ML20069D934

Proposed Tech Spec Change 87,revising Westinghouse 14x14 Optimized Fuel Assembly Design

ML20069D934
Person / Time
Site:	Point Beach
Issue date:	03/14/1983
From:	WISCONSIN ELECTRIC POWER CO.
To:
Shared Package
ML20069D931	List: ML20069D927 ML20069D934
References
TAC-49969, TAC-49970, NUDOCS 8303210116
Download: ML20069D934 (38)

Text

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-o.

s-TABLE 1 POINT BEACH NUCLEAR PLANT

, FUEL ASSEMBLY DESIGN COMPARISON Standard 0FA

'1. Rod Array 14 x 14 14 x 14

2. Rod Pitch.(Inches) 0.556 same

3. Overall dimensions (Inches) 7.763x7.763* same

4. Fuel Weight (U in U02 ) kg's 402.7 356.8

5. Active Fuel length (Inches) 144 same

6. Fuel Rod 0.D. (Inches) .422 .400

7. Clad Thickness (Inches) .0243 same

8. Clad Material Zircaloy-4 same

9. Fuel Pellet 0.D. (Inches) .3659 .3444

10. Number of Grids:

Inconel 7 2 Zircaloy-4 --

5

11. Nominal Assembly Weight (1bs.) 1274 1135

12. End Fittings Material Top Nozzle - Hold down Springs Inconel-718 same

- Fitting and Tie Plate SS-304 same Bottom Nozzle - SS-304 SS-304 (non-reconstitutable) (reconstitutable)

13. Guide and Instrument Tube Materials Zircaloy-4 same i

(14 x 0.556") x (14 x 0.556") = 7.784" x 7.784". Used for spent fuel storage criticality calculations.

0303210116 830314 PDR ADOCK 05000266 P PDR s 1

.f' Attachment A POINT BEACH - UNITS 1 AND 2 TECHNICAL SPECIFICATION CHANGE REQUEST NO. 87

SUMMARY

OF CHANGES Description of Page Specification Change . Comments 15.1-4 3) Refueling Shutdown, change This change reflects a change in refueling "10 percent" to "5 percent" shutdown margin and'is shown acceptable for the Uncontrolled Boron Dilution Accident at Refueling conditions, independent of tran-sition to optimized fuel. .

15.2.1-1 15.2.1 Specification 1., line 3, change As a result of conversion to 0FA, two

. " Figure 15.2.1-1" to " Figure 15.2.1-1 for safety limit figures are presented, one a transition core

• and Figure 15.2.1.-2 for for transition cores and one for full 0FA the full optimized fuel assembly (OFA) core." cores.

Add to the bottom of page: "* Transition core is defined as being any core loading pattern consisting of standard and optimized 14 x14 fuel assemblies, not including 0FA demonstration assemblies.

15.2.1-2 and The basis has been rewritten to address both The basis has been rewritten to reflect 15.2.1-3 transition and full 0FA cores. Discussion of a more current Technical Specifications DNB correlations and the .3 FAH multiplier is (tech spec) description on this subject.

l included. Figures 15.2.1-1 and Figures 15.2.1-2 As a result the tech spec has been sin-l were revised to reflect the safety limits with plified and abbreviated.

l optimized fuel applicable to both Units 1 and 2.

1

Descriptien of ,

Prge Specificatian Chrnge Comments 15.2.3-2 Change the following to read: The overtemperature AT and overpower AT '

(a) "for q t

~9 b within -17, + 5 percent" setpoint equations were determined for: ,

(b) "... equivalent of 2.0 percent" (1) both units; (2) transition and 0FA cores; (3) 2250 psia and 2000 psia operation; and (4) pressure ranges within the range of the high and low pressure trips-respectively.

All analyses confirm acceptability of OTAT, OPAT and low pressurizer trip of 1790 psig changes for above described conditions.

15.2.3-3 (c) change "...two percent" to "2.0 percent" Consistent with (b) 15.2.3-6 Basis for 15.2.3, line 2 should read "... Figure Reference to two safety limit figures.

15.2.1-1 and 15.2.1-2". Second paragraph, line 2, remove "have been revised", and line 3 remove

" revised". Paragraph 2, last line should read

" Figure 15.2.1-1 and 15.2.1-2.

15.2.3-7 Third Paragraph - Add explanation of the This is inconsistent with the original design difference between the nominal and allowed methodology used. However, setpoints are values. correct.

9

Description of ,.

Pige Specificttien Change Comments 15.3.1-17 thru Inclusion of positive moderator temperature This change:is independent of 0FA. Safety *~ '

15.3.1-18 coefficient (MTC) specification and basis to analyses (to be documented) have been -

reflect the change. analyzed or evaluated to show acceptability of this change.

Removal of specification for subcriticality Reactivity effect is very small and l due to depressurization effects. would be overshadowed by temperature effects.

15.3.1-19 G. 3. the "178,000" value is changed to The minimum measured flow is consistent-j reflect a measurement plus uncertainty value with the assumptions in analyzing Condi-J of "181,800 gps". Basis changed, tion I and II events with the Improved s

Thermal Design Procedure. Basis discusses measurement uncertainties and verification of thermal design flow.

15.3.6-1 Spec 15.3.6.A.c) - withdrawal of one bank pf. This change reflects the definition change 1 control rod is allowed for rod drop testing; in refueling shutdown margin, and clari-last line, change "10 percent" to "5 percent" fies rod drop testing requirements.

15.3.6-2 Basis for 15.3.6 - Second paragraph, lines 4 , Change in refueling shutdown margin, and and 7, change "10 percent" to "5 percent", margin for rod drop tests.

Second paragraph last line, chnge "FFDSAR" i to "FSAR". s 15.3.8-4 First line, change "10 percent" to "5 percent". Change in refueling shutdown margin.

i t

! )

y Descriptien of ,

P_Ege Specificaticn Chrnoe Comments 15.3.10-2 The LOCAq F limit was revised to reflect the new Resulted from change from standard to

• value of 2.21. (4.42 for P < .5)_

optimized fuel.

The Fg limit was changed to reflect the values Resulted from change in Fg multiplier assumed in generating the core limits (Figures from .2 to .3.

15.2.1-1 and 15.2.1-2).

15.3.10-3 thru Specification 15.3.10.B.2 was completely re- The tech spec wording is similar to that 15.3.10-4 written to incorporate the Relaxed Axial Offset employed in the RAOC generic report re-Control strategy (RAOC). C l' references to tar- cently submitted to the NRC and informally get axial offset and the +6 - 9 percent target approved for use on the Sequoyah reload.

bands were deleted, since these concepts are Some changes to the specifications have not appropriate for RAOC. This specification been made to reflect Point Beach operat-defines: (a) the AI operating envelope, (b) ing procedures. This specification change the actions required when the AI envelope is was not necessary to accommodate optimized violated, (c) acceptable conditions for power fuel.

escalation above 50 percent rated thermal power, (d) the surveillance requirements on axial flux difference (AI), (e) the criterion for a AI viola-tion.

15.3.10-7 The maximum rod drop time was relaxed to 2.2 This accommodates the change to optimized secs. fuel.

15.3.10-9 ' -

1.30 DNBR value was changed to "the limit Wording change.

W R".

15.3.10-11 The LOCAqF limit was changed to 2.21. The Resulted from change from standard to reference to rod bow was deleted. optimized fuel.

Descripticn of -

PC2e Specificntirn Change Comments 15.3.10-12 thru The basis for Axial Power Distribution was com- The tech spec wording is consistent with-15.3.10-13 pletely rewritten to incorporate RAOC. The AI that employed in the RAOC generic report envelope constructed assures that the LOCA F q submitted to the NRC. I,t should be noted limit is not violated during normal operation. that the RAOC analysis methodology de-scribed in the report is consistent with the methodology utilized for Point Beach.

Figure 15.3.10-1 The control rod insertion limits were changed Variations in fuel management, such as slightly, as a contingency measure. abnormally short or long cycles could necessitate future revisions to the AI operating envelope. Raising the insertion limits slightly will make a revision less likely. This change is not necessary for optimized fuel.

Figure 15.3.10-3 The LOCA K(Z) curve was revised. Resulted from optimized fuel LOCA analysis.

Figure 15.3.10-4 Flux difference operating envelope. Required for RAOC.

Table 15.4.1-1 1. Nuclear Power Range under " check" monthly Clarification of incore/excore calibration j add "(4)" and under " Remarks" add "(4) Com- to within 3 percent. Credit was taken pare incore... 3 percent." for this requirement in OP and OTAT Setpoint.

15.5.3-1 thru Modified Reactor Core specifications (Descrip- Required for core loading with optimized 15.5.3-2 tion). fuel l 15.5.4-1 Modified U-235 axial loading and storage loca- Required for storage of new and spent tion restrictions. OFA fuel.

ATTAC10ENT B POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 TECHNICAL SPECIFICATION CHANGE REQbT.ST NO. 87 REVISta.' PAGES l

f f

I l 3 l 4 I l

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2) Cold Shutdown The reactor is in the cold shutdown condition when the reactor has a shutdown margin of at' least 1 percent Ak/k and reactor coolant temperature is <200*F

3) Refueling Shutdown The reactor is in the refueling shutdown condition when the reactor is suberitical by at least 5 percent Ak/k and T,y is il40*F. A refueling shutdown refers to a shutdown to move fuel to and from the reactor core.

4) Shutdown. Margin Shutdown margin is the instantaneous amount of reactivity by which the reactor core would be suberitical if all withdrawn control rods were tripped into the core but the highest worth withdrawn RCCA remains fully withdrawn. If the reactor is shut down from a power 4 condition, the hot shutdown temperature should be assumed. In other cases, no change in temperature should be assumed.

h. Power Operation The reactor is in power operating condition when the reactor is critical and the average neutron flux of the power range instrumentation indicates greater than 2 percent of rated power.

i. Refueling Operation Refueling operation is any operation involving movement of core components (those that could affect the reactivity of the core) within the containment when the vessel head is unbolted or removed.

J. Rated Power Rated power is here defined as a steady-state reactor core output of 1518.5 NWT.

k. Thermal Power Thermal power is defined as the total core heat transferred from the fuel to the coolant.

15.1-4

15.2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 15.2.1 SAFETY LIMIT, REACTOR CORE Applicability: ,

Applies to the limiting combinations of thermal power, reactor coolant system pressure, and coolant temperature during operation.

Objective:

To maintain the integrity of the fuel cladding.

Specification:

1. The combination of thermal power level, coolant pressure, and coolant temperature shall not exceed the limits shown in Figure 15.2.1-1 for a transition core

• and 15.2.1-2 for the full a

optimi^ed fuel assembly (OFA) core. The safety limit is exceeded if the point defined by the combination of reactor coolant system average temperature and power level is at any time above the appropriate pressure line.

• Transition core is defined as being any core loading pattern consisting of standard and 0FA 14x14 fuel assemblies, but not including cores consisting of standard assemblies and 0FA demonstration assemblies.

i l

1 15.2.1-1

Basis:

The restrictions of this safety limit prevent overheating of the fuel and j possible cladding perforation which would result in the release of fission l products to the reactor coolant. Overheating of the fuel cladding is pre-vented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temper-ature is slightly above the coolant saturation temperature.

Operation above the upper boundary of the nucleate boiling regime could result in excess cladding temperature because of the onset of departure from nucleate boiling (DN8) and the resultant charp reduction in heat transfer coefficient.

DN8 is not a directly measurable parameter during operation and therefore thermal power and Reactor Coolant Temperature and Pressure have been related to DNB.

This relation has been developed to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux distributions. The local DNB heat flux ratio, DNBR, defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the margin 7 to DNB.

The DNB design basis is as follows: there must be at least a 95 percent proba-4 bility at a 95 percent confidence level that DNB will not occur during steady state operation, normal operational transients, and anticipated transients and is chosen as an appropriate margin to DNB for all operating conditions.

• The curves of Figure 15.2.1-1 are applicable to transition cores. The W-3 correlation is used to generate these curves. The DNBR limit of this correlation is shown to be met in plant safety analyses using values of input parameters with uncertainties considered at fixed conservative values.

l The curves of Figure 15.2.1-2 are applicable for a full core of 14 x 14 0FA.

The WRB-1 correlation is used to generate these curves. Uncertainties in plant parameters are statistically convoluted to determine the plant DNBR uncertainty. 1 This DNBR uncertainty, combined with the correlation DNBR limit, establishes a value of design limit DNBR. This value of design limit DNBR is shown to be met in plant safety analyses, using values of input parameters considered at their nominal values.

15.2.1-2

These curves represent the loci of points of thermal power, Reactor Coolant System pressure and average temperature for which the calculated DNBR is no less than the design Ifmit-DN8R or the average enthalpy at the vessel exit is less than the enthalpy of saturated liquid. Appropriate rod bow penalties have been included in the generation of these curves. The effects of fuel densification and possible clad flattening have also been taken into account.

An allowance is included in these curves for an increase in F g at reduced power based on the expression:

F H= Full Power F g [1 + 0.3 (1 - P)]

where P is a fraction of rated thermal power.

The hot channel factors are sufficiently large to account for the degree of mal-positioning of full-length rods that is allowed before the reactor trip set-points are reduced and rod withdrawal block and load runback may be required.

Rod withdrawal block and load runback occur before reactor trip setpoints are reached. The Reactor Control and Protective System is designed to prevent any anticipated combination of transient conditions that would result in a DNB ratio of less than the design limit.ONBR.

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. (3)* Low pressurizer pressure - )1855 psig for operation at 2250 psia primary system pressure

>1790 psig for operation at 2000 psia primary system pressure (4) Overtemperature AT 1

STo (K 1 - K 2 (T-T') ( 1*T 3 ) + K3 (P-P') - f(AI))

1+t28 where

! ATo = indicated AT at rated power, "F

! T = average temperature, F'  ;

T' = 574.2 'F

P = pressurizer pressure, psig  !

P' = 2235 psig  !

• K y 11.117 for operation at 2250 psia primary system pressure

$1.30 for operation at 2000 psia primary system pressure K = 0.0150 2

K = 0.000791 3

Il = 25 sec

, t2 = 3 sec 2

and f(AI) is an even function of the indicated difference between top and bottom detectors of the power-range nuclear ion chambers; with gains to be

selected based on measured instrument response during plant startup tests, where qt and qb are the percent power in the top and bottom halves of the core respectively, and qt*9b is total core power in percent of rated power, such that

(a) ~9 b within -17, +5 percent, f(AI) = 0.

for q t

(b) for each percer.t that the magnitude of qt ~9 b exceeds +5 percent, the AT trip set point shall be automatically reduced by an equivalent of 2.0 percent of rated power.

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e Appropriate safety analyses shall be performed prior to shifting operation from one primary system pressure to the other.

j 15.2.3-2

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-(c) for each percent that the magnitude of gt " 9 bexceeds -17 percent, the AT trip setpoint shall be automatically reduced by an equivalent of 2.0 percent of rated power.

[1.B. (5)] Overpower AT 1AT, [K4 -K 5 T35. T-K 6(T-T') - f (AI)]

TS+1 3

where AT, = indicated AT at rated power, *F T = average temperature, *F T' = 574.2*F K4 1 1.089 of rated power K5 = 0.0262 for increasing T

= 0.0 for decreasing T K6 = 0.00123 for T 1 T'

= 0.0 for T < T' T

3 = 10 sec.

f (AI) as defined in (4) above, (6) Undervoltage - >75 percent of normal voltage (7) Lower indicated reactor coolant flow per loop - >90 percent of normal indicated loop flow (8) Reactor coolant pump motor breaker open (a) Lew frequency set point 157.5 cps (b) Low voltage set point >75 percent of normal voltage.

~

1 15.2.3-3

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e power distribution, the reactor trip limit, with allowance for errors (2) ,y, always below the core safety limit as shown on Figures 15.2.1-1 and 15.2.1-2.

If axial peaks are greater than design, as indicated by the difference between the top and bottom power range nuclear detectors, the reactor trip limit is automatically reduced (6)(7) , j The overpower, overtemperature and pressurizer pressure system setpoints include the effect of reduced system pressure operation (including the effects of fuel densification). The setpoints will not exceed the core safety limits ,

as shown in Figures 15.2.1-1 and 15.2.1-2.

The overpower limit criteria is that core power be prevented from reaching a value at which fuel pellet centerline melt,ing would occur. The reactor is pre. vented from reaching the overpower limit condition by action of the nuclear overpower and overpower AT trips.

The high and low pressure reactor trips limit the pressure range in which reactor operation is permitted. The high pressurizer pressure reactor trip setting is lower than the set pressure for the safety valves (2485 psig) such that the reactor is tripped before the safety valves actuate. The low pressurizer pressure reactor trip trips the reactor in the unlikely event of a loss-of-coolant accident (4) .

The low flow reactor trip protects the core against DNB in the event of either a decreasing actual measured flow in the loops or a sudden loss of power to one or both reactor coolant pumps. The setpoint specified is consistent with the value used in the accident analysis (0) The low loop flow signal is caused by a condition of less than 90 percent flow as measured by the loop flow instrumentation. The loss of power signal is caused by 15.2.3-6 l

_ 0

e the reactor coolant pump breaker opening as actuated by either high current, low supply voltage or low electrical frequency, or by a manual control switch.

The significant feature of the breaker trip is the frequency setpoint, 57.5 cps, which assures a trip signal before the pump inertia is reduced to an unacceptable value. The high pressurizer water level reactor trip protects the pressurizer safety valves against water relief. The specified setpoint allows adequate oper-ating instrument errorI2) and transient overshoot in level before the reactor trips.

The low-low steam generator water level reactor trip protects against loss of feedwater flow accidents. The specifed setpoint assures that there will 'be sufficient water inventory in the steam generators at the time of trip to allow for starting delays for the auxiliary feedwater system.(9)

Numerous reactor trips are blocked at low power where they are not required for protection and would otherwise interfere with normal plant operations. The pre-scribed setpoint above which these trips are unblocked assures their availability in the power range where needed. Specifications 15.2.3.2.A(1) and 15.2.3.2.C have 11% tolerance to allow for a 2% deadband of the P10 bistable which is used to set the limit of both items. The difference between the nominal and maximum allowed value (or minimum allowed value) is to account for "as measured" rack drift effects.

Sustained operation with only one pump will not be permitted above 10 percent power. If a pump is lost while operating between 10 percent and 50 percent pc;.er, an orderly and immediate reduction in power level to below 10 percent is allowed.

The power-to-flow ratio will be maintained equal to or less than unity, which ensures that the minimum DNB ratio increases at lower flow, because the maximum enthalpy rise does not increase above the maximum enthalpy rise wnich occurs during full power and full flow operation.

1 References (1) FSAR 14.1.1 (4) FSAR 14.3.1 (7) FSAR 3.2.1 (2) FSAR, Page 14-3 (5) FSAR 14.1.2 (8) FSAR 14.1.9 (3) FSAR 14.2.6 (6) FSAR 7.2, 7.3 (9) FSAR 14.1.11 15.2.3-7

F. MINIMUM CONDITIONS FOR CRITICALITY Specification:

1. Except during low power physics tests, the reactor shall not be made critical when the moderator temperature coefficient is more positive than:

a. 0.5 x 10 ~4 Ak/k/'F (applicable below 70 percent of rated power), or

b. 0.0 Ak/k/*F (applicable at or above 70 percent or rated power).

2. In no case shall the reactor be made critical (other than for the purpose of low level physics tests) to the left of the reactor core criticality curve presented in Figures 15.3.1-1 for Unit 1 and 15.3.1-3 for Unit 2.

l

3. O Thereactorshallbemaintainedsubcriticalbyatleast1percentfuntil normal water level is established in the pressurizer.

Basis:

During the early part of the fuel cycle, the moderator temperature coefficient is calculated to be slightly positive at coolant temperatures below 70 percent of rated thermal power.(1) (2) The moderator coefficient at low temperatures will be most positive at the beginning of life of the fuel cycle, when the boron concentration in the coolant is the greatest. Later in the life of the fuel cycle, the boron concentrations in the coolant will be lower and the moderator coefficients will be either less positive or will be negative. At all times, the moderator coefficient is negative when > 70 percent of rated thermal power. Suitable physics measurements of moderator coefficient of

, reactivity will be made as part of the startup program to verify analytic predictions.

l l

I 15.3.1-17 l

g  %

The limitations of the moderator temperature coefficient are provided to ensure that the auumptions used in the accident and transient analyses remain valid through each fuel cycle. This requirement is waived during low power physics tests to permit measurement of reactor moderator coefficient and other physics design parameters of interest. During physics tests, special operating precautions will be taken. In addition, the strong negative Doppler coefficient (3) and the small integrated ak/k would limit the magnitude of a power excursion resulting from a reduction of moderator density.

The requirement that the reactor is not to be made critical below the Reactor Core Criticality Curve provides assurance that a proper relationship between l reactor coolant pressure and temperature will be maintained during system heatup and pressurization. Heatup to this temperature will be accomplished by operating the reactor coolant pumps. However, as provided in 10 CFR Part 50, Appendix G Section IV.A.2.c., the reactor core may be taken critical below this curve for the purpose of low level physics tests.

If the specified shutdown margin is maintained (Section 15.3.10), there is no possibility of an accidental criticality as a result of an increase of moderator temperature or a decrease of coolant pressure (I) .

The requirement for bubble formation in the pressurizer when the reactor has passed the threshhold of 1 percent subcriticality will assure that the reactor coolant system will not be solid when criticality is achieved.

References:

(1) FSAR Table 3.2.1-1 (2) FSAR Figure 3.2.1-9 (3) FSAR Figure 3.2.1-10 I

15.3.1-18

4 G. OPERATIONAL LINITATIONS The following DNB related parameters shall be maintained within the limits shown during Rated Power operation:

1. T shall be maintained at or below 578"F.

Avg

• 2. Reactor Coolant System (RCS) pressure shall be maintained:

1 2205 psig during operation at 2250 psia or, 1 1955 psig during operation at 2000 psia.

3. Reactor Coolant System raw measured Total Flow Rate 1 181,800 gpm (See Basis).

Basis:

The reactor coolant system total flow rate of 181,800 gpm is based on an assumed measurement uncertainty of 2.1 percent over thermal design flow (178,000 gpm).

The raw measured flow is based upon the use of normalized elbow tap differential pressure which is calibrated against a precision flow calorimeter at the begin-ning of each cycle.

Appropriate safety analyses shall be performed prior to shifting operation from one primary system pressure to the other.

15.3.1-19

l e.

15.3.6 CONTAINMENT SYSTEM Applicability:

Applies to the integrity of reactor containment.

Objective:

To define the operating status of the reactor containment for plant operation.

Specification:

A. Containment Integrity a) The containment integrity (as defined in 15.1) shall not be violated when a nuclear core is installed in the reactor unless the reactor is in the cold shutdown condition, b) The containment integrity shall not be violated when the reactor vessel head is removed unless the reactor is in the refueling shutdown condi-tion.

c) Positive reactivity changes shall not be made by rod drive motion when the containment integrity is not intact except for the testing of one bank of rods at a time, rod discon~necting, and rod reconnecting pro-vided the reactor is initially subcritical by at least 5 percent ak/k.

d) Positive reactivity changes shall not be made by boron dilution when the containment integrity is not intact unless the boron concentration in the reactor is maintained > 1800 ppm.

B. Internal Pressure If the internal pressure exceeds 3 psig or the internal vacuum exceeds 2.0 psig, the condition shall be corrected or the reactor rendered sub-critical.

1 I

i 15.3.6-1  !

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C. Containment Purge Supply and Exhaust Valves The containment purge supply and exnaust valves shall be locked closed and may not be opened unless the reactor is in the cold shutdown or refueling shutdown condition.

Basis The Reactor Coolant System conditions of cold shutdown assure that no steam will be formed and hence there would be no pressure buildup in the containment if the reactor coolant system ruptures.

The shutdown conditions of the reactor are selected based on the type of activ-ities that are being carried out. When the reactor head is not to be removed, the specified cold shutdown margin of 1% ak/k precludes criticality under any occurrence. During refueling, the reactor is subcritical by at least 5% Ak/k.

Positive reactivity changes for the purpose of rod assembly testing will not result in criticality because no control bank worth exceeds 3%. Positive reac-tivity changes by boron dilution may be required or small concentration fluctua-tions may occur during preparation for, recovery from, or during refueling but maintaining the boron concentration greater than 1800 ppm precludes criticality under these circumstances. 1800 ppm is a nominal value that ensures 5% shutdown for typical reload cores. Should continuous dilution occur, the time intervals -

for this incident are discussed in Section 14.1.5 of the FSAR.

Regarding internal pressure limitations, the containment design pressure of 60 psig would not be exceeded if the internal pressure before a major loss-of-coolant accident were as much as 6 psig(1) The containment is designed to withstand an internal vacuum of 2.0 psig( ).

The containment purge supply and exhaust valves are required to be locked closed during plant operations since these valves have not been demonstrated capable of closing from the full open position during a design basis loss-of-coolant accident. Maintaining these valves locked closed during plant operation ensures that excessive quantities of radioactive materials will not be released via the 15.3.6-2

e containment purge system in the event of a design basis loss-of-coolant accident.

The containment purge supply and exhaust valves will be locked closed by providing locking devices on the control board operators for these valves.

1 References (1) FSAR - Section 14.3.4 (2) FSAR - Section 5.5.2 i

b 4

9 6

1 4

15.3.6-3

i subcritical approximately by 5 percent ak/k in the cold condition with all rods inserted.(2) Periodic checks of refueling water boron concentration insure that proper shutdown margin is maintained. Part A6 allows the control room operator to inform the manipulator operator of any impending unsafe condition detected from the main control board indicators during fuel movement.

During the refueling operation a substantial number of station personnel and perhaps some regulatory people will be in the containment. The requirements are to prevent an unsafe amount of radioactivity from escaping to the environment in the case of a refueling accident, and also to allow safe avenues of escape for the personnel inside the containment as required by the Wisconsin Department of Industry, Labor 'and Human Relations. To provide for these requirements, the personnel locks (both doors) are open for the normal refueling operations with a third temporary door which opens outward installed across the outside end of the personnel lock.I ) This hollow metal third door is equipped with weather strip-ing and an automatic door closer to minimize the exchange of inside air with the outside atmosphere under the very small differential pressures expected while in the refueling condition. Upon sounding of the containment evacuation alarm, all personnel will exit through the temporary door (s) and then all personnel lock doors shall be closed. As soon as possible, the fuel transfer gate valve shall be closed to back up the 30 foot water seal to, prevent escape of fission products.

The spent fuel storage pool at the Point Beach Nuclear Plant consists of a single pool with a four foot thick reinforced concrete divider wall which separ-ates the pool into a north half and south half. The divider wall is notched to a point sixteen feet above the pool floor to allow transfer of assemblies from one half of the pool to the other. -

l I

15.3.8-4

..(

B. Power Distribution Limits

1. a. Except during low power physics tests, the hot channel factors defined in the basis must meet the following limits:

Fq (Z) 1 (2.21) x K(Z) for P> 0.5 ,

P Fq(Z)

_ 1 4.42 x K(Z) for P1 0.5

F < l.58 x [1 + 0.3 (1-P)]

Where P is the fraction of full power at which the core is operating, K(Z) is the function in Figure 15.3.10-3, and Z is the core height location of F .

q

b. Following a refueling shutdown prior to exceeding 90 percent of

! rated power and at effective full power monthly intervals thereafter, power distribution maps using the moveable incore

detector system shall be made to confirm that the hot channel factor limits are satisfied. The measured hot channel factors 4

shall be increased in the following way:

(1) Themeasurementoftotalpeakingfactor,feas , shall be increased by three percent to account for manufacturing tolerances and further increased by five percent to account for measurement error.

(2) The measurement of enthalpy rise hot channel factor, F ,

shall be increased by four percent to account for measurement error.

(c) If a measured hot channel factor exceeds the full power limit of Specification 15.3.10.B.I.a, the reactor power and power range high setpoints shall be reduced until those limits are met. If subsequent flux mapping cannot, within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, demonstrate that the full power hot channel factor limits are met, the overpower 15.3.10-2

I e and overtemperature AT trip setpoints shall be similarly reduced and reactor power limited such that Specification 15.3.10.B.1.a above is met.

2. a. The indicated axial flux difference (AFD) shall be maintained within the allowed operational space defined by Figure 15.3.10-4 except during physics tests. The physics test exemption applies provided that the thermal powar is less than or equal to 85% of rated thermal power and the limits of specification 15.3.10.B.1.a are satisfied.

During suspension of the specification, the thermal power shall be determined to be less than or equal to 85% of rated thermal power at least once per hour. In addition, the curveillance requirements of 15.3.10.8.1.b shall be performed at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

b. If the indicated AFD deviates from the Figure 15.3.10-4 limits, the AFD shall be restored to within the Figure 15.3.10-4 limits within 15 minutes. If this cannot be accomplished, then reactor power shall be reduced until the AFD is within the envelope or the power level is less than 50 percent of rated thermal power. Normally the rate of power reduction is 15% per hour. Once AFD has been returned to and maintained within the operating envelope, power level is no longer restricted. If it is necessary to reduce power to 50%, the Power Range Neutro.n Flux-High Trip Setpoints shall be reduced to less than or equal to 55 percent 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 />.

c. A power increase to a level greater than 50 percent of rated power is contingent upon the indicated AFD being within the Figure 15.3.10-4 limits.

d. Alarms shall normally be used to indicate non-conformance with the flux difference requirements of 15.3.10.B.2.a and 15.3.10.B.2.b. If the alarms are totally out of service, the AFD shall be noted and l

conformance with the limits assessed every hour for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />,  ;

and half-hourly thereafter.  :

I l

e. The indicated AFD shall be considered outside of its limits when at least 2 operable excore channels are indicating the AFD to be outside the limits.

15.3.10-3

O 4

D. Misalianed or Dropped RCCA

1. If the rod position indicator channel is functional and the associated RCCA is more than 7.5 inches indicated out of alignment with its bank and cannot be aligned when the bank is between 215 steps and 30 steps, then unless the hot channel factors are shown to be within design limits as specified in Section 15.3.10.B-1 within eight (8) hours, power shall be reduced to less than 75 percent of rated power. When the bank position is greater than or equal to 215 steps, or, less than or equal to 30 steps, the allowable indicated misalignment is 15 inches.

2. To increase power above 75 percent with an RCCA more than 7.5 inches indicated out of alignment with its bank when the bank position is between 215 steps and 30 steps, an analysis shall first be made to determine the hot channel factors and the resulting allowable power level based on Section 15.3.10.B. When the bank position is greater than or equal to 215 steps, Ror, less than or equal to 30 steps, the allowable indicated misalignment is 15 inches.

3. If it is determined that the apparent misalignment or dropped RCCA indication was caused by rod position indicator channel failure, sustained power operation may be continued if the following conditions are met:

a. For cperation between 10 percent power and rated power, the position of the RCCA(s) with the failed rod position indicator channel (s) will be checked indirectly by core instrumentation (excore detectors, and/or thermocouples, and/or moveable incore detectors) every shift and after associated bank motion exceeding 24 steps in one direction.

b. For operation below 10 percent of rated power, no special monitoring is required.

E. RCCA Drop Times

1. At operating temperature and full flow, the drop time of each RCCA shall be no greater than 2.2 seconds from the loss of stationary gripper coil voltage to dashpot entry. l 15.3.10-7 1

4 of axial power distribution. One may assume no change in core poisoning due to xenon, samarium or soluble boron.

Power Distribution l I

Design criteria have been chosen which are consistent with the fuel integrity '

~

analyses. These relate to fission gas release, pellet temperature and cladding mechanical properties. Also the minimum DNBR in the core must not be less than the limit DNBR in normal operation or in short-term transients.

In addition to the above, the peak linear power density must not exceed the limiting kw/ft values which result from the large break loss-of-coolant accident analysis based upon the ECCS acceptance criteria limit of 2200'F. This is required to meet the initial conditions assumed for loss-of-coolant accident.

To aid in specifying the limits on power distribution, the following hot channel factors are defined:

Fq (Z), Height Dependent Heat Flux Hot Channel Factor, is defined as the local heat flux on the surface of a fuel rod at core elevation Z divided by the average fuel rod heat flux, allowing for manufacturing tolerances on fuel pellets and rods. Imposed limits pertain to the maximum F (Z) 9 in the core.

F , Engineering Heat Flux Hot Channel Factor, is defined as the allowance on heat flux required for manufacturing tolerances. The engineering factor allows for local variatons in enrichment, pellet density end diameter, surface area of the fuel rod and eccentricity of the gap between pellet and clad. Combined statistically, the net effect is a factor of 1.03 to be applied to fuel rod surface heat flux.

15.3.10-9

s An upper bound Envelope of 2.21 times the normalized peaking factor axial dependence of Figure 15.3.10-3 con-istent with the Technical Specifications on power distribution control as given in Section 15.3.10 was used in the LOCA analysis. The results of the analyses based on this upper bound envelope indicate a peak clad temperature of less than 2200*F limit. When an F measurement q

is taken, both experimental error and manufacturing tolerance must be allowed for. Five percent is the appropriate allowance for a full core map taken with the uoveable incore detector flux mapping system and three percent is the appropriate allowance for manufacturing tolerance. In the design limit of Fh,thereiseightpercentallowanceforuncertaintieswhichmeansthat normaloperationofthecoreisexpectedtoresultinadesignFh 5,1.58/1.08.

The logic behind the larger uncertainty in this case is that (a) normal perturba-i tions in the radial power shape (i.e. , rod misaligryment) affect F H, in m st cases without necessarily affecting F q

, (b) while the operator has a direct influence on Fq through movement of rods, and can limit it to the desired value,hehasnodirectcontroloverFhand(c)anerrorinthepredictions for radial power shape which may be detected during startup physics tests can be compensated for inqF by tighter axial contro ; but compensation for F is H

less readily available. When a measurement of Fg is taken, experimental

{ error must be allowed for and four percent is the appropriate allowance for a j

full core map taken with the moveable incore detector flux mapping system.

Measurements of the hot channel factors are required as part of startup physics tests, at least each full power month operation, and whenever abnormal power distribution conditions require a reduction of core power to a level based upon measured hot channel factors. The incore map taken following initial 1 loading provides confirmation of the basic nuclear design bases including proper fuel loading patterns. The periodic monthly incore mapping provides additional assurance that the nuclear design bases remain inviolate and identify operational 15.3.10-11

O anomalies which would, otherwise, affect these bases.

Axial Power Distribution The limits on axial flux difference (AFD) assure that the axial power distribu-imit tion is maintained such that the F (Z) q upper bound envelope of F times the normalized axial peaking factor [K(Z)] is not exceeded during either normal opera-tion or in the event of xenon redistribution following power changes.

Provisions for monitoring the AFD on an automatic basis are derived from the plant process computer through the AFD monitor alarm. The computer determines the one minute average of each of the operable excore detector outputs and pro-vides an alarm message immediately if the AFD for at least 2 of 4 or 2 of 3 oper-able excore channels are outside the AFD limits and the thermal power is greater than 50 percent of Rated Thermal Power.

+

15.3.10-12

r ,,

i e

J J

i

, Quadrant Tilt /

4 The excore detectors are somewhat insensitive to disturbances near the core center such as misaligned inner control rods It is therefore possible that a five percent tilt might actually be oresent in the core when the excore detectors respond with a two percent indicated quadrant tilt. On the other hand, they are overly responsive to distrubances near the periphery.

4 15.3,10-13

0

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TABLE 15.4.1-1 :(1 of 4)

MINIMUM FREQUENCIES FOR CHECKS, CALIBRATIONS AND TEST OF INSTRUMENT CHANNELS-Channel" Description Check Calibrate Test Remarks

1. Nuclear Power Range S(1)** D(1)** B/W(2)** (1) Heat Balance M*(3)**(4) Q*(3)**(4) (2) Signal to AT; bistable action (permissive, rod stops, trips)

(3) Upper and lower chambers for axial off-set' (4) Compare incore to excore axial flux difference. Recalibrate if the absolute difference is greater than or equal'to 3 percent.

2. Nuclear Intermediate Range S(1)** N.A. P(2) (1) Once/ shift when in service (2) Log level; bistable action (permissive, rod stop, trips)

3. Nuclear Source Range S(l) N.A. P(2) (1) Once/ shift when in service (2) Bistable action (alarm, trips)

4. Reactor Coolant Temperature S R B/W(1)** (1) Overtemperature -AT (2) (2) Overpower - AT

5. Reactor Coolant Flow S** R M**

6. Pressurizer Water Level S** R M**

7. Pressurizer Pressure S** R N**

8. 4 Kv Voltage N.A. R M** Reactor protection circuits only

9. Analog Rod Position S(l)** R M** (1) With step counters

• By means of the moveable incore detector system.

• • Not required during periods of refueling shutdown, but must be performed prior to starting up if it has not been performed during the previous surveillance period. Tests of permissive and low power trip bistable setpoints which cannot be done during power operations shall be conducted prior to startup if not done in the previous two weeks.

j 1

o 4.

'15.5.3 REACTOR

! Applicability Applies to the reactor core, Reactor Coolant System, and Emergency Core Cooling Systems.

Objective

~

To define those design features which are essential in providing for safe system operation.

Specifications f'

A. Reactor Core

1. Standard reactor cores consisting entirely of standard design fuel, contain approximately 48 metric tons of slightly enriched uranium.

The initial-transition reactor core contains approximately 47 metric tons of uranium in the form of slightly enriched uranium dioxide pellets. The transition core is, defined as being any core loading pattern consisting of standard and 0FA 14x14 fuel assemblies. Use of 0FA demonstration assemblies in cores of standard design fuel does not constitute a transition core. The pellets are encapsulated in Zircaloy-4 tubing to form fuel rods. The reactor core is made up of 121 fuel assemblies. Each fuel assembly nominally contains 179 fuel rods (I) Where safety limits are not violated, individual fuel rods suspected of leaking may be replaced with an inert rod or the assembly left with a water hole to prevent possible reinsertion of leaking fuel rods. No more than one fuel rod may be replaced in any single assembly and no more than six (6) such modified assemblies r

may reside in the core at any time.

2. The average reload region enrichment of the initial transition core is a nominal 3.20 weight percent of U-235.

3. The transition reload fuel will be similar in design to the initial core of standard fuel.

15.5.3-1 l

4-  ;

4. Burnable poison rods are incorporated for reactivity and/or power aistribution control. The burnable poison rods consist of borated pyrex glass clad with stainless steel (4) .

5. There are 33 full-length RCC assemblies in the reactor core. The full-length RCC assemblies contain a 142-inch length of silver-indium-cadmium alloy clad with the stainless steel.

6. Up to ten (10) grams of enriched fissionable material may be used either in the core, or available on the plant site, in the form of fabricated neutron flux detectors for the purposes of monitoring

-core neutron flux.

B. Reactor Coolant System

1. The design of the Reactor Coolant System complies with the code requirements (0) .

2. All high pressure piping, components of the Reactor Coolant System and their supporting structures are designed to Class I requirements, and have been designed to withstand: ,

a. The design seismic ground acceleration, 0.06g, acting in the horizoatal and 0.04g acting in the vertical planes simultaneously, with stresses maintained within code allowable working stresses.

b. The maximum potential seismic ground acceleration, 0.12g, act-ing in the horizontal and 0.08g acting in the vertical planes -

simultaneously with no loss of function.,

3. The nominal liquid volume of the Reactor Coolant Systen, at rated operating conditions, is 6040 cubic feet.

References (1) FSAR Section 3.2.3 (4) FSAR Section 3.2.3 (2) FSAR Section 3.2.1 (5) FSAR Sections 3.2.1 & 3.2.3 (3) FSAF Section 3.2.3 (6) FSAR Table 4.1-9 15.5.3-2

0 15.5.4 FUEL STORAGE Applicability Applies to the capacity and storage arrays of new and spent fuel.

Objective To define those aspects of fuel storage relating to prevention of criticality in fuel storage areas.

Specification

1. The new fuel storage and spent fuel pool structures are designed to withstand the anticipated earthquake loadings as Class I structures. The spent fuel pool has a stainicss steel liner to ensure against loss of water.

2. The new and spent fuel storage racks are designed so that it.is impossible to store assemblies in other than the prescribed storage locations. The fuel is stored vertically in an array with sufficient center-to-center '

distance between assemblies to assure K,77 < 0.95 with the storage pool filled with unborated water and with the fuel loading in the assemblies limited to 44.8 grams of U-235 per axial centimeter of standard fuel assemblies and 39.4 grams of U-235 per axial centimeter of 0FA fuel

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assemblies. An inspection area shall allow rotation of fuel assemblies for visual inspection, but shall not be used for storage.

3. The spent fuel storage pool shall be filled with borated water at a concentration of at least 1800 ppm boron whenever taere are spent fuel assemblies in the storage pool.

4. Each storage location immediately adjacent to a wall shall be restricted to storage of fuel assemblies having a cooling time of one year or more.

References FSAR Section 9.3 15.5.4-1

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