ML111540120

License Amendment Request - Reactivity Anomalies Surveillance

ML111540120
Person / Time
Site:	Limerick
Issue date:	06/02/2011
From:	Jesse M Exelon Generation Co, Exelon Nuclear
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML111540120 (18)

Text

10 CFR 50.90 June 2,2011 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555-0001 Limerick Generating Station, Units 1 and 2 Facility Operating License Nos. NPF-39 and NPF-85 NRC Docket Nos. 50-352 and 50-353 SUbject:

License Amendment Request - Reactivity Anomalies Surveillance In accordance with 10 CFR 50.90, Exelon Generation Company, LLC (Exelon) requests a proposed change to modify the Technical Specifications (TSs) concerning a change to the method of calculating core reactivity for the purpose of performing the reactivity anomaly surveillance at Limerick Generating Station, Units 1 and 2.

The proposed changes have been reviewed by the Limerick Generating Station Plant Operations Review Committee, and approved by the Nuclear Safety Review Board in accordance with the requirements of the Exelon Quality Assurance Program.

Exelon requests approval of the proposed amendment by June 2, 2012. Once approved, this amendment shall be implemented within 60 days of issuance. Additionally, there are no commitments contained within this letter. contains the evaluation of the proposed changes. Attachment 2 provides the marked up TS and Bases pages. The Bases pages are being provided for information only.

License Amendment Request Reactivity Anomalies Surveillance June 2011 Page 2 In accordance with 10 CFR 50.91, Exelon is notifying the State of Pennsylvania of this application for a license amendment by transmitting a copy of this letter and its attachments to the designated State Officials.

Should you have any questions concerning this letter, please contact Tom Loomis at (610) 765-5510.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 2nd day of June 2011.

Respectfully, ichael D. Jesse Manager, Licensin gulatory Affairs Exelon Generation Company, LLC : Evaluation of Proposed Changes : Markup of Technical Specifications and Bases Pages cc:

USNRC Region I, Regional Administrator USNRC Senior Resident Inspector, LGS USNRC Project Manager, LGS R. R. Janati, Commonwealth of Pennsylvania

ATTACHMENT 1 EVALUATION OF PROPOSED CHANGES

ATTACHMENT 1 EVALUATION OF PROPOSED CHANGES CONTENTS

SUBJECT:

Reactivity Anomalies Surveillance 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria

4.2 Precedents

4.3 No Significant Hazards Consideration 4.4 Conclusions

6.0 REFERENCES

Evaluation of Proposed Changes License Amendment Request Reactivity Anomalies Surveillance Page 1 1.0

SUMMARY

DESCRIPTION This evaluation supports a request to amend Facility Operating License Nos. NPF-39 and NPF-85 for Limerick Generating Station (LGS), Units 1 and 2.

The proposed change would revise the Technical Specifications to allow performance of the surveillance on a comparison of predicted to actual (or monitored) core reactivity. The reactivity anomaly verification is currently determined by a comparison of predicted vs. actual control rod density.

2.0 DETAILED DESCRIPTION The purpose of the reactivity anomaly surveillance is to compare the observed reactivity behavior of the core (at hot operating conditions) with the predicted reactivity behavior.

Currently, the LGS Technical Specifications (TSs) require that the surveillance be done by comparing predicted control rod density (calculated prior to the start of operation for a particular cycle) to an actual control rod density. The comparison is done, as required by the surveillance requirements. The proposed revision will change the method by which the reactivity anomaly surveillance is performed and not the specified frequency for performing the surveillance.

The current TS require that the reactivity equivalence of the difference between the actual rod density and the predicted rod density shall not exceed 1% ilk/k.

The proposed TS and Bases would be revised to state that the reactivity difference between the actual keffective (keff) and the predicted keff shall not exceed 1% ilk/k.

The current method of performing the reactivity anomaly surveillance uses rod density for the comparison primarily because early core monitoring systems did not calculate core critical keff values for comparison to design values. Instead, rod density was used as a convenient representation of core reactivity.

Allowing the use of a direct comparison of keff, as opposed to rod density, provides for a more direct measurement of core reactivity conditions and eliminates the limitations that exist for performing the core reactivity comparisons with rod density.

Marked up TS Bases pages are provided in Attachment 2, and are provided for information only.

3.0 TECHNICAL EVALUATION

If a significant deviation between the reactivity observed during operation and the expected reactivity occurs, the reactivity anomaly surveillance alerts the reactor operating staff to a potentially anomalous situation, indicating that something in the core design process, the manufacturing of the fuel, or in the plant operation may be different than assumed. This situation would trigger an investigation and further actions as needed.

License Amendment Request Reactivity Anomalies Surveillance Page 2 The current method for the development of the reactivity anomaly curves used to perform the TS surveillance actually begins with the predicted kelt at rated conditions and the companion rod patterns derived using those predicted values of kelt. A calculation is made of the number of notches inserted in the rod patterns, and also the number of equivalent notches required to make a change of +/-1 % Llk/k around the predicted kelt. The rod density is converted to notches and plotted with an upper and lower bound representing the

% Llk/k acceptance band as a function of cycle exposure. This curve is then used as the predicted rod density during the cycle. In effect, the comparison is indirect to critical kelt with a "translation" of acceptance criteria to rod density.

While being a convenient measurement of core reactivity, control rod density has its limitations, most obviously that all control rod insertion does not have the same impact on core reactivity.

For example, edge rods and shallow rods (inserted 1/3 of the way into the core or less) have very little impact on reactivity while deeply inserted central control rods have a larger effect.

Thus, it is not uncommon for reactivity anomaly concerns to arise during operations simply because of greater use of near-edge and shallow control rods than anticipated, when in fact no true anomaly exists. Use of actual to predicted kelt instead of rod density eliminates the limitations described above, provides for a technically superior comparison, and is a very simple and straightforward approach.

These proposed changes will not affect transient and accident analyses because only the method of performing the reactivity anomaly surveillance is changing, and the proposed method will provide a technically superior comparison as discussed above. Furthermore, the reactivity anomaly surveillance will continue to be performed at the current required frequency.

Consequently, core reactivity assumptions made in safety analyses will continue to be adequately verified, and no margins of safety will be reduced.

The following additional information is being provided concerning the core monitoring software as discussed in the Reference 1 request for additional information for the Edwin I. Hatch Nuclear Plant.

LGS utilizes the Global Nuclear Fuel (GNF) 3D MONICORE (Reference 2) core monitoring software system. The latest version of this product incorporates the PANACEA Version 11 (PANAC11) (Reference 3) core simulator code to calculate parameters such as core nodal powers, fuel thermal limits, etc., using actual, measured plant input data. PANAC11 is the same 3D core simulator code used in core design and licensing activities. When a 3D MONICORE core monitoring case is run, the core kelt (as computed by PANAC11) is also calculated and printed directly on each 3D MONICORE case output. This value can then be directly compared to the predicted value of core kelt as a measure of reactivity anomaly.

No plant hardware or operational changes are required with this proposed change.

Evaluation of Proposed Changes License Amendment Request Reactivity Anomalies Surveillance Page 3

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Reguirements/Criteria General Design Criteria 26, 28, and 29 require that reactivity be controllable such that sUbcriticality is maintained under cold conditions and specified applicable fuel design limits are not exceeded during normal operations and anticipated operational occurrences. The reactivity anomaly surveillance required by the LGS, Units 1 and 2 Technical Specifications serves to partly satisfy the above General Design Criteria by verifying that core reactivity remains within expected/predicted values.

Ensuring that no reactivity anomaly exists provides confidence of adequate shutdown margin as well as providing verification that the assumptions of safety analyses associated with core reactivity remain valid.

4.2 Precedents

A similar TS amendment was approved for:

1)

Letter from R. E. Martin (U.S. Nuclear Regulatory Commission) to M. J. Ajluni (Southern Nuclear Operating Company, Inc), "Edwin I. Hatch Nuclear Plant, Unit Nos. 1 and 2, Issuance of Amendments Regarding Revision to Technical Specifications Limiting Condition for Operation 3.1.2, "Reactivity Anomalies" (TAC NOS. ME3006 and ME3007)," dated November 4,2010.

Currently, the Dresden Nuclear Power Station, Units 2 and 3, LaSalle County Station, Units 1 and 2, and Quad Cities Nuclear Power Station, Units 1 and 2 TS use the kelt comparison.

4.3 No Significant Hazards Consideration Exelon Generation Company, LLC (Exelon) has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1.

Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed Technical Specifications changes do not affect any plant systems, structures, or components designed for the prevention or mitigation of previously evaluated accidents. The amendment would only change how the reactivity anomaly surveillance is performed. Verifying that the core reactivity is consistent with predicted values ensures that accident and transient safety analyses remain valid. This amendment changes the Technical Specification requirements such that, rather than performing the surveillance by comparing predicted to actual control rod density, the surveillance is performed by a direct comparison of kelt. Present day on-line core

License Amendment Request Reactivity Anomalies Surveillance Page 4 monitoring systems, such as the one in use at Limerick Generating Station (LGS), Units 1 and 2 are capable of performing the direct measurement of reactivity.

Therefore, since the reactivity anomaly surveillance will continue to be performed by a viable method, the proposed amendment does not involve a significant increase in the probability or consequence of a previously evaluated accident.

2.

Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

This Technical Specifications amendment request does not involve any changes to the operation, testing, or maintenance of any safety-related, or otherwise important to safety systems. All systems important to safety will continue to be operated and maintained within their design bases. The proposed changes to the reactivity anomaly Technical Specifications will only provide a new, more efficient method of detecting an unexpected change in core reactivity.

Since all systems continue to be operated within their design bases, no new failure modes are introduced and the possibility of a new or different kind of accident is not created.

3.

Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

This proposed Technical Specifications amendment proposes to change the method for performing the reactivity anomaly surveillance from a comparison of predicted to actual control rod density to a comparison of predicted to actual kef!. The direct comparison of kef! provides a technically superior method of calculating any differences in the expected core reactivity. The reactivity anomaly surveillance will continue to be performed at the same frequency as is currently required by the Technical Specifications, only the method of performing the surveillance will be changed. Consequently, core reactivity assumptions made in safety analyses will continue to be adequately verified.

Therefore, the proposed amendment does not involve a significant reduction in a margin of safety.

Based on the above, Exelon Generation Company, LLC, concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of no significant hazards consideration is justified.

Evaluation of Proposed Changes License Amendment Request Reactivity Anomalies Surveillance Page 5 4.4 Conclusions In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

5.0 ENVIRONMENTAL CONSIDERATION

A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

6.0 REFERENCES

1)

Letter from M. J. Ajluni (Southern Nuclear Operating Company, Inc.) to U.S. Nuclear Regulatory Commission, "Information on 3D Monicore Core Monitoring Software," dated October 5,2010.

2)

MFN-003-99, F. Akstulewicz (NRC) to G. Watford (GE), Safety Evaluation Report for GE Licensing Topical Report NEDC-32694P, "Power Distribution Uncertainties for Safety Limit MCPR Evaluations" (TAC No. M99069), March 11, 1999 [provides NRC acceptance of 3D MONICORE core surveillance system power distribution uncertainties].

3)

MFN-035-99, S. Richards (NRC) to G. Watford (GE), "Amendment 26 to GE Licensing Topical Report NEDE-24011-P-A, "GESTAR II" -Implementing Improved GE Steady-State Methods (TAC No. MA6481)," November 10, 1999 [provides NRC acceptance of PANACEA Version 11].

ATTACHMENT 2 Limerick Generating Station (LGS), Units 1 and 2 Markup of Technical Specifications and Bases Pages Revised Pages (Units 1 and 2) ii 1-7 3/41 B 3/41-1

)

I. 1

)....

............................. 1 I

OJlAPlHGR FLOW

)

1 4 MUlTIPl 4

ON SYSTEM RESPONSE TIME.......*.............* 1 6 REACTOR IONAl CONDITION CONDITION..*......................... 1 PHYS I

.. *. * * *. *..... *......... *.... *.... *... *........* 1

.JJU. BOUNDARY LEAKAGE.........*.**.*..................... 1 PRIMARY CONTAINMENT INTEGRITy.......*...*..****..*..........* 1 5

[\\V"LJJ CONTROL PROGRAM..........*............*.....*.....*.* 1 PU PURGING.................*....**...**......*....***... 1 6 D THERMAL POWER..........................................

~6 REACTOR ENCLOSURE SECONDARY CONTA NMENT INTEGRITy........*..* 1 6 ICAL POWER RATIO (MCPR) 1~4 LCULAT ON

~1ANUAl.**.**......*........**...**. 1 4 OPERABLE OPERABILITY.**............*......***.......**..... 1 4 MINIMU~1 I.

I.

I.

I.

I.

I.

I.

I.

I.

I. 33 l.

l. 5 I.

RECENTLY RRADIATED FUEL **.....*..........*.....*.*.*...*...* 1 6 FUELING FLOOR SECONDARY CONTAINMENT INTEGRITy..*.......****

1~6 1.37 REPORTABLE EVENT *......*.........*..*.....*..............*... 1 ~ 7 7a 1 7 1.38 1 ~ 7 1.39 SHUTDm<JN MARG IN.....*.....*............*.................*.*. 1 7 1.40 ITE BOUNDARy.*..........*......*....................*....*.. 1-7 1.41 SOURCE CHECK.......*....***....*...*..**....*.*...*...*.**... 1-7 LIMERICK

~ UNIT 1 i i Amendment No. JJ, 48,

~, 64, 192

\\nuedl L

n 0 RAB or nment AII 10

',HI 'HillAry containmen ha hes and [)lowout pa em i in compliance with the requirements r in h a 10 ed.

to ling floor mechanism a soci ted with each refueling floor econdary

ration,

.g., welds, bellows, or O-ring, is OPERABLE.

f.

rhe pre sure within the refueling floor secondary containment is less than or equal to the value required by ification 4.6.. 1.213.

1.37 A REPORTABLE EVENT hall be any of those conditions specified in Section to 10 CFR Part 50.

1.

rRICTED AREA mean an rea, a s to for the purpose of protecting individuals to radiation and radioactive materials.

rea used as residential quarters, but s buil ing may be apar;t~~~_,~~*,l whi h isl imited by the 1i cen ee in t undue ri ks from exposure ICTED AREA does not

~clude rate rooms in a residential 1.39 SHUTDOWN MARGIN hall be the amount of reactivity by which the reactor is ubcriti al or would be subcritical assuming all control rods are fully inserted for the single control rod of highest reactivity worth which is umed to be fully withdrawn and the reactor is in the shutdown condition; cold, i.e. 68°F; and xenon free.

1.40 fhe SITE BOUNDARY shall be that line as defined in Figure 5.1.3-113.

1.41 A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a radioactive source.

LIMERICK - UNIT 1 1-7 Amendment No. 4B,

~,

~,

~, 187

1 ttl i th

\\kl d.

itii th n I

hour of h

tivi I

plained and rform n analy i d termine nd explain the cause ration may ntinue 1

th,,; dif rence r

b.

Otherwi e.

be in at 8a t

HUTOOWN within the next 1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

provi ion of ERATION.

, and E AU RATI 110wing cat on 4.0.4 re not appl i able.

feet ve full power day during POWER t

t rtu rin

!\\

lea t onee per 31 J

• b.

LIMERICK - UNIT 1 3/4 1 Amendment No.

29

3/4. 1 REACTIVITY CONTROL SYSTEMS 3/4.1,1 SHUTDOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that (l) the reactor can be made succritical from all operating conditions. (Z) the reactivity transients associated with postulated accident conditions are controllable within acceptable limits. and (3) the reactor will be maintained sufficiently SUbcritical to preclude inadvertent criticality in the shutdown condition.

Since core reactivity values will vary through core life as a function of fuel depletion and poison burnup. the demonstration of SHUTOOWN MARGIN will be performed in the cold. xenon-free condition and shall show the core to be subcritical by at least R + O.38¥ A k/k or R + O.28¥ A k/k. as a~~ropriate, The O.3~ A k/k includes uncertainties and calculation biases.

The value of R in units of S A k/k is the difference between the calculated value of minimum shutdown margin during the operating cycle and the calculated shutdown margin at the time of the shutdown margin test at the beginning of cycle.

The value of R must be positive or zero and must* be determined for each fuel loading cycle.

Two different values are supplied in the Limiting Condition for Operation to provide for the different methods of demonstration of the SHUTDOWN MARGIN, The highest worth rod may be determined analytically or by test.'

The SHUTOOWN MARGIN is demonstrated by (an insequence) control rod withdrawal at the beginning of lite tuel cycle conditions, and. if necessary, at an~ future time in the cycle if the first demonstration indicates that the required margin could be reduced as a function of exposure.

Observation of subcriticality in this condition assures sUbcriticality with the most reactive control rod fully withdrawn.

This reactivity charac~ristic has been a basic assumption in the analysis of plant performance and can be best demonstrated at the time of fuel loading, but the margin must also be determined anytime a control rod is incapable of insertion.

--~.~%~

3/4,1.2 REACTIVITY ANOMALIES Since the SHUTDOWN MARGIN requirement for the reactor is s~11, check on actual conditions to the predicted conditions is nec~

changes in reactivity can be inferred from these comparisons o~'~~ro~~~"~~

Since the comparisons are easily done. frequent checks are not an 1

1 on normal operations.

A 1% change is larger than is expected for normal operation so a change of this magnitude should be thoroughly evaluated.

A change as large as 1% would not exceed the design conditions of the reactor and is on the safe side of the postulated transients.

LIMERICX - UNIT 1 a 3/4 1-1 AUG 8 i9ti

4 1 4 1 4 1

1 1 5 1 5 1

1-6 1-6 MAPLHGR LOW FACTOR)...........*..*............ I 4 POWER DEPENDENT MAPLHGR MULTIPL R)

PHYSICS T r~APFAC(

r*1APFAC(P)

LETED)...........................*......*................. 1 4 r~Er*1B R(

0 FH PUBLIC MINIMUM RITICAL POWER RAFIO (MCPR)

OF I

DOS CALCULATION MANUAL *..... '"....*...*..... '"

OPERABLE - OPERABILITy OP RATIONAL CONDITION - CONDITION............*..............

1.

L 1

L L

1.

1. 23

1. 24 1.

5

1. 6 1.

7 1.28 PRESSURE BOUNDARY LEAKAGE '"

1.29 PRIMARY CONTAINMENT INTEGRITy..............................*

1.30 PROCESS CONTROL PROGRAM.............................*.......

1.31 PURGE PURGING....*...........................*............

1.

RflTED FHERMAL POWER..........................*.*............

1. 33 REACTOR ENCLOSURE ECONDARY CONTAINMENT INTEGRITy 1 6 1.34 REACTOR PROTECTION SYSTEM RESPONSE TIME..*.................. 1 6

~~~~ (Continued)

LOW (POW

)

TR P FPOINT L

1.35 RECENTLY IRRADIATED FUEL..................*...*..........*.. 1 6

1. 36 REFUELING FLOOR SECONDARY CONTAINMENT INTEGRITy..*.......... 1-6

1. 37 REPORTABLE EVENT 1-7 1.37a 1 7 1.38 1-7 1.39 SHUTDOWN

~1ARGIN 17 1.40 SITE BOUNDARy...*........................................... 1-7 1.41 SOURCE CHECK............................*...............*... 1-7 LIMERICK - UNIT 2 ii Amendment No. l+, +6,

~,

153

(Con in ry con inment inment nd blowout panel are In compli nee wi h the requirements ling floor ry ling ontainment hanism

ration, i ted with each refueling floor secondary

.g., weld, bellows, r 0 rings, i OPERABLE.

f.

pres ure within the refueling floor secondary containment is less than r equal to the value red by ifi tion 4.6.5.1.2a.

PORTABLE EVENT hall be any of those conditions specified in Section 3

10 C R Part 50.

1.

7 RESTRICTED AREA mean n area, a the purpo of protecting individual radiation and radioactiv rna rial u

idential qua r, but be rt a a RESTRICTED AREA.

to which i limited by the licensee for again t undue ri ks from expo ure to RESTRICTED AREA does not include areas rate rooms in a residential building may e the number of control rod notches inse control rod notches.

All rods fully inserted 00% ROD DENS lTY.

1.39 SHUTDOWN MARGIN hall be the amount of reactivity by which the reactor is ubcritical or would be ubcritical a suming all control rods are fully in rted for the ingle control rod of highest reactivity worth which i

as umed to be fully withdrawn and the reactor is in the shutdown condition; cold, i.e. 6soF; and xenon free.

1.40 The SITE BOUNDARY hall be that line as defined in Figure 5.1.3-1a.

1.41 A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a radioactive source.

LIMERICK - UNIT 2 1-7 Amendment No. ++/-, 4g, B9, +4&, 148

W h eXl:eE~(l i ng c.k/k:

rform an analy i to determine and lain the cause operation may continue f the dif b.

Otherwi. be in at 1 st HOT SHUTDOWN wi thi n the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

rovi ion of RATIONS. and llowing CORE ive full power days during POWER OPERATION.

ification 4.0.4 re not applicable.

r t once Dur ng the f r At 1 b.

LIMERICK UNIT 2 3/4 1-2

3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 SHUTDOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that (1) the reactor can be made subcritical from all operating conditions, (2) the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and (3) the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.

Since core reactivity values will vary through core life as a function of fuel depletion and poison burnup. the demonstration of SHUTDOWN MARGIN will be performed in the cold, xenon-free condition and shall show the core to be subcritical by at least R + 0.381 4 k/k or R + 0.281 4 k/k. as appropriate.

The 0.381 4 k/k includes uncertainties and calculation biases.

The value of R in units of % 4 k/k is the difference between the calculated value of minimum shutdown margin during the operating cycle and the calculated shutdown margin at the time of the shutdown margin test at the beginning of cycle.

The value of R must be positive or zero and must be determined for each fuel loading cycle.

Two different values are supplied in the Limiting Condition for Operation to provide for the different methods of demonstration of the SHUTDOWN MARGIN.

The highest worth rod ma~ be determined analytically or by test.

The SHUTDOWN MARGIN is demonstrated by (an insequence) control rod withdrawal at the beginning of life fuel cycle conditions. and, if necessary, at any future time in the cycle if the first demonstration indicates that the required margin could be reduced as a function of exposure.

Observation of subcriticality in this condition assures sUbcriticality with the most reactive control rod fully withdrawn.

This reactivity characteristic has been a basic assumption in the analysis of plant performance and can be best demonstrated at the time of fuel loading, but the margin must also be determined anytime a control rod is i e of insertion.

3/4.1.2 REACTIVITY ANOMALIES Since the SHUTDOWN MARGIN requirement for the reactor is cheCK on actual conditions to the predicted conditions is nece changes in reactivity can be inferred from these comparisons of a

nSk Since the comparisons are easily done, frequent checks are not an imposltl0n on normal operations.

A 1% change is larger than is expected for normal operation so a change of this magnitude should be thoroughly evaluated.

A change as large as 1% would not exceed the design conditions of the reactor and is on the safe side of the postulated transients.

LIMERICK - UNIT 2 B 3/4 1-1