ML20052F597

Proposed Changes to Tech Specs Re Moderator Temp Coefficient

ML20052F597
Person / Time
Site:	Trojan File:Portland General Electric icon.png
Issue date:	05/10/1982
From:	PORTLAND GENERAL ELECTRIC CO.
To:
Shared Package
ML20052F590	List: ML20052F589 ML20052F593 ML20052F597
References
NUDOCS 8205130226
Download: ML20052F597 (62)

Text

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r-LCA 87 Attachmint 1

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.8 1.0 1.2 540 FRACTION OF RATED THERMAL POWER l

REACTORCORESAFETYLIMIT-FduRLOOPSINOPERATION FIGURE 2.1-1 TROJAN UNIT 1 8205130226 820510 PDR ADOCK 05000344 PDR P

S SAFETY LIMITS BASES The curves are based on an enthalpy hot channel factor, F H, of 1.49 for axial power shape. An and a reference cosine with a peak of 1.55N at reduced power based on allowance is included for an increase in FaH the expression:

= 1.49[1 + 0.3(1-P)]

H where P is the fraction of RATED THERMAL POWER These limiting heat flux conditions are higher than those calculated for the range of all control rods fully withdrawn to the maximum allowable control rod insertion assuming the axial power imbalance is within the limits of the f(aI) function of the Overtemperature AT trip. When the l

axial power imbalance is not within the tolerance, the axial power imbalance effect on the Overtemperature AT trips will reduce the set-points to provide protection consistent with core safety limits.

2.1.2 REACTOR COOLANT SYSTEM PRESSURE I

The restriction of this Safety Limit protects the integrity of the Reactor Coolant System from overpressurization and therby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.

l The reactor pressure vessel and pressurizer are designed to Section III of the ASME Code for Nuclear Power Plant which permits a maximum transient pressure of 110% (2735 psig) of design pressure. The Reactor Coolant System piping, valves and fittings, are designed to ANSI B 31.7-1969, which permits a maximum transient pressure of 120%

(2985 psig) of component design pressure. The Safety Limit of 2735 psig is therefore consistent with the design criteria and associated code requirements.

The entire Reactor Coolant System is hydrotested at 3107 psig to demonstrate integrity prior to initial operation.

f TROJAN-UNIT 1 B 2-2 Amendment No. #8 l

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REACTIVITY CONTROL SYSTEMS MODERATOR TEMPERATURE COEFFICIENT LIMITING CONDITION FOR OPERATION 3.1.1.4 The moderator te.perature coefficient (MTC) shall be:

<0.5 x 10~4 a.

Ak/k/*F below 70 percent RATED THERMAL POWER.

<0.0 x 10-4 Ak/k/*F at or above 70 percent RATED THERMAL POWER.

Less negative than -5.36 x 10-4 Ak/k/*F at RATED THERMAL b.

POWER.

APPLICABILITY: MODES 1 and 2*#

ACTION:

With the moderator temperature coefficient outside any one of the above limits, be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.1.1.4.1 The MTC shall be determined to be within its limits by confirma-tory measurements. MTC measured values shall be extrapolated and/or compensated to permit direct comparison with the above limits.

4.1.1.4.2 The MTC shall be determined at the following frequencies and THERMAL POWER conditions during each fuel cycle:

Prior to initial operation above 5% of RATED THERMAL POWER, a.

after each fuel loading.

t l

b.

At any THERMAL POWER, within 7 EFPD after reaching a RATED THERMAL POWER equilibrium boron concentration of 300 ppm.

• With K 1.0.

1. SeeSpb>alTestException3.10.4.

i l

TROJ AN-UNIT 1 3/4 1-5

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS 4.2.2.1 The provisions of Specification 4.0.4 are not applicable.

4.2.2.2 F*Y shall be evaluated to determine if F (Z) is within its 0

limit by:

Using the movable incore detectors to obtain a power distribu-a.

tion map at any THERMAL POWER greater than 5% of RATED THERMAL POWER.

b.

Increasing the measured F component of the power distribution xy map by 3% to account for manufacturing tolerances and further increasing the value by 5% to account for measurement uncertainties.

C c.

Comparing the F computed (Fx ) obtained in b, above to:

xy 1.

The F limits for RATED THERMAL POWER (FRW ) for the approhiatemeasuredcoreplanesgivenineandfbelow, and 2.

The relationship:

R F

=F

[1 + 0.3(1-P)]

x x

b where F is the limit for fractional THERMAL POWER RTP operatioNexpressedasafunctionofF and P is the fraction of RATED THERMAL POWER at$hich F*Y was measured.

d.

Remeasurirg F according to the following schedule:

xy C

RTP 1.

When F is greater than the F limit for the appropriate measured core plane but less th n the F ' relationship, adgitional power gtribution maps shalFbe taken and and Fxy :

F compared to F,

xy x

a) Either within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after exceeding by 20% of RATED THERMAL POWER or greater, the THERMAL POWER at which F was last determined, or x

TROJ AN-UNIT 1 3/4 2-6 Amendment No. 5

,1 POWER DISTRIBUTION LIMITS l

SURVEILLANCE REQUIREMENTS (Continued) b)

At least once per 31 EFPD, whichever occurs first.

When the Fj is less than or equal to the F,TP R

limit for 2.

the appropriate measured core plane, additional power distribution maps shall be taken and F compared to F{P and F ) at least once per 31 EFPD.

x e.

Changes in the F limits for RATED THERMAL POWER (F P) xy shall be provided for all core planes containing bank "D" control rods and all unrodded core planes in a Radial Peaking Factor Limit Report per Specification 6.9.1.10.

f.

The F limits of e, above, are not applicable in the fol-xy lowing core plane regions as measured in percent of core height from the bottom of the fuel:

1.

Lower core region from 0 to 15%, inclusive.

2.

Upper core region from 85 to 100% inclusive.

3.

Grid plane regions at 17.8 + 2%, 32.1 + 2%, 47.4 + 2%,

60.6 + 2% and 74.9 + 2%, inclusive.

4.

Core plane regions within + 2% of core height (+ 2.88 inches) about the bank demand position of the bank "D" rods.

g.

Evaluating the effects of F on F (Z) to determine if F (Z) xy g

g is within its limit whenever F CexceedsFh.

x x

4.2.2.3 F (Z) shall be measured at least once per 31 EFPD. When F (Z) n n

is measured, an overall measured value shall be obtained from a power distribution map and increased by 3% to account for manufacturing tolerances and further increased by 5% to account for measurement uncertainty.

TROJAN-UNIT 1 3/4 2-6a Amendment No. 6, N, ##

POWER DISTRIBUTION LIMITS RCS FLOWRATE AND F R LIMITING CONDITION FOR OPERATION 3.2.3 The combination of indicated Reactor Coolant System (RCS) total flow rate and F shall be maintained within the region of allowable R

operation (above and to the left of the line) shown on Figures 3.2-3 and 3.2-4 for 4-and 3-loop operation, respectively.

Where:

N FAH

' and a.

FR " 1.49L1.0 + 0.3(1.0 - P )J THERMAL POWER b.

P = RATED THERMAL POWER APPLICABILITY: MODE 1 ACTION:

outside the region With the combination of RCS total flow rate and Fp of acceptable operation shown on Figure 3.2-3 or 3.2-4 (as applicable):

a.

Within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s:

1.

Either restore the combination of RCS flow rate and FR to within the above limits, or 2.

Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER and reduce the Power Range Neutron Flux - High trip setpoint to <55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.-

b.

Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of initially being outside the above limits, verify through incore flux mapping and RCS total flow rate comparison that the combination of F. and RCS total flow g

rate are restored to within the above limits, or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

t l

l TROJ AN-UNIT 1 3/4 2-8 Amendment No. 30, ##, (8 1

POWER DISTRIBUTION LIMITS ACTION:

(Continued)

Identify and correct the cause of the out-of-limit condition c.

prior to increasing THERMAL POWER above the reduced THERMAL POWER limit required by ACTION items a.2 and/or b above; subsequent POWER OPERATION may proceed provided that the combination of R and indicated RCS total flow rate are demonstrated, through incore flux mapping and RCS total flow rate comparison, to be within the region of acceptable opera-tion shown on Figure 3.2-3 or 3.2-4 (as applicable) prior to exceeding the following THERMAL POWER levels:

1.

A nominal 50% of RATED THERMAL POWER, 2.

A nominal 75% of RATED THERMAL POWER, and 3.

Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of attaining >95% of RATED THERMAL POWER.

SURVEILLANCE REQUIREMENTS 4.2.3.1 The provisions of Specification 4.0.4 are not applicable.

4.2.3.2 The combination of indicated RCS total flow rate and FR shall be determined to be within the region of acceptable operation of Figure 3.2-3 or 3.2-4 (as applicable):

Prior to operation above 75% of RATED THERMAL POWER after each a.

fuel loading, and b.

At least once per 31 Effective Full Power Days.

l Where:

N i

FAH

, and R " 1.49 Ll.0 + 0.3 (1.0 - P )J F

= Measured values of F$Hobtainedbyusingthemovable H

incore detectors to obtain a power distribution map.

The measured values of F{g shall be used to calculate FR since Figures 3.2-3 and 3.2-4 include measurement calculational uncertainties of 3.5% for flow and I

4% for incore measurement of F{g, 4.2.3.3 The RCS total flow rate indicators shall be subjected to a CHANNEL CALIBRATION at least once per 18 months.

4.2.3.4 The RCS total flow rate shall be determined by measurement at least once per 18 months.

TROJ AN-UNIT 1 3/4 2-9 Amendment No. #$

1

48 Measurement uncertainties of 3.5Y.

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3g 7CJAN-UNIT I 3/.1 2 9a Amenc.en-No. y e

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730JAN. UNIT 1 3/4 2 9b Amend:nent No. //

POWER DISTRIBUTION LIMITS BASES 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR, RCS FLOWRATE,'AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR The limits on heat flux hot channel hctor, RCS flowrate, and nuclear enthalpy rise hot channel factor ensure that 1) the design limits on peak local power density and minimum DNBR are not exceeded and 2) in the event of a LOCA the peak fuel clad temperature will not exceed the 2200'F ECCS acceptance criteria limit.

Each of these is measurable but will nomally only be determihed periodically as specified in Specifications 4.2.2 and 4.2.3.

This periodic surveillance is sufficient to insure that the limits are maintained provided:

a.

Control rods in a single group move together with no individual rod insertion differing by more than + 12 steps from the group demand position.

b.

Control rod groups are sequenced with overlapping groups as described in Specification 3.1.3.5.

c.

The control rod insertion limits of Specifications 3.1.3.5 are maintained.

d.

The axial power distribution, expressed in tems of AXIAL FLUX DIFFERENCE, is maintained within the limits.

FWH will be maintained within its limits provided conditions a.

through d. above are maintained. As noted on Figures 3.2-3 and 3.2-4, RCS flow rate and F$g may be " traded off" against one another (i.e., a low measured RCS flow rate is acceptable if the measured F$g is also low) to ensure that the calculated DNBR will not be below the design DNBR value. This tradeoff is allowed up to a maximum F{g of 1.49 (1+0.3(1-P))

l which is consistent with the initial conditions assumed for the LOCA analysis.

The relaxation of FyH as a function of THERMAL POWER allows changes in the radial power shape for all pemissible rod insertion limits.

When an Fg measurement is taken, both experimental error and manufactur-ing tolerance must be allowed for. 5% is the appropriate allowance for a full core map taken with the incore detector flux mapping system and 3% is the appropriate allowance for manufacturing tolerance. Application of these two penalties in a multiplication fashion is sufficient to provide a correction for the effect of rod bow on F, which has been conservatively estimated 0

as 5% in WCAP-8692, " Fuel Rod Bowing". The appropriate statistical combina-tion of local power, manufacturing tolerance and rod bow uncertainties, results l

in a penalty on F0 of 7.68%, whereas multiplying measured values of Fg by 1.03 x 1.05 results in a penalty of 8.15%.

TROJAN-UNIT 1 B 3/4 2-4 Amendment No. 30, #S, 70 l

ADMINISTRATIVE CONTROLS additicnal narrative material to provide complete explanation of the circumstances surrounding the event.

Reactor protection system or engineered safety feature instru-a.

ment settings which are found to be less conservative than those established by the technical specifications but which do not prevent the fulfillment of the functional requirements of affected systems.

b.

Conditions leading to operation in a degraded mode permitted by a limiting condition for operation or plant shutdown required by a limiting condition for operation.

Observed inadequacies in the implementation of administrative c.

or procedural controls which threaten to cause reduction of degree of redundancy provided in reactor protection systems or engineered safety feature systems.

d.

Abnomal degradation of systems other than those specified in 6.9.1.8.c above designed to contain radioactive material resultinc #-om the fission process.

.s IMIT REFORT RADIAL PEAKING I 6.9.1.10 Changes in the F Itnits for Rated Thermal Power (FRTP) xy shall be provided to the Director of the Regional Office of Inspection and Enforcement, with a copy to the Director, Nuclear Reactor Regulations, Attention Chief of the Core Performance Branch, U. S. Nuclear Regulatory Commission, Washington, D.C. 20555 for all core pisnes containing bank"D" control rods and all unrodded core planes at least 30 days prior to cycle initial criticality.

In the event that the limit would be submitted at some other time during core life, it will be submitted 30 days prior to the date the limit would become effective unless otherwise exempted by the Commission.

P Any infomation needed to support F will be by request fram the NRC and need not be included in this report.

SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Director of the Office of Inspection and Enforcement Regional Office within the time period specified for each report. These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference. specification:

Inoperable Seismic Monitoring Instrumentation, Speci-d.

fication 3.3.3.3.

TROJAN-UNIT 1 6-18 Amendment No. 56 h

i

b.

Inoperable Meteorological Monitoring Instrumentation, Specification 3.3.3.4.

c.

Inservice Inspection Program Reviews, Specifications 4.4.10.1 and 4.4.10.2.

d.

ECCS Actuation, Specifications 3.5.2 and 3.5.3.

e.

Sealed Source Leakage in excess of limits, Specifica-tion 4.7.7.1.3.

f.

Seismic Event Analysis, Specification 4.3.3.3.2.

g.

Fire Detection Instrumentation, Specification 3.3.3.7.

h.

Fire Suppression Systems, Specifications 3.7.8.1 and 3.7.8.2.

1.

Accident Monitoring Instrumentation, Specification 3.3.3.9.

j. Control Building Modification Connection Bolts, Specifications 3.7.11 and 4.7.11.1.

l TROJ AN-UNIT 1 6-18a l

LCA 87 Attechm2nt 2 RADIAL PEAKING FACTOR LIMIT REPORT This Radial Peaking Factor Limit Report is provided in accordance with Paragraph 6.9.1.10 of the Tro.ian Nuclear Plant Technical Specifications.

The Fxy limits for RATED THERV.AL POWER within specific core planes for Cycle 5 shall be:

RTP 1.

F less than or equal to 1.71 for all core planes containing bank "D" control rods, and RTP 2.

F less than or equal to 1.65 for all unrodded core planes.

These F (z) limits were used to confirm that the heat flux hot channel factor (z) will be limited to the Technical Specification values of:

[2 32]

[K(z)]

for P > 0.5 and, F (z) 3, g

p F (z) 3,

[4.64]

[K(z)]

for P 3, 0. 5 n

assuming the most limiting axial power distributions expected to result from the insertion and removal of control banks C and D during operation, including the accompanying variations in the axial xenon and power distri-butions as described in the " Power Distribution Control and Load Following Procedures", WCAP-8385, September,1974. Therefore, these Fxy limits provide assurance that the initial conditions assumed in the LOCA analysis are met, along with the ECCS acceptance criteria of 100FR50.46.

l t

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J WESTINGHOUSE PROPRIETARY CMSSi[

LCA 87 Attachm:nt 3 SAFETY ANALYSIS FOR OPERATION OF TROJAN WUCLEAR PLANT WITH A POSITIVE MODERATOR COEFFICIENT SE.CTION I INTRODUCTION, I.

Introduction and Purpose This safety analysis' has been performed to support the proposed Technical Specifications change for Trojan which would allow a small, positive moderator temperature coefficient to exist at power levels below 70 percent power. The results of the analysis, which are presented in the following section, show that the proposed change can be accommodated with margin to applicable FSAR safety limits.

The present Trojan Technical Specifications require the moderator temperature coefficient (MTC) to be zero or negative at all times

~

while the reactor is critical. This requirement is overly restric-tive, since a small positive coefficient at reduced power levels could result in a significant increase in fuel cycle flexibility, but would' have only a minor affect on the safety analysis of the accident events presented in the FSAR.

The proposed Technical Specifica'tions change, given in Appendix A, allows a +5 pcm/*F* MTC below 70 percent of rated power, changing to a O pcm/*F MTC at 70 percent power and above. This MTC is diagrammed in Figure 1.

A power-level dependent FUC was chosen to minimize the effect of the specification on postulated accidents at high power levels.

Moreover, as the power 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 coefficient more negative.

Also, the baron concentration can be reduced as xenon builds into the core. Thus, there is less

' I pca = 13-5 ;k/%

need to allow a positive coefficient as full power is approached.

As fuel burnup is achieved, boron is further reduced and the moderator coeff.icient will' become negative over the entire opera-

ing power range.

O S

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f SECTION II AC IDENT ANALYSIS I.

Intr: duction The impact of a positive coderator temperature coefficient on the accident analyses presented in Chapter 15 of the Trojan FSAR (1) has been assessed. Those incidents which were found to be sensi-tive to minimum or near-zero moderator coefficients were reana-lyzed.

In general, these incidents are limited to transients which cause reae. tor coolant temperature to increase. With one exception, the analyses presented hereth were based on a +5 pcm/*F* moderator temperature coefficient, which was assumed to remain constant for variations in temperature.

The control rod ejection analysis was based on a coefficient which was at least +5 pcm/*F at zero power nominal average temperature, This was and which became less ' positive for higher temperatures necessary since the TWINKLE computer code, on which the analysis is based, is a diffusion-theeny> cede ratheMherwa-point-kinetics approximation and the moderator temperature feedback cannot be For all accidents artificially held constant with temperature.

which were reanalyzed, the assumption of a positive moderator tem-perature coefficient existing at full power is conservative since as shown in Appendix A, the proposed Technical Specification requires that the coefficient be zero or negative at or above 70 percent power.

In general, reanalysis was based on the identical analysis rathods, computer codes, and assumptions employed in the FSAR; any excep-l tions are noted in the discussion of each incident. Accidents not l

reanalyzed included those resulting in excessive heat removal from t

the reactor coolant system for which a large negative moderator

~

l coefficient is conservative, and those for which heatup l

1 pcm = 1.0 x 10-5 sk/k.

e

-e i

l effects following reactor trip are investigated, which are not sensitive to the moderator coefficient.

Table I gives a list of accidents presented in the Trojan FSAR, and denotes those events reanaly:ed for a positive coefficient.

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I 1

e l

1 4

9 e

e 9

e o

G a

,-,-n

,.,,,-_.,,_--,n

r II. T ra nsi ent: Not Affected By a Po ':fve Moderator Coef ficient The following transients weri act reanalyzed since they either esu!! in a ' reduction 11 reactor coolant system temperature, and are therefore sensitive to a negative moderator temperature coeffi-cient, or are otherwise not affected by a positive moderator tem-perature coefficient.

A.

RCCA Misalignment / Drop Only the RCCA drop case presented in Section 15.2.3 of the FSAR and subsequent reload safety analyses (2,3,4) is potentially affected by a positive moderator temperature coef ficient. Use of a positive coefficient in the analysis would result in a larger reduction in core power level following the RCCA drop, thereby increasing the probability of a reactor trip. For the

[

return to power automatic rod contro,1 case, a positive coeffi-cient (which would only exist below 70 percent power) would result in a small increase in the power overshoot. Since, the limiting conditions for this accident are at or near 100 per-cent power where the moderator temperature coef ficient must be zero or negative, this accident is unaffected by the proposed Technical Specification and thus the analysis was not repeated.

I i

B.

Startup of an Inactive Reactor Coolant Loop

~

An inadvertent' startup of an idle reactor coolant pump results in a decrease in core average temperature. As the most nega-tive values of moderator reactivity coef ficient produce the greatest reactivity addition, the most limiting case is repre-sented by the analysis reported in the FSAR, Section 15.2.6, and the Cycle 2 Reload Safety Evaluation (2).

1ACOs

4 C.

Excessive Heat Removal Due to Feedwater System Malfunctions.

The socition of excessive feedwater or the reduction of feed-water teccerature are excessive heat removal incidents, and are consequently most sensitive to a negative moderator temperature coef ficient.

Results presented in Section 15.2.10 of the FSAR and the Cycle 2 Reload Safety Evaluation (2), based on a nega-tive coefficient, represent the limiting case. Therefore, this incident was not reanalyzed.

D.

Excessive Load Increase An excessive load increase event, in which the steam load exceeds the core power, results in a decrease in reactor cool-ant system temperature. With the raector in manual control, the analysis presented in Section 15.2.11 of the FSAR and the Cycle 2 Reload Safety Evaluation (2) shows that the limiting If the case is with a large negative moderator coefficient.

~

reactor is in automatic control, the control rods are withdrawn to increase power and restore the average temperature to the programmed valoe. The analysis o,f this case in the-FSAR show that the minimum DNBR is not sensitive to moderator temperature coef ficient. Therefore, the results presented in the FSAR are still applicable to this incident.

E.

Loss of Normal Feedwater, Loss of Offsite Power l

The loss of normal feedwater and loss of offsite power acci-dents (Sections 15.2.8 and 15.2.9 of the FSAR) are analyzed to determine the ability of the secondary system to remove decay These events are not sensitive to a positive moderator heat.

coefficient since the reactor trip occurs at the beginning of the transient before the reactor coolant system temperature l

increases significantly. Therefore, tnese events were not reanalyzed.

F.

Accidental Depressurization of the Reactor Coolant System An accidental depressurication af the res: tor coolant system results from an inadvertent opening of a pressurizer safety valve (FSAR Se: tion 15.2.12). The most liiniting case assumes the reactor is in automatic control, where the rod control system functions to keep the power and average coolant teapera-ture essentially constant until reactor trip. This portion of

~

the transient is insensitive to a positive moderator tempera-ture coefficient. Following reactor trip, the average coolant temperature decreases slowly. Therefore, the results presented i'n the FSAR represent the most limiting conditions, and this incident was not reanalyzed with a positive moderator tempera-ture coefficient.

Spurious Actuation of Safety Injection G.

Analysis of 'a spurious actuation of safety injection at. power is presented in Section 15.2.14 of the FSAit. This transient results in a decrease in average coolant temperature and is most sensitive to a negative moderator temperature coeffi-Therefore, this incident was not reanalyzed with a cient.

positive moderator coefficient.

H.

Rupture of a Main Steam Pipe Since the rupt'ure of a main steam pipe is a temperature reduc-tion transient, minimum core shutdown margin is associated with The worst a strong negative moderator temperature coefficient.

conditions for a steamline break are therefore those analyzed in the FSAR (Section 15.2.13 and 15.4.2).

I.

Rupture of'a Main Feedwater Pipe The rupture of,a main feedwater pipe accident (FSAR Section 15.4.2) is analyzed to determine the ability of the secondary This event is not sensitive to a system to remove decay heat.

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positive moderator coefficient since the reactor trd; occurs at the beginning of tne transient before the reactor esslant system temperature increases significantly.

Therefore, this event was not reanalyze,d with a positive moderator coefficient.

J.

Loss of Coolant Accident (LOCA)

The loss of coolant accident (Sections 15.3 and 15.4 of the FSAR) is analyzed to determine the core heatup consequences caused by a rupture of the reactor coolant system boundary.

The event results in a depressurization of the RCS ard a reactor shutdown at the beginning of the transient. This acci-dent was not reanalyzed since the Technical Specification requirement that the temperature coefficient be zero or nega-tive at 70 percent power cr above ensures that the previous analysis basis for this event is not affected.

III. Transients Sensitive to a Positive Moderator Coefficient A.

Baron Dilution As stated in Section 15.2.4 of the FSAR, if a boron dilution incident occurs.during refueling or startup, the FSAR shows that the operator.has sufficient time to identify the problem and terminate the dilution before the reactor returns criti-cal. Therefore, since a return to critical is prevented the value of the moderator coefficient has no effect on a boron dilution incident during refueling or startup. The reactivity addition due to a boron dilution at power will cause an Due increase in power and reactor coolant system temperature.

to the temperature increase, a positive moderator coefficient would add additional reactivity and increase the severity of the transiint. With the reactor in automatic control, however, the rod insertion alarms provide the operator with adequate time to terminate the dilutio. before shutdown margin is lost.

n e

A boron dilution incident with the reactor in manual control is no more severe than a rod withdrawal at power, which-is ana-lyzed in Section III.C. and therefore this case was not speci-fically analyzed. Following reactor trip, the amount of time available before shutdown margin is lost is not if fected by the moderator coefficient.

B.

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

.suberitical results in an uncontrolled addition of reactivity The leading to a power excursion (Section 15.2.1 of the FSAR).

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 l

of the moderator and fuel. The reac'tivity addition due,to a positive moderator coefficient results in increases in pe'ak heat flux and peak fuel and clad temperatures.

Method of Analysis The analysis was performed in the FSAR for a reactivity inser-tion rate of 75 x 10-5 Ak/sec. This reactivity insertion rate was used in this analysis and is greater than that for the

)

I simultaneous withdrawal of the combination of the two sequen-tial control banks having the greatest combined worth at maxi-I mum speed (45 inches / minute). A constant moderator temperature I

coefficient of +5 pcm/*F was used in the analysis. The digital computer codes, initial power level, and res: tor trip instru-ment delays and setpoint errors used in the analysis were the same as used in the FSAR.

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Results and Conclusions The nuclear power, coolant temperature, heat flux, fuel average temperature, and clad temperature versus time for a 75 x 10-5 Thi s 3.k/sec insertion rate are shown in Figures 2 through 4.

insertion rate, coupled with a positive moderator temperature coefficient of +5 pcm/*F, yields values for peak heat flux, peak coolant temperature and thermal power which do not exceed nominal full power values. Therefore the conclusions presented in the FSAR are still valid.

C.

Uncontrolled Control Rod Assembly Withdrawal at Power Introduc tinn An uncontrolled control,$d assembly withdrawal at power pro-r duces a mismatch in steam flow and core power, resulting in an

. increase in reactor coolant temperat'ure. A positive moderator coefficient would augment the power mismatch and could reduce the margin to DNB. A discussion of this incident is presented in Section 15.2.2 of the FSAR and the Cycle 2 Reload Safety Evaluation (2).

Method of Analysis l

The transient was reanalyzed employing the same digital compu-ter code and assumptions regarding inst umentation and setpoint T hi s errors used for the Cycle 2 Reload N::, E nbation (2).

transient. was only analyzed at 100 perec7: power with a posi-tive moderator coef ficient since this case is the most limiting A constant moderator of those presented in previous analyses.

The assump-coefficient of +5 pcm/*F was used in the analysis.

tion that a positive moderator coefficient exists at full power is conservative since at full power, tne moderator coefficient will actually be zero or negative. For this case, the DNB evaluation was performed using the improved thermal design prc:edure (5).

e 1ACO*

Results Figure 5 shows the minimum DNBR as a function of reactivity inser:f on rate. The 'imi-ing :ase for DNB margin is a reac-tivity insertion rate of 2.0 x 10-5 ak/sec from full power initial conditions which results in a minimum DNBR of 1.86.

A positive moderator coefficient therefore does not lower the DNBR associated with a control rod assembly withdrawal at power below the limit value of 1.73.

Co nclusions These results demonstrate that the conclusions presented in the Cycle 2 Reload Safety Evaluation (2) are still valid. That is, the core and reactor coolant system are not adversely af fected since nuclear flux and overtemperature aT trips prevent the core minimum DNB ratio from falling below 1.73 for this inci-dent.

D.

Loss of Reactor Coolant Flow Introduction As demonstrated in the FSAR, Sections 15.2.5 and 15.3.4, and in f

- the Cycle 2 Reload Safety Evaluation (2) the most severe loss of flow transient is caused by the simultaneous loss of elec-trical power to all four reactor coolant pumps. This transient was reanalyzed to determine the effect of a positive moderator temperature coefficient on the nuclear power transient and the resultant effect on the minimum DNBR reached during the inci-The effect on the nuclear power transient would be dent.

limited to the initial stages of the incident during which reactor coolant temperature increases; this increase is termi-nated shortly after reactor trip.

i 1

i 1454t

Method of Analysis Analysis methods and assumptions used in the reevaluation were consistent with those employed in the Cycle 2 Rel,oad Safety Evaluation (2).

The digital computer codes used to calculate the flow coastdown and resulting system transient were the same as those used to perform the analysis described in the Cycle 2 Reload Safety Evaluation (2). The analysis was done with a constant modera-ter coef ficient of +5 pcm/*F. The DNB evaluation was performed using the improved thermal design procedure (5).

Results For the analysis performed with a +5 pcm/*F moderator coef fi-cient, the reactor coolant average temperature increases le'ss than 2*F above the initial value. Therefore, a positive-

~

moderator coefficient does not appreciably affect the reactor coolant system response or the minimum DNBR reached during the tra'nsi e nt. Figures 6 through 9 show the flow coastdown, the nuclear power and heat flux transients, and the minimum DNBR versus time.

Concl usions A positive moderator temperature coef ficient does not apprecia-bly affect the result of the complete loss of flow transient, and the minimum DNBR remains above the limit value of 1.73 for this incident. This case was analyzed since it is the most limiting one presented in the FSAR and in the Cycle 2 Reload l

Safety Evaluation (2).

Since the transient causes only a smc11 change in core average moderator temperature, and the positive moderator coefficient does not appreciably affect the nuclear power transient, the single pump ' loss of flow cases will also not be appreciably affected.

4 64

E.

Loss of External Electrical. Load Introduction Two cases, analyzed for both beginning and end of life condi-tions, are presented in Section 15.2.7 of the FSAR and the Cycle 2 Reload Safety Evaluation (2):

1.

Reactor in automatic rod control with operation of the pressurizer spray and the pressurizer power operated relief valves; and 2.

Reactor in manual rod control with no credit for pressur-izer spray or power operated relief valves.

As the moderator temperature coefficient will be ' negative at end of life, only beginning of 11.fe cases were repeated. 'The resul.t of a. loss of load is a core p'ower, level which momen-tarily exceeds the secondary system power extraction causing an increase (n core watef}emperature. The consequences of the reactivity addition due to a positive moderator coefficient are increases in both peak nuclear power and pressurizer pressure.

Method of Analysis A constant moderator temperature coefficient of +5 pcm/*F was assumed. The hethod of analysis and assumptions used were otherwise in accordance with those presented in the Cycle 2 Reload Safety Evaluation (2). The improved thermal procedure (5) was utilized in the DNB evaluation.

f Resul ts The system transient response to a total loss of load frcm 102

(

percent power, with control rods in automatic control, assuming 1494t

E pressurizer relief and spray valves, is shown in Figures 10 and

11. Peak pressurizer pressure reaches 2497 osia following a reactor trip on overtemperature aT. This compares to a value of 2467 psia presented in the Cycle 2 Reload Safety Evaluation (2). A minimum DNBR of 2.15 is reached shor ly af ter reactor trip.

Figures 12 and 13 illustrate reactor coolant system response to a loss of load with rods in manual control,, assuming no credit for pressure control. Peak pressurizer pres'sure reaches 2547 The psia following reactor trip on high pressurizer pressure.

peak pressure reached in the Cycle 2 Reload Safety Evaluation (2) analysis for this case was 2542 psia. The minimum DNBR is initially 2.26 and increases throughout the transient.

Conclusions The analysis demonstrates that the integrity. of the core and 9e reactor coolant system pressure boundary d'uring a loss of load transient will not be affected by a positive moderator reactivity coefficient since the minimum DNB ratio ' remains well above the 1.73 limit, and the peak reactor coolant pressure is less than 110 percent of design. Therefore, the conclusions presented in the Cycle 2 Reload Safety Evaluation (2) are still applicabl e.

I F.

Locked Rotor Introduction Two cases are analyzed in the FSAR '(Section 15.4.4) for this transient: four loops operating, one locked rotor; and three

~

loops operating, one locked rotor. Following a locked rotor incident, reactor coolant system temperature rises until shortly af ter reactor trip. A positive moderator coef ficient will not af fect the time to DNB since DNB is conservatively e

The tran-3ssumed to occur at the beginning of the incident.

sient was reanalyzed, however,' due to the potential effect on the nuclear power transient and thus on the peak reactor cool-ant system pressure and fuel temperatures.

Method of Analysis The digital computer codes used in the reanalysis to evaluate the pressure transient and thermal transient were the same as those used in the FSAR. The assumptions used were also consis-tent with those employed in the FSAR. An analysis was done at 102 percent power with four loops operating and 72 percent

' power for three loops operating with a moderator / coefficient of

+5 pcm/*F.

Results Table II compares results obtained for these two cases with those presented in the FSAR.. Core flow, reactor coolant pres-sure, nuclear power and heat flux versus time for both four loops operating and three l' oops operating are presented in Figures 14 through 21. Figure 22 illustrates clad temperature versus time for four loops operating, one locked rotor, which is the worst case.

Conclusions l

Analysis of the locked rotor incident with a positive moderator temperature coefficient shows that the peak reactor coolant system pressure remains below that which would cause stresses to exceed the fault.ed condition stress limits. The peak clad temperature for the hot spot during the worst transient remains much less than 2700*F and the amount of Zirconium - water reac-tion is small. Therefore, the conclusions presented in the FSAR are still valid.

~r

3.

Ruoture of a Control Rod Drive iiechanism Housing Control Rod Ejection

• ntroduction The rod ejection transient is analyzed at full power and hot standby for both beginning and end of life conditions. Si nc e the moderator temperature coefficient is negative at end of life, only the beginning of life cases were reanalyzed. The high nuclear power levels and hot spot fuel temperatures resulting from a rod ejection are increased by a positive moderator coefficient. A discussion of this transient is pre-sented in Section 15.4.6 of the FSAR.

Method of Analysis The digital computer codes for analy,ses of the nuclear power transjent and hot spot heat transfer are the,,same as those used in the FSAR. The ejected rod worths and transient peaking factors were the same as reporteLip.the FSAR@r_ ths _beginning The moderator coef ficient used for this tran-of life cases.

sient was slightly greater than +5 pcm/*F at zero power nominal average temperature, decreasing to approximately +4.5 pcm/*F at full power T-average. This is still a conseryative assumption since the moderator coefficient actually is zero or negative above 70 percent power.

Results and Conclusions Peak fuel and clad temperatures and nuclear power versus time for both full power and hot standby are presented in Figures 23 through 26. The limiting peak hot spot clad temperature, Maximum fuel 2675*F, was reached in the hot full power case.

temperatures were also assor.iated with the full power case.

Although the peak hot spot fuel centerline temperature for this transient exceeded the melting point, melting was restricted to less than the innermost 10 percent of the pellet.

As fuel and clad temperatures do not exceed the fuel and clad limits specified in the FSAR, tnere is no danger of sudden fuel dispersal into the coolant, or consequential damage to the pr4 mary coo' ant loop. The results are sumarized in Table III.

l t

l 1ACQt

I

'SECTION III j

CONCLUSIONS i

13ent analysis Of sceration of Trojan Nuclear l

To sssess the effect on 3::

Slant with a sligntly positive moderator temperature coefficient, a safety analysis of transients sensitive to a zero or positive moderator coefficient' was performe 1.

These transients included control rod assem-bly withdrawal from subcritical, control rod assembly withdrawal at power, loss of reactor coolant flow, loss of external load, and control This study indicated that a small positive moderator rod ejection.

I coefficient does not result in the violation of safety limits for the l

transients analyzed.

Except as noted, the analyses employed a const' ant modera' tor coef ficient The results of this study are of +5 pcm/*F, independent of power level.

conservative for the accidents investigated at full power, since the

' proposed Technical Specification shown in Appendix A required that the coefficient be zero or negative at or above 70 percent power.

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1459t

REFERENCES T.rojan 'Juclear 'lant, Final Safety Analysis Report, Amendment 13, 1.

dated ;tay 1974 and supplements, Docket Jo. 50-344 Reload Safety Evaluatio 1, Trojan Nuclear Plant, Cycle 2, November 2.

1977.

Reload Safety Evaluation, Trojan Nuclear P1arit, Cycle 3, May 1980.

3.

Reload Safety Evaluati'on, Trojan Nuclear Plant, Cycle 4, May 1981.

1 4.

Chelemer, H., Boman, L. H. and Sharp, D. R., " Improved Thermal 5.

Design Procedure," WCAP-8567-P, July 1975 (Proprietary) and WCAP-8568, July 1975 (Non-Proprietary).

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TABLE I ACCIDENTS EVALUATED FOR POSITIVE MODERATOR COEFFICIENT EFFECTS FSAR

' Accident Time in Life 15.2.1 RCCA Withdrawal from Suberitical BOC 15.2.2 RCCA Withdrawal from Power BOC/EOC BOC 15.2.3 RCCA Misalignment / Drop BOC 15.2.4 Baron Dilution BOC 15.2.5/3.4 Loss of Flow EOC 15.2.6 Startup of an Inactive Loop 15.2.7 Loss of Load / Turbine Trip BOC/EOC 15'.2.8 Loss of Feedwater 15.2.9 Station Blackout 15.2.10 Feedwater Malfunction EOC 15.2.11 Excessive Load Increase BOC/EOC

~

15.2.12 Accidental Depressurization of RCS BOC 15.2.13/4.2 Steam Line Break EOC BOC 15.2.14 Spurious Actuation of SI BOC 15.3.1/4.1 LOCA 15.4.2 Feed Line Break BOC 15.4.4 Locked Rotor BOC/EOC' 15.4.6 RCCA Ejection

• Accidents Evaluated BOC - Beginning of Cycle EOC - End of Cycle 1494t

TABLE II COMPARISON OF RESULTS FOR LOCKED ROTOR ANALYSES Four Loops Operating

+5 pcm/*F FSAR Moderator temperature coefficient, ak/*F 5 x 10-5 0

Initial power level, percent of nominal 102 102 Peak reactor coolant system pressure, psia 2665 2632 Peak clad temperature during transient, *F 2032 2017 Amount of Ir-H O at core hot spot, percent

.59

.59 2

hy weight Three Loops Operating

+5 pcm/*F FSAR Moderator temperature coefficient, ak/*F 5 x 10-5 0

72 72 Initial power level, percent of nominal Peak reactor coolant system pressure, psia 2695 2630 5

m 1aEQt

TABLE III

SUMMARY

OF ROD EJECTION RESULTS BEGINNING OF CYCLE I

Hot Zero Power Hot Full Power l

Maximum fuel pellet average temperature, *F 3224 4225 Maximum fuel center temperature, *F 3743 5153 Maximum clad average temperature, 'F 2438 2575 i

Maximum fuel enthalpy, cal /gm 135 186 0

< 10 Fuel pellet melting, percent i

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APPENDIX A i

TECHNICAL SPECIFICATION CHANGE o

Replace paragraph 3.1.1.4a with:

10.5 x 10-4 ak/k/*F below 70 percent RATED THERMAL POWER a.

10.0 x 10-4 ak/k/*F at or above 70 percent RATED THERMAL l

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10 20 30 40 50 60 70 80 90 100 POWER 1

l FIGURE 1 MODERATOR TEMPERATURE COEFFICIENT VS

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FIGURE 2 R00 WITHDRAWAL FROM SUBCRITICAL NUCLEAR POWER V5 TIME l

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FIGURE 3 ROD WITHDRAWAL FROM SUSCRITICAL TEMPERATURE VS TIME k

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FIGURE 11 LOSS OF_ LOAD' AUTOMATIC R00 CONTROL '41TH PRESSURIZER RELIEF 'UD SPCAY y-

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500.00 5.0000 4.0000 -

3.0000 -

E 3

2.0000 -

1.0000 -

e y

a e

e 0.0

-c o

e o

o o

e o

o o

C o

e

~

o TIME (SEC)

FIGURE 12 LOSS OF LOAD MANUAL R0D CONTROL. --......

\\

1400 0 1350.0 -

1300.0 -

m

' E E

1250.0 - -

E W

1200.0 -

.E M i

"_ b.

1150.0 -

2

~

5i 1100.0 -

n b

E 1050.0 -

1000.00 -

950.00 i

i

..=.; w.,.

610.00 mx tE 600.00 -

m m

i 590.00 - -

m.-

m c.

580.00 - -

x o-570.00 -

m e

s._

560.00 -

5;:

550.00 f

e e

e e

e

- o e

o o

O o

O C

C C

o o

e oe e

~

m TIME (SEC)

IGURE n LCSS OF LOAD mat!UAL 3.00 CC'!30L.

1 ::r..vp..... n.....-

e - :..r g

,-u

li I

1.2000 l-1.0000 4 E

.80000 -

m5 e=

66 C

.30000 -

m85 y-b E

.40000 -

~

.20000 -

e e

e o

6 0.0 e

e o

o eo e

o e

o O

G O

O C

e w

=

a

,m TIME (SEC)

FIGURE 14 LOCKED ROTOR 1/4 PRESSURE CORE FLOW VS TIME o

e n

r.-

---

=_---,r%

p 2700.0 ;-

2600.0 -

4 2500.0 -

=$

u E.

2400.0 -

e5

<m

~

JL 8~

2300.0 -

L,)

x C:

>=

M 2200.0 -

d 2100.0 -

2000.0 o

e o

e e

e e

o e

e e

o e

e e

e e

e e

e o

e o

e Q

O e

c d

3 2

2 8

c 4

e

[

TIME (SEC) 1 1

FIGURE 15 LOCKED ROTOR

~

l/4 PRES 5URE I

REACTOR COOLANT PRESSURE VS TIME i

I i

G e

-. _. _. ~ -.

\\

1

)

i l

1.2000 1 0000 -

.3

.80000 -

a=

wa 2=

c c.

,c

.60000 -

et =

WC d-

=u z<

=

,40000 - -

s

.20000 -

0.0 o

e e

o e

e o

e e

e e

o e

e o

e e

e e

e D

e D

e O

D e

D eu D

~

e e

o eu D

~

~

TIME (SEC)

FIGURE 16 LOCKED ROTOR 1/4 PRESSURE NUCLEAR POWER VS TIME

i

-s

==

1.2000 i

1.0000 ->

3<z G

z

.80000 -

wo zo P

.60000 -

u ac w

~

g

.40000 -

d Y

20000 -

i

~

1 0.0 o

o e

o o

o e-e o

o e

e e

o o

.o o

o o

e e

m

=

o cu TIME (SEC) l FIGURE 17 LOCKED ROTOR 1/4 PRESSURE HEAT FLUX VS TIME

1.2000 1.0000 -

~ ::

.80000 -

. <5

~E ez du

.60000 -

C E

g 6.

E

.40000 -

e

.20000 -

o e

o o

o 0.0 O

O O

O O

e o

e o

8 e

e O

e o

e m

.o eu TIME (SEC)

~

FIGURE 18 LOCKED ROTOR 1/3 PRESSURE CORE FLOW VS TIME

e 2700.0 2600.0 --

2500.0 -

w=

~m M

wg 2400.0 -

5$

. <. ~.

JM 8 S.

2300.0 -

us

'

• e ~. w n c

• W 2200.0 -

W GT.

2100.0 e

0 t

e e

.g g

g 2000.0 4

4 i

e e

e g

g a

g

.~

e a

e e

e g

a N

~

~

~

I f

M c

TIME (SEC) -

l 1

~

FIGURE 19 LOCKED ROTOR 1/3 PRESSURE REACTOR COOLANT PRESSURE VS TIME

,n 7

I l

I i

f 1.2000 1 0000 -

m E

80000 -

55 gz c_ w

.60000 -

a:

55 J-b

~

~

z$

40000 -

20000 e

e o

o e

o e

o 00 e

o e

o e

o o

e r

c o

o o

c-o e

e o

e o

e g

e N

\\

~

m e

~

.o TIME (SEC) i l

FIGURE 20 LOCKED ROTOR 1/3 PRESSURE NUCLEAR POWER VS TIME

~ " " ' " ~ ~ ~

mn_,.w,,

1 2000 1 0000 --

<z 1

e

~

z 80000 -

uo v

g p

.60000 -

V

.c E

' x 40000 -

3

>=<

E

.20000 e

e o

e e

0.0 C

C C

C C

o C

o

=

=

C C

C C

C C

cc J

a N

C TIME (SEC)

FIGURE 21 LOCKED ROTOR 1/3 PRESSURE HEAT FLUX VS TIME

~

2100.0 2000.0 --

l 4

1750.0 --

C o

1500.0 -

w o

W

=

52 1250.0 -

-e

=

1 W 1000. 00 -

o<av 750.00 -

600.00 l o

o e

.o o

.o o

e o

e o

a C

C O

o o

o e

o e

o N

a up CD TIME (SEC)

FIGURE ~22 LOCKED ROTOR 1/4 CLAD CLAD TEMPERATURE V5 TIME

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

d 3.0000 I

2.5000 -

E 2.0000 -

e. -

5E Cz

c. u E$

1.5000 -

b.a v-EG 1.0000 -

i

~~-

4,..... _

.50000 -

e e

C C

6 C

e C

0.0 C

C C

C C

C C

C C

D C

C C

C' C

D.

C N

W N

C O

c C

c

~

o

=

n

~.

TIME (SEC)

FIGURE 23 ROD EJECTION BOL HFP NUCLEAR POWER VS TIME e

-m n - -

-w

-w-w,-

.r-,

6000.0 I

MELTING

---

~

l 5000.0 -

50A0' F FUEL CENTER TEMPERATURE 8:'00. 0 -

FUEL AVERAGE C

TEMPERATURE 3000.0 -

E

?

2000.0 -

CLAD OUTER c.6 TEMPERATURE 1000.00

0. 0 6

6 e

o o

e o

e o

o e

o o

e o

e o

e o

o a

J J

S o

TIME (SEC)

FIGURE 24 R00 EJECTION BOL HFP TEMPERATURE VS TIME v

-w

,.-n e

m e r --,

-e

-w

l n

1:

11 10 l H

I 5

L f

-1 g

-a 10 l

f

=

z-mm

f..

2m c

mm C

ec

=

w E. 10-o v-I z<*

- l CU CC m

w

-5 10 u.

10~7 f

5 g

g 8

e e

g 8

8 8

e

=

n

=

a e

~

e M

g 4

o

~

T1HE (SEC)

FIGURE 25 RCD EJECTION BOL HZP NUCLEAR POWER V5 TIME

.(,.

=

6000.0 5000 0 -

l l

4000.0 - -

FUEL CENTEP. TEMPERATURE, C

N

=

~

~

3000.0. -

FUEL

~

m At*CRAGE TEMPERATURE m

5 l

CLA0 4tTER, -

2 2000.O w="-

TEMPERATURE 1000.00 -

.J 0.0

c e

e o

e C

C O

C o

o e

o e

o e

o e

e o

c.,,.

g,0 gn C

tJ g

l TIME (SEC)

FIGURE 26 ROD EJECTION BOL NZP TEMPERATURE VS TIME 1

_