conclusions presented in the UFSAR are still valid. That is, the core and reactor coolant system are not adversely affected since the nuclear flux and 0vertemperature Delta-T trips prevent the core minimum DNB L ratio from falling below 1.30 for this incident.

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4. . Loss of Reactor Coolant Flow-Introduction As demonstrated in UFSAR Section 15.3.4, the most severe loss of flow transient is caused by_ the simultaneous loss of I-electrical power to all three reactor coolant pumps. This

transient was reanalyzed to determine the effect of a_ positive imoderator temperature coefficient on the nuclear power

, transient .and_the-resultant effect on the minimum DNBR' reached during the incident.

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. Method-of Analysis Analysis methods and assumptions used in the reevaluation were consistent .with those employed in the UFSAR. The analysis uns Performed using a constant +6 pcm/'T moderator temperature coefficient coupled with the maximum Doppler temperature coefficient.

Results For the case remnalyzed, the reactor coolant average temperature increases less than 3'T above the initial value.

The' impact of the positive moderator coefficient on the nuclear power transient would be limited to the initial stages of the incident during- which the average reactor coolant temperature

. increases. This' increase is terminated shortly:after reactor trip. The. reactor coolant system response and the transient

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- DNBR - response .are similar .to those for the-UFSAR case which assumed.a.zero MTC. A lower DNBR value is expected as a result of the~ slight. increase..in the nuclear power' transient. However, l the- positive moderator temperature coefficient analysis with confirmed that the minimum DNBR- was greater- than~_ 1.30.

- Therefore, .the effect of the positive moderator temperature coefficient on this transient is acceptable. Figures 6 through the nuclear power and-heat flux

[ 8 shou' the flou 'coastdown, transients, and. the minimum DNB ratio vs. time'for the +6 I pcm/*F case.

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changes in the results of the complete loss of flow transient, and the minimum DNBR remains above 1.30 for this incident. This case was analyzed since it is the most limiting one presented in. the UFSAR.-Loss-of a single pump with all loops in service or with a single loop out of service and the loop stop valves open or closed were -less limiting. Since this type of transient causes only a small change in core average moderator temperature, and this change does not significantly affect the nuclear power transient, the single pump loss of flow cases are not appreciably affected and therefore remain less limiting.

5.' Locked Rotor Introduction The UFSAR (Section 15.4.4) shows that the most severe locked rotor incident is an instantaneous seizure of a reactor coolant pump rotor at 100X power with three loops operating. Following i

the. incident,' reactor coolant system temperature rises until l shortly .after reactor trip. A positive moderator temperature coefficient :will not affect the time to DNB since DNB is

. conservatively ' assumed to occur at _the beginning of the incident. The transient was reanalyzed, however, due to the potential effect on the nuclear power transient and thus on the peak reactor coolant system pressure and fuel temperatures.

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PAGE 19 Method.of~ Analysis The' analysis was performed for a +6pcm/*F moderator temperature coefficient. The initial conditions and assumptions used in

'this evaluation were consistent with those employed in the

-UTSAR. Tna RETRAN Hot Spot Model described in Reference 5.

using the locked rotor transient assumptions, was used to evaluate the-fuel. rod thermal transient.

Results Figures- ~ 9 through 12 show system response for the case rennaly=ed. .The nucicar peuar transiant was most =ignificantly affected by the positive moderator temperature coefficient..

peak nuclear - power- for this. case was 106% of nominal ~. This affect :resulted from the'use of a conservatively large bypass flou- fraction which was assumed in order to accentuate the pressure transient. This large bypass flow produces ant even greater core uater temperature, which causes the

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? reactivity contribution of the. positive moderator temperature

. coefficient to be overestimated.

Table 2 provides results consistent with those presented in the UFSAR.. The values of peak clad temperature and peak pressure

,, are~well below the applicable limit for this event.

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JConclusions' A positive moderator temperature coefficient does not adversely

. affect ~ .the- consequences of a locked rotor at full power with 1

-three -loops operating. -The integrity of the reactor coolant

. system .is not endangered as peak pressure during the transient is 2616 -psia. Since.a locked rotor with three loops operating 4 is; 'the limiting case present4d in the UFSAR, a positive moderator . temperature coefficient will also not significantly affect the consequences of the two loop operation cases.

- 6. Losst of External Electrical Load

. Introduction Two cases,. -analyzed' for both beginning and tend- of life conditions, are presented in Section 15.2.7 of the UFSAR

1. Reactor in manual rod controliuithioperation of the pressurizer spray and the pressurizer power operated

-relief valves; and 4.

2. ' Reactor in manual rod control with no credit for pressurizer spray or pressurizer power operated relief valv Since the moderator temperature coefficient will be negative at and .of life, only the beginning of life' cases were reanalyzed.

.The- result of a.~1oss of . load is - a core power level which

. momentarily exceeds. the secondary system power extraction causing 1 an-increase ~in core water' temperature. The consequences

< _ of this . reactivity addition du'e to- a- positive moderator

. temperature coefficient.-are increases in both peak nuclear

power and' pressurizer pressure.

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PAGE 21 Method-of' Analysis 1

A ' constant moderator temperature coefficient of +6 pcm/*F was assumed for the beginning:of-life cases reanalyzed. The method of analysis and assumptions used were otherwise in accordance with~ those presented in the UFSAR. These assumptions included

=an initial ~ reactor power and coolant temperature which were assumed - to be at their maximum values consistent with steady state. full power' operation, including allowances for calibration and instrument errors. The reactor coolant system pressure was assumed at its minimum value. The steam dump and u .t - direct. reactor trip on turbine trip were not assumed to e-operate.

4 Results

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System transient, response to- a total loss of load from 102X

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power, - at beginning 'of life, assuming pressurizer spray and F pressurizer . power operated relief valves,, is shown in' Figures

' 13 'and 14. The reactor trips on high pressurizer pressure, assumed to occur' at 2425 psias. pressurizer: pressure subsequently rises -to 2520' psia. The minimum DNBR decreases' lfrom_ its initial.value.-to-its minimum value' shortly after the reactor trip. The minimum DNBR remains greater than 1.30 during

-the. event.

Figures,15 and:16 illustrate reactor. coolant system response to a- loss Sf load' at beginning of life, assuming no credit for

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. minimum DNBR. increases from its initial value throughout the transient.

Conclusions The analysis demonstrates that the integrity of the core and the reactor coolant system pressure boundary during a loss of load ' transient will not be affected by a positive moderator reactivity' coefficient since'the minimum DMB ratio remains well above. the 1.3 limit, and the peak reactor coolant pressure is

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less than-110% of design. Therefore, the conclusions presented ' '

in'the UFSAR are'still applicable.

7. LRupture.of a Control Rod-Drive Mechanism Housing,. Control Rod

-Ejection-Introduction-i Thel rod ejection . transient is analyzed at full power and. hot Estandby- forf both beginning and and of life conditions. Since

-the- moderator temperature coefficient is negative at and of life, only .the beginning of life cases;were reanalyzed.

The

. re ac tivity .- addition increases nuclear power'and hot spot fuel

~ -

temperatures.

Mme-w=<4we t w 4 m W 0 -4 mWi

1

. . ... l PAGE 23 3

i

-Method of' Analysis l

'The method'of analysis is the same as reported in Reference 5.

The ejected- rod worths and transient peaking factors are the same as those in the UFSAR. The acceptance criteria are the same as the Westinghouse limit criteria, which are discussed in

. Reference- 5. Reference 6 provides the- basis for these criteria. The- values of significant' input parameters used in the. analysis are presented in Table 3. The moderator temperature- coefficient used was a constant +6 pcm/*r over the ranga Of coolant average temperature involved.

t- . Results and Conclusions peak 'f ue l- and clad temperatures and nuclear power versus time 2for'-both cases are presented ~in Figures 17 through 20. The

" limiting peak hot spot clad temperature, 2493*F, was reached in the h'ot- full power transient.. Maximum fuel temperatures were also associated with the full power case. Although peak hot spot' -fuel centerline 1tamparature reached-4900*F, the assumed melting . point of irradiated. fuel, melting was restricted to

' ~

less than'the innermost-10X of.the pellet.

' 1As ; peak fuel and clad-temperatures do not exceed the fuel and

-clad limits presented in Section 1.3 Eof the Vapco rod ejection

~

topicalI (5),- there is.no danger of. sudden fuel dispersal into

'the coolant, or consequential damage to the primary coolant Iloop. The results are summarized in Table 3.

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.r .

, , , , - - . . . . - . . - - - . . . - - . - - . . - . _ _ - . = . ... -

PAGE 24 SECTION III CONCLUSIONS l

To- assess the effect on accident analysis of operation of-North I

- Anna Units 1 and 2 with a slightly positive moderator temperatur6

. coefficient,-a safety analysis of transients sensitive to a zero or

. positive moderator coefficient was performed. These transients

- include'd. control rod assembly withdrawal from subcritical, control rod- assembly withdrawal at power, loss of reactor coolant flow, loss of-external load, locked rotor, and control rod ejection. This study- . indicated that. a small positive moderator temperature coefficient does not result in the violation of. safety limits for 7 -

tNe transients' analyzed.

- The analyses employed a. constant moderator temperature coefficient Hof : 1+6 'pcm/*F, independent of power level. The results of this study -are therefore ' conservative, since a1 positive moderator temperature coefficient is precluded by -the proposed Technical

- Specifications for full power operation.

LAnal'yses of the . transients in: Section 15 of the UTSAR that are

' affected -by the . change to a positive moderator temperature

- coefficient .have been performed to demonstrate that these transients meet the appropriate transient acceptance criteria. As

, such,. it ,can beLeoncludad that the_ change to a positive moderator temperature coefficient will not cause safety limits to be exceeded -

for any incident and consequently no unreviewed safety questions as defined. 'in 3 10CFR50.59 exist as la result of this proposed change.

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25-PAGE 25 ll The results of this evaluation can be stated as follous.

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=A

1. No increase in the probability of occurrence or consequences a-m of an accident will result from this proposed change. None SE of the plant systems will undergo physical changes for the {}

change to a positivie moderator temperature coefficient and yg therefore no change in the associated transient probabilities is expected.

lk -

T "E_

2. Since the proposed change causes no other system changes g (e.g., alterations in plant configuration), and given that "[

the effects upon system accident response are fully sea described by the parameters evaluated, operation with this ,se proposed change does not create the possibility of an j{

accident of different type than any evaluated previously in 20 the Safety Analysis Report. ;5*

.2l-

3. The margin of safety as defined in the bases for the II Technical Specifications is not reduced. The calculated EF safety parameters for the affected transients are all within the allouable limits for the respective transients. fI

=

SE It has been determined that the proposed change in moderator jh .

EL temperature coefficient does not pose a significant ha=ard jh d&

consideration. This is based upon example vi of those types of g{

sp .

license amendments that are considered unlikely to involve -at l

$N-significant hazards considerations (7). Example vi partially it-AW states, " A change which either may result in some increase to the EE hh probability or consequences of a previously analyzed accident or 'f may reduce in some way a safety margin, but where the results of the change are clearly within all acceptable criteria with respect  :

to the systems or component specified in the Standard Review plan." y{

Some analysis results do shou incremental increase in accident I bh consequences. However, the analysis results clearly shou'that all of the acceptance criteria for these types of transients are met w.

84 and the appropriate safety margins are maintained. ][

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PAGE 26 References

.i.

1) " Updated Final Safety Analysis Report - North Anna Power Station, Units 1 and 2," Docket Nos. 50-338, 50-339, Rev. 2, June 1984. ,

2) "Vepco Reactor System Transient Analysis Using the RETRAN 14 c+

Computer Code," VEP-FRD-41, March 1981. L

3) "RETRAN-02--A Program for Transient Thermal-Hydraulic Analysis of Complex Fluid Flow Systems," EPRI-NP-1850-CCM, May 1981.

.d(

4) "Vepco Reactor Core 'hermal-Hydraulic Analysis Using the Ib; y

COBRA IIIC/MIT Con er Code," VEP-FRD-33-A, October 1983.

5) "Vepco Evaluation of the Control Rod Ejection Transient,"

VEP-NFE-2, October 1983.

6) "An Evaluation of the Rod Ejection Accident ir Westinghouse PWR's Using Spatial Kinetics Methods", WCAP-7588, Rev. 1-A, January, 1975.

7) Federal Register, Vol. 48, No. 67, April 6, 1983, p. 14864

" Standards for Determining Whether License Amendments Involve No Significant Hazards Considerations," Interim Final Rule.

t l

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TABLE 1 ACCIDENTS EVALUATED FOR POSITIVE MODERATOR COEFFICIENT EFFECTS FSAR ACCIDENT TIME IN LIFE

• 15.2.1 RCCA Withdrawal from Subcritical BOC

• 15.2.2 RCCA Withdrawal from Power BOC/E0C 15.2.3 RCCA Misalignment / Drop BOC

• 15.2.4 Baron Dilution BOC I'

• 15.2.5/3.4 Loss of Flow B0C 15.2.6 Startup of an Inactive Loop E0C

• 15.2.7 Loss of Load / Turbine Trip BOC/E0C 15.2.8 Loss of Feedwater -

15. 0 9 Loss of Offsite Power -

A . 15.2.10 Feedwater Malfunction EOC

,'- , 15.2.11 Excessive Load Increase BOC/EOC 15.2.12 Accidental Depressurization of RCS BOC o

15.2.13/4.2 Steam Line Break E0C 15.2.14 Spurious Operation of SI B0C 8

15'.3.1/4.1 #LOCA. B0C

~ 15.4.2 Feed Lin't Break -

• c15.4.4 Locked Rctor BOC

• 15.4.6 RCC'A Ejection B0C/E0C t

• Accidents Reanalyzed B0C - Beginning of Cycle E0C - End of Cycle

I; . .. -

TABLE 2 SumARY OF 'RESULTS FOR LOCKED-ROTOR TRANSIENTS-Maximum primary coolant system 2616 pressure (psia)

Maximum clad temperature-(DF), 2273 core hot' spot Amount of Zr-H3 0 at core hot 1.374 spot (% by weight)

)

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- TABLE 3- <

' StPNARY OF R0D EJECTION ANALYSIS PARAETERS AND RESULTS' BEGINNING 0F CYCLE -

~ Power' Level,I

!102 ~ ~0 Ejected rod worth,J%Ak '0.20 0.878 Delayed neutron fraction, %. .52 .52 -

Feedback reactivity weighting 1.68 '3.23' Trip rod shutdown, %ak2 ., .4.0 ' 2.0 FQ before rod' ejection 2.52~ --

Fq after rod ejection' 7.07. 16.07 Number of operating pumps- 3. 2 Maximum fuel. pellet ~ average temperature,.OF 4046 '3502

- Maximum fuel center- temperature, OF 4904 4119 Maximum clad temperature,'0F. 2493 2486 Maximum fuel stored. energy, cal /gm .188' 150..

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FIGURE I J-

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. 30 40 50 60 70 80 90 100 0 10 20 l

- POWER (PERCENT) i I

4 .,

N00ERATOR TEMPERATURE COEFFIC1ENT VS. POWER LEVEL

l

+ . .

- 10 0 REACTIVITY INSERTION RATE = 75. x 10-5 AX/X/SEC NOTE: Neutron Flux Starts at 10-13 at Time Zero w

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Figure 2 Uncontrolled Rod Withdrawal from a Subcritical Condition, Neutron Flux versus Time

k. w - - - . - v -,r- ,-w- ---~re v-w,, ~vw-n-- w ,---w-~-~n- c, - - *- ew- -

.. .n . w - . . . _ _ . . .

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1 REACTIVITY INSERTION l RATE = 75. x 10-5 aK/K/SEC l f

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w jE s .2 t nc tsEcopos Figure 3 Uncontrolled Rod Withdrawal from a Subcritical condition Thermal Flux versus . Time T -*+e- ye e- w g- - g y#-- , 'w+ gy e ' d vymerwr- w e wr e,w-v

• wwyree-y,ww e v- w n --eme - w**y--w-, -www-+y--+-rweiWww--- - + sum-

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3000 ' I REACTIVITY INSEPTION RATE = 75. x 10-5 AK/K/SEC 900 -

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C 300 -

FUEL E

R w

b w 100 . -

Y y

CLAD 800 -

CORE WATER -

9 500 -

le 20 -

8 If O 4-fint tsECON081 Figure 4 Uncontrolled Rod Withdrawal from a Suberitical Condtion,

. Temperature versus Time

. .- . . . . - . . , . . . - - . . . . . - . - . . - . - . - . - . . - . - . _ - - - - . - - . - . ~ , - - - . -

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1. 7 - i 11 MININUM FEEDBACK.

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-- l

@ 1.5 -

'C-OVERTEMPERATURE r AT' TRIP E

.E HIGH NEUTRON E I

• 4 ' ._ FLUX TRIP

1.3 -

1.2' - - -

10-5 .: 2 4 .

6 8 10-4' 2 4 6 8 10-3 REACTIVITY: INSERTION RATE (oK/ K/SEC)

Figure 5 Effect of Peactivity Insertion Rate on Minimum DNBR for a Rod Withdrawal Accident from 100% power

i l

1.2.

.1.0 d

m:-

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5 0.6 -

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< .g. 0.4 d

0.2 0.0 0 2 4 '6 8 10 TIME'(SECONDS)

Figure 6 All Loops Operating, All Loops Coasting Down, Core Flow versus Time

c 1.2 1.0 <

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8

.5 z

s b HEAT FLUX

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5 NEUTRON p FLUX u

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FIGURE 7 All Loops Operating, All Loops Coasting Down, Flux Transients versis Time

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1.8 .

1.7 -

1.6' --

E'~ .1.5 -

5 1.'4 -

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0 .) 2' 3 4 5

'TIld'(SECONDS) t Figure 8 All Loops Operating, All Loops Coasting Down, DNBR versus Time

[.:

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/

'2500 - _/ \\

m / N w:

/ s 2400 _/ \

./

s-E y 'NN Y. \ .

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'2100 8 10 0' 2 4 6

- TIME (SECONDS)-

Figure ! 9 A11 Loops Operating One Locked Rotor, Pressure versus Time i.'.

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.4 1.0 5

g g 0.8 -

= -

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8. 0.2 0- 0 6 8 1 0- 2 4 TIE (SECONDS)

All Loops Doerating,- One Locked Rotor, Core Flow ver Figure 10 Time m.

Y 1 1.2 NUCLEAR POWER 1.0-4 0.8-5 HEAT FLUX A 0.6- -

8 p

j 0.4 --

.w d.2- _

0.0 4 6 8 1 0 0 2 TIME (SECONpS) 9 Figure 11: All Loops Operating, One Locked Rotor, Flux Transients versus Time

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.g 2000- N h; 1750 -

.h E 1500 <.

5 E

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g dL 1000 - '

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TIME (SECONDS)

Figure 12 All Loops Operating, One Locked Rotor, Clad Temperature versus Time

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t .= j,4 j i 1.2 - ! l 1.0 - l Eg 0.8 - ;

.E g 0.6 - l 0.4 - ;

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Figure 13 Loss of Load Accident, With Pressurizer Spray and Power Operated P.elief Valves, Beginning of Life L +

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Figure 14 . Loss of Load Accident, With Pressurizer Spray and Power Operated Relief Yalves, Beginning.of Life

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Figure 15 Loss of Load Accident, Without pressurizer Soray and Power Relief Valves, Beginning of Life

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(a 1200 -

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Figure 16 Loss of ' Load Accident, Without Pressurizer Spray and Power Operated Relief Valves, Beginning of Life

- . . . . . . ._.- -. -._..._ _ .._ . _ . -- . ---__.2. ,.

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_i Figure 17 Nuclear Power Transient BOC HFP Rod Ejectior. Accident a

4-

c. a.

l l

5000 " ~ ~ ~ ~ ~ ~ ~

' "4 Fuel Center Temperature 0 4900 F 4000 <

Fuel Avg Temperature o- -3000-Clad Temperature g

2000<

! 1000' l-

'O , , , , , , , , ,

6 7 8 9 10 2 3 4 5 0 1 l

TIME (SECONDS)

L Figure 18 Hot Spot Fuel and Clad Temperatures versus Time BOC HFP Rod Ejection Accident k

10 2. - 10 "

5 10 3 10 3 10 10 2 e

'N 2- 10 1

, 10-5 O

E p-o -

E 10 0 6- 10 ' a: m w

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10 10-1 10 10-2

' 10-3 10 9 - - . . .

0 .5 1.0 1.5 2.0 2.5 3.0 TIME (SECONDS)

Figure 19 Nuclear Power Transient BOC 11ZP Rod Ejection Accident

5000 ____

felting :

49000F 00 Fuel Center Temperature

^

g-Fuel Avg Temperature %_,'

- 3000 w /

g ,

y Clad Temperatu

___~

U 2000- I 1000 -

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2 3 4 5 0 1

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TI'E (SECONDS)

Figure 20 Hot Spot Fuel and Clad Temperature versus Time; BOC HZP Rod Ejection Accident

. '. , s, T

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ENCLOSURE 2 TECHNICAL SPECIFICATIONS CHANGES FOR

, A'+6 PCM/0F' MODERATOR TEMPERATURE COEFFICIENT NORTH ANNA POWER STATION UNIT NOS. 1 AND 2 i

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