ML19309B035

Safety Evaluation Supporting Order Modifying OL to Conform W/Eccs Evaluation Results.Acrs Encl

ML19309B035
Person / Time
Site:	Rancho Seco
Issue date:	12/27/1974
From:	US ATOMIC ENERGY COMMISSION (AEC)
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ML17054D216	List: ML17054D214 ML19309B029 ML19309B032 ML19309B035
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NUDOCS 8004020510
Download: ML19309B035 (28)

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SAFETY EVALUATION REPORT BY THE DIRECTORATE OF LICENSING fff

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U.S. ATOMIC ENERGY COMMISSION-E:ii IN THE MATTER OF

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SACRAMENTO MUNICIPAL UTILITY DISTRICT RANCHO SECO NUCLEAR GD!ERATING STATION UNIT 1 DOCl;ET NO. 50-312 "Y

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TABLE OF CONTENTS t

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g EE 1.0 I NT RO DU CT IO N...........................................

1

- 2.0 BABCOCK AND WILCOX ECCS EVALU ATION 110 DEL................ 5

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30 APPLICABILITY OF CENERIC EVALUATION MODEL..............

8 km 4.0 RESULTS OF LOCA CALCULATIONS...........................

9 5 0 CONCLUSIONS............................................

17 6.0 R E F E R EN CE S.............................................

20 E

e APPENDIX A - OPERATING RESTRICTIONS us APPENDIX B - LETTER FROM ADVISORY C0!O!ITTEE ON REACTOR

..E SAFEGUARDS, NOVD4BER 20, 1974

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LIST OF TABLES l

Pare TABLE 1.

A COMPA9ISON OF RAtlCHO SECO UNIT 1 TO KEY PARAMETERS EMPLOYED IN THE GENERIC EVALUATION MODEL........

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TABLE 2.

SU! MARY OF' SENSITIVITY STUDIES....................

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LIST Of' FIGURES

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Rod Position Limits.........................

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FIGURE A-2.. LOCA Limited Maximum Allowable Linear Heat

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1.0 INTRODUCTION

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On January 4, 1974, the Commissior published its acceptance criteria for emergency core cooling systems for light water power reactors (39 FR 1003). This rule included Appendix % to 10 CFR 50 which specifies analytical techniquec to be employed for the evaluation of ECCS effectiveness. On August 5, 1974,

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Babcock and Wilcox officially submitted a five volume package E

(1,2,3,4,5) of topical reports constituting their proposed ECCS evaluation model. The information contained in these reports had been the subject of a number of informal conferences and discussions between the staff and Babcock and Wilcox, starting just prior to the publication of the Acceptance Criteria in January, 1974.

The Regulatory staff reviewed these documents and published (6) a Status Report on October 15, 1974, which addressed each item

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required by Appendix K and identified areas which were acceptable to the staff and areas of concern which were to be resolved.

On November 13, 1974, the Regulatory staff published a Supplement (7) to the Status Report which addressed each of these areas of concern.

As reflected in the Supplement, for some items adequate additional information was provided to enable'the staff to accept.

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&E the, Babcock and Wilcox approach.- Fo.r certain other items, the staff

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concluded that adequate justification had not been provided and

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that further modification of the August 5, 1974 model was required.

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Babenk and Wilcox will modify their model to reflect these e5ff

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reonf rements and haa evaluated the effect of all changes upon (9) 1 the previous calculations. Accordingly, the Babcock and 'Vilcox model* with the modifications presented in Section 2.0 and 4.0 of this SER is acceptable and would conform to Appendix K.

A report of the Advisory Committee on Reactor Safeguards,

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attached as Appendix B, was issued on November 20, 1974 regarding W

l the generic review and the acceptability of the Babcock and Wilcox ECCS Evaluation Model.

On August 2,1974, Sacramento Municipal Utility Dist'rict (the licensee) submitted an analysis or ECCS performance for the Rancho Seco Unit 1, (the plant or facility) along with proposed Technical Specification (8) changes to reflect the new ECCS evaluation model calculations.

This evaluation was based upon the Babcock and Wilcox August 5,1974 Evaluation Model. Section 3 0 of this SER discusses the applicability of the generic evaluation model to the specific Bincho Seco Unit 1 plant.

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As stated in the Status Report and its Supplement, the August 5th Babcock and Wilcox Evaluation Model was not completely acceptable

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Supplement were required. These changes are now being made to the generic Babcock and Wilcox evaluation model.

Since the Rancho Seco. Unit 1 cvaluation was based upon a model which was not acceptable, it also requires some changes.

A revised set

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of computations for the plant (and for other facilities in a like position), using the newly revised and acceptable evaluation

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model, cannot be submitted for a number of months.

' To determine the effect of the model changes made to the August 5,1974 Eabcock and Wilcox Evaluation Mo' del, the staff requested, and Babcock and Wilcox submitted, a series of generic

= "i plant sensitivity studies which quantified the effect of the (9) model changes on the relults of the previous calculations.

s The staff followed the performance of these sensitivity studies While they were in progress and has re. viewed the results. These results are presented in Section 4.0 along with'a discussion of

s their effects on the evaluation submitted for Rancho Seco (8)

Unit 1.

From these studies, it appears that certain operating restrictions are req'. ired to ensure that in the event of a

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E postulated loss-of-coolant accident, ECCS cooling performance will not exceed the values for calculated peak clad temperature and oxidation and hydrogen sencration limits set forth in 10 CpR 50.46(b). These restrictions on maximum heat generation rate are set forth in the proposed Technical Specifications submitted on August 2,1974, and are set forth in Appendix A hereto along with the other appropriate operating limits.

To verify the limitations con-

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tained in the licensee's submittal of August 2,1974, a reevaluation of ECCS performance in conformity with 10 CFR 50.46 and Appendi:: K, 3..q:;

and based upon an approved evaluation model should be submitted for Rancho Seco Nuclear Generating Station Unit 1, along with appropriate Technical Specifications based on sue.h evaluation, as soon as practi-cable.

During the interim, before an evaluation wholly in conformity

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with the requirements of 10 CFR 50.46 can be submitted and evaluated, continued conformance to'the requirements of the Commission's A

Interim Acceptance Criteria (IAC) and the restrictions contained (8) in' the licensee's August 2,1974, submittal, combined with the additional limitations set forth in Appendix A hereto, will provide reasonable assurance that the public health an'd safety will not be ei:

endangered.

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e 2.0 BABCOCK AND WILCOX ECCS EVALUATION MODEL ggg;[

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The staff Status Report provides a complete evaluation of Eis ~

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the Babcock and Wilcox ECCS Evaluation Model.

Each part of 10.CFR

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50, Appendix K was addressed and appropriate comments regarding compliance to each aspect of the model were included.

All phases of the Babcock and Wilcox analytical method' were concluded to s

.:ja meet Appendix K requirements with the exceptions noted in Supplement

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(7) 1 of the Status Report.

Of the fourteen areas of concern

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addressed in Supplecent 1 to the Status Report, five were identified as model deficiencies for Oconee Class reactors (177 fuel assembly plants with a lowered loop arrangement) requiring modification or additional dat.a to justify conformance to Appendix K.

These areas are briefly discussed below.

Additional detail of each deficiency is presented in Section 4.0 of this SER and in the staff

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(6,7)

Status Report.

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'A complete listing of each computer program, in the same form as used in.the evaluation model, was furnished to the Regulatory

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staff.

These listings, combined with the Babcock and Wilcox impact

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studies, constitute the currently acceptable ECCS model.

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2.1 Metal k'ater Reaction

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The staff required that the Babcock and Wilcox ECCS model

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and outside of the fuel cladding.

In addition, an improved

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calculational technique for arriving at a predicted value

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for total core-wide metal-water reaction resulted from staff comments. Babcock and Wilcox is modifying its ECCS model to incorporate these features.

See Section 4.0 for an assessment of impact upon the current plant operating restrictions.'

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55 2.2 Swelling and Rupture of the Cladding and Fuel Rod Thermal Parameters

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(7)

.As noted in the Status Report and Supplement 1, the staff accepted the Babcock and Wilcox modeling of swelling and rupture with three limitations.

As discussed in Section 4.0 of this SER, these limitations were satisfied in the Rancho Seco Unit 1 9

evaluation.

Babcock and Wilco:: has proposed to modify its model to incorporate a plastic swelling model, discussed in the 5 atus Report Supplement, and a transient pin pressure codel, which would eliminate two of the staff limitations. These modifications have not yet been completed.

At present, the e.xisting swelling and rupture model is acceptable if the staff limitations are observed.

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23 End-of-Blowdown As indicated in the Status Report and Supplement 1, the staff lh, accepted the modeling of end-of-blowdown with the conditions

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that the definition of end-of-bypass be char.ged and that the down-comer noding representation be changed to ase a homogeneous noding.'

gg Babcock and Wilcox is modifying its model to incorporate these

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Section 4.0 discusses the impact of these deficiencies on the licensee's calculations.

2.4 Containment Pressure (6)

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Page 4-41 of the Status Report states that the containnent backpressure calculation performed for the Oconee Class plants

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is conservative and acceptable. For plants of a different type, specific input assumptions must be justified on an individual plant bas.ls.

Although the backpressure model is acccptable for the Oconee 4

Class plants, the effect of the use of the conservatively assumed parameters should be assessed by comparison with actual as-built values. Accordingly, the licensee has been requested to provide as-built values and to discuss the methods useo

',o determine the passive containment heat sinks for Rancho Seco Unit 1.

Also required is an identification of each sink by category (e.g., cable tray

. equipment supports, floor grating, crane wall) and surface area,.

thickness, materials'or construction, thermal conductivity and l

volumetric heat capacity by component catege.ry'.

Values of paint thickness, thermal conductivity and volumetric heat capacity for ennt.<n..nt intorn.1.tenntuvo. nr..1 n v nn t a.

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=Ed5 2.5 Steam Interaction with Emergency Core Cooling Water in Pressurized

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h5 Water Reactorr Two concerns discussed by the staff in Supplement 1 to the

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Status Report.

are related to.the effect of hot walls on the E

ECC water being injected in the downcomer and the appropriateness of the value used for vent valve resistance. Babcock and Wilcox will modify their model to incorporate the resolution of these

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concerns. Section 4.0 assesses the impact of these concerns upon

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1:nu the plant operating restrictions.

,.j.]l:.h3 30 APPLICABILITY OF GENERIC EVALUATION MODEL

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As noted in BAW-10091 and in the staff's Status (6)

Report, the' development of the generic Babcock and Wilcox

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Evaluati5n Model involved the utilization of a plant desigli appropriate to all Oconee Class reactors. The series of sensiti-

=r vity studies described in BAW-10091, Section 5 0 were therefore

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directly applic'able to Rancho Seco Unit 1.

Also worthy of note are the actual key parameters utilized in the generic model calculations. Babcock and Wilcox stated that they ~ b5unded the variations in key. parameters within the Oco.iee Class plants by choosing values-in their generic calculations which conservatively include any plant-to-plant variations. Table 1 provides a list k_

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of such key parameters employed in the generic. evaluation and-E.7=

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Unit 1.

This list shows that the generic calculation sufficiently incorporated the differences in these key parameters found in this plant.

4.0 RESULTS OF LOCA CALCULATIONS From a break spectrum analysis, the worst break cramined by 2

Babcock and Wilcox using the August 5,1974 model was an 8.55 ft

.(1) double-ended rupture at the reactor coolant pump discharge. This (8) generic analysis was the basis for the licensee's submittal.

o This calculation resulted in a peak clad temperature of 2062 F, 3 38% local =etal-water reaction, and 0.14% whole core metal-water reaction.

These values are.within the criteria of 10 CFR 50.46 o

(2200 F, 17%, and 15, respectively).

e All of the model deficiencies noted in Section 2.0 of this SER were examined by Babcock and Wilcox with regard to an (9) impact assessment on current operating reactors.

The following sections address each of the relevant model deficiencies and their effects on the August 5, 1974 LOCA analysis.

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4.1 Metal-Water' Reaction E.E=;

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E As indicated in Section 2.1 of this SER, the staff his i

requested that the Babcock and Wilcox ECCS Evaluation Model be

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revised to account for thinning of the oxide layer on the inside and outside of the fuel cladding. The generic model LOCA limit ei (1)

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calculations assumed initial values of 0.0001 inches oxide

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layer thickness and 1800 psia internal pin pressure.

An' oxide

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(7) thickness sensitivity study conducted by Babcock and Wilcox

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yielded the conclusion that the value of internal pin pressure combined with the value of the oxide thickness used by Babcock and Wilcox in their generic calculations conservatively predicted the highest peak cladding temperature for fuel cycle 1 operations.

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The Babcock and Wilcox study thinned the oxide layers consistent with the degree of pin swelling predicted.

.g The staff also noted in the Status Report that further

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a calculational technique for predicting total core-wide metal-water reaction.

In the Supplement, the staff reported that

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Babcock and Wilcox had chosen to modify their codel in a manner 7g E

which the staff found would be adequate.

These codifications

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To deterc'ne

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i whether this todification would affect the calculations submitted ws H

by the 1,icensee, the staff considered sensitivity studics performed using staff codels previously developed for' confircation of analyses'

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submitted under the IAC.

Although these models do not fully incorporate all required evaluation features, they are adequate y=

to deconstrate that the results will fall well within the hydrogen generation criteria of 10 CFR 50.46(b)(3).

Therefore, this modification has no impact en the licensec's calculations.

4.2 Swelline and Ruoture of the Claddine and Fuel Rod Thermal Para :eters a

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(6)

As noted in the Status Report, the, staff accepted the generic calculation if three limitations were observed:

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The internal pin pressure selected for the initial condition value must exceed the naximum predicted during normal

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b)

If the rod with the highest peak clad temperature ruptures,

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then the time of rapture is restricted to a time period prior to the end of blowdown.

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e c) 70% circumferential swelling for certain rupture tempera-

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It is permissible to increase

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TE=E linearly from 1700 F (about 40% circumferential swelling) 3..

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to the 70% plateau at 2000 F.

.,.m The Rancho Seco Unit 1 analyses satisfy each of these limitations (maximum pin pressure was assumed, ruptures occurred g

prior to end of blowdown, and rupture temperatures were less than 1700'F). Accordingly, there is no impact on the licensee's calculation.

j2f 43 End-of-B1cudown (1)

Since the generic calculation showed that end-of-bypass always occurred prior to, or at the same time as, end-of-blowdown, the model change regarding the definition of end-of-bypass has gg.

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no effect on peak clad temperatures for this plant.

With regard to the staff concern that the downcomer model did not appear to be properly represented, Babcock and Wilcox has now changed the downcomer noding to a homogeneous noding representation as required in the Status Report Supplement.

In connection with this change, a number of other areas previously modeledonaheterogerAousbasishavealsobeenchangedto homogeneous noding.

This is acceptable.

These modifications will require related changes to the generic model sensitivity

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studies. These are being performed by Babcock and Wilcox.

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However, in assessing the impact of this required change upon-the

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calculations made using the August 5th model, Babcock and Wilcox

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5 found that two counteracting phenomena occur to result in, an overall decrease in peak clad temperature at the 6-foot elevation of about 2=E 23::

n 80 F.

Although less water remains in the versel at the end-of-

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bypass (leading to a longer adiabatic heatup), reduced water s+

head in the downcomer allows a significantly higher negative flow

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through the core for a longer period of time. As previo,usly

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indicated, the overall effect is to decrease the peak clad

.jf temperature, especially at the higher core elevations.

4.4 Contair. ment Pressure For the reasons stated in Section 2.4 of this SER, staff concerns in the area of the containment backpressure calculation

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have no effect on the licensee's calculations.

4.5 L eam, Interaction with Emerrency Core Cooline Water in Pressurized Water Reactors (7)

As noted in Supplement 1 to the Status Report, the staff i

required that Babcock and Wilcox correct the vent valve resistance

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(K) for two-phase. flow by applying a factor of 1.5 to the single phase value. With respect to the vent valve flow resistance hp.i 4

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.;y factor used by Babcock and Wilcox (K = 3 9), the staff required lie. q r

correction of th'is factor for two-phase flow.

As indicated in'

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the Supplement, a correcti6n factor of C = 1.5 based upon

. appropriate experimental' data for gate valves was proper along with a further correction to account for the pressure dependence of C. In the Reynolds number range of interest

• during reflood (starting with a rcrerence K of 3 3 based on single-phase data),

=eE a multiplier of 0.85 is acceptable to correct for pressure effcets..

Therefore, the required vent valve K-factor to be used in reflood sE calculation is:

K = 3 3 x 1 5 x 0.85 = 4.2 s;

Babcock and Wilcox will modify its model to use this value.

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sensitivity studies were performed by Babcock and Wilcox to assess the impact of this change of assumed vent valve K.

The results of these studies showed that an increase in vent valve (1) resistance from the value of 3 9 used in the generic calculations o

to 4.2 showed about a 20 F increase in peak clad temperature.

With regard to the effect of hot walls on the ECC water being injected in the downcomer, the staff has provided Babcock and Wil-(6) cox a description of an acceptable hot wall time delay model.

During the hot wall delay period, ECC water, which is delayed in

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passing through the downcomer, accumulates In'available storage

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volumes in the following manner:

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1) Lower downcomer - region between the bottom of the i=~

downcomer and the lower lip of the celd leg.

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m-maximum of 1/3 of this' volume will become available essw;

.e:-wem bE" linearly over the hot wall delay period.

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2) Upper downcomer - region of downcocer abov,e the lower lip of cold leg pipe.

If the lower downcomer volume

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spill'out tne break.

A storage volume is available in

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55 head above the bottom of the cold leg. The sace elevation head should be used to determine the break flow rate.

3) Cold leg piping from the reactor coolant pucp discharge to

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the vessel nozzle. A storage volume consistent with.the upper down-comer water level is available.

Once the hot wall delay time has elasped and flow through t'he downcomer begins, a further period of tice is required for the ECC water to flow from the available storage volumes to the lower plenum. To reflect this period, a downcocer transport (free fall) delay time is calculated which is added to the hot wall delay time to yield the total time required for ECC water to fall from the inlet elevation to the bottom of the downcoter (lower plenum).

Once the hot wall delay time is ended and free fall starts, no further spillage of ECC water out the break would occur.- Babcock

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and Wilcox has indicated that sufficient storage capacity exists Sh to account for the volume of water which could be accumulated.

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during the hot wall delay time. 'Therefore, there is no. net change

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Z-4.6 Summarv of Results A review of preceding Sections 4.1 through 4.5 shows that f')f s m:

the two model deficiencies which have an impact on the previous generic calculations are region noding (Section 4 3) and vent

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valve K-factor (Section 4.5).

Table 2 shows a summary o,f the Sj{

results of sensitivity studies by Babcock and Wilcox on peak c' lad

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temperature, local metal-water reaction. and whole core metal-water.

reaction.

These calculations indicate that, while the model corrections could'cause an increase in peak clad temperature, this increase would not be large enough to exceed the criteria of 10 CFR 50.46, provided that the LOCA limit curve in Appendix A are observed

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+5 in facility operation.

The LOCA limit curve submitted on August 2, 1974, establishes proper limits for ECCS ecoling performance.

However, the accompanying

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rod position limit curve could have adversely affected limitations controlling rod ejection accident consequences.

In accordance with stsff instructions, the licensee submitted a proper rod position limit curve on December 6, 1974. This curve, included as Figure A-1 in Appendix A, establishes appropriate limits for the rod ejection acci-dent considerations and slightly more restrictive limits on maximum linear heat generation rate for ECCS performance considerations.

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=+.#v 50 CONCLUS_LOJE O

Based on the analysis set forth in this. Safety Evaluation, the limitations contained in.th'e licensee's submittals, particularly 7]j E."

the LOCA limit curve set forth in Appendix A, along with the' addi-

?:Mi ditional restrictions for rod ejection accident considerations, will

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assure conformance with the peak clad temperature limit, and maximum l] jf oxidation and hydrogen generation criteria of 10 CFR 50.46(b).

However, these restrictions should be verified by a re-analysis based on the Babcock and Wilcox Evaluation Model, modified as

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described in this Safety Evaluation Report.

In addition, Rancho Seco Unit 1 satisfies the two remaining (6)

criteria, i.e., maintenance of a coolable geometry and long-term (10) cooling.

The heat removal system for long-term cooling of the plant as d'escribed in the FSAR is satisfactory for these requirements.

An evaluation of ECCS performance wholly in conformance with

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10 CFR 50.46 and Appendix K, based on an approved evaluation model#

should be subtaitted for this facility as soon as practicable, but within six months or before any refueling is authorized. During the interim, until each evaluation is submitted and evaluated by the staff operation should conform to the requirements of the Interim

. Acceptance Criteria

, as well as.to the requirements (8) of the licensce's submittals and the requirements of Appendix A.

e The Eabcock & Wilcox ECCS Evaluation Model, which i's in conformance with Appendix K of 10 CFR 50.46, is described in a letter from Babcock and Wilcox dated December 18, 1974.(9)

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A COMPARISON OF RANCHO SECO UNIT 1 TO KEY' PARAMETERS EMPLOYED IN THE E 5:E.

GENERIC EVALUATION MODEL E!=

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PARAMETER GENERIC MODEL

. UNIT 1

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- Rated Core Power,'Mwt 2,772 2,772 Reactor Vessel Flow, (1)

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. Ibm /sec 38,306 39,193

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Reactor Coolant System

.l-j Pressure at Core Outlet, psig 2,182 2,185

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1 Core Inlet' Fluid (2) o Temperature, F 556 557

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Volume Average Fuel ni.4!:=

Temperature at 18 Kw/ft N."'

with a Sink Temperature 4

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of 580 F, F

3,105 2,980 ECCS Delay Time, seconds 35 25 Reactor Building Free.

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Volume ft 2.205x10 1 98x10

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Flows are total systems flows because core flow is' not measured.

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These are estimates since full power has not yet been achieved.

Other vessel flows and fluid temperatures are measured values.

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EE:eEE ll5V

' TABLE 2.

SUMMARY

OF SENSITIVITY STUDIES EE

_.==:

(8.55 ft Double-Ended Rupture)

T'.E-Axial

1. Peak Clad

1. Local M-W

• Whole-Core M-W Kw/ft Position, ft Temocrature, F

_ Reaction, 5 Reaction, 5

'22 16.0 2

2167 3 77

<0 5

. 5.

E: E4i?

'S 4

2112 3 01

<0 5-

,,[h 18.0 6

2122 3 53

<0 5

=;;;;;.

~

17*1 8

2059 2'.21

- - ~ ~

<0 5

"""~

16.0 10 1877 1.68

<0 5

=.:...

• :::b.h:~~ ~

sj 5 CRITERIA

~.....

o

._ I Peak clad temperature............ 2200 F H

Local Metal-Water Reaction........

17%

Yhole-Core Metal-Water Reaction...

1%

...)

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d

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f e

1 Ei5I t'

)

H 1

E

.=.=

=-.

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u.t

6.0 REFERENCES

lie +

1. BAW-10091, "B&W's ECC5 Evaluation Model. Report with Specific

-,{I[l Tj Application to 177 FA Class Plants with Lowered Loop

[

E+i Arrangement," August 1974

~

2. BAW-10092, " CRAFT 2-Fortran Program for Digital Simulation of a Multinode Reactor Plant During Loss of Coolant," July 1974.

3.. BAW-10093, "REFLOOD - Description of Model for Mult.inode Core -

Reflood Analysis," July 1974

4. BAW-10094, " Revisions to THETA 1-B, A Computer Code for Nuclear

=i

~

Reactor Core Thermal Analysis," IN-1445, July 1974.

[

5. BAW-10095, " CONTEMPT - Computer Program for Predicting Containment Pressure-Temperature Response to a Loss-of-Coolant Accident," July 1974.

6. " Status Report by the Directorate of Licensing in the Matter of Babcock and Wilcox ECCS Evaluation Model Conforma *nce to 10 CFR 50, Appendix K," October 1974.

7. " Supplement 1 to the Status Report by the Directorate of Licensing in the Matter of Ba' :ock and Wilcox ECCS Evaluation Model Conformance to 10 CFR 50, Appendix K,"

November 13, 1974.

~

8. Letter from J. J. Mattimoe to Mr. A. Giambusso dated

~

August 2,1974.

10.. Letter from James F. Mallay to T.M. Novak dated

~

4=

December 18, 1974.

10. Letter from James F. Malley to T.M. Novak dated November. ~25, 1974.

...me==

=

~=

+

==

'::::::= :; :p

• 1

- EE=;E-APPENDIX'A 55

=

OPERATING RESTRICTIONS-

= &&

The Regulatory staff has reviewed the methods used by Babceck'

..;jg and Wilcox - to derive the LOCA-related operating limits for'its plants.

The review considered.the basic calculation method,' the range.of

~jjg l

!.i.5-operating conditions calculated, the types of uncertainties and their ez.E:;

@h:.~ ~

magnitude, and the instrumentation provided to monitor plant operation.

Based on this review, we conclude that sufficient monitoring instru-mentation is~present to provide assurance that the plant may be operated within LOCA related operating restrictions. We further conclude that

ii

~

operation of Rancho Seco Unit 1, within the restrictions shown on

.. = =

Figure A-1, will assure that the heat generation limits of Figure

~ ' " ~

A-2 will not be exceeded. Operation of the facility shall conform with the values' set forth in these Figures.

G

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100 195.7,

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94.5 e.e 177.7,94.51

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175.9,87.1 j,

=._l:jiih so m

4 i

166.8,76.0 70 RESTRICTED REGION 60

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. s 50 N,'

122 3,46.3 u.

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M

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PERMISSIBLE O

1 OPERATING REGION 20

- - 80,15 10 49,0 0

1 I

1 I

I I

I O

20 40 60 80 100 120 140 160 180 200 R0D INDEX,YWITHDRAWN O

25 50 75 100 i

i I

I I

O 25 50 75 100 GROUP 6 & 7 I

I I

I I

GROUP 5 1.

ROD INDEX-IS THE PERCENTAGE SUM OF THE WITHDRAWAL OF CONTROL R0D GROUPS 5,6, & 7.

4 ROD POSITION LIMITS FibuRE A-1 0

l m

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r

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1 E tc=if 20 2:a M

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12

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2 4

8 8

. 10 12 Elevation frca Bottc= of Core, it L O C A L ill l T E D i.! A X 'I t!Ul.1 A'L L 0il A B!. E

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Lil! EAR liEAT RATE FIF A-2 m

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