ML20217G307

Proposed Tech Specs Re Distribution of Inoperable Control Rods

ML20217G307
Person / Time
Site:	FitzPatrick
Issue date:	10/08/1997
From:	POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:
Shared Package
ML20217G297	List: JPN-97-032, Forwards Application for Amend to License DPR-59,reflecting Change in LCO ML20217G299 ML20217G307
References
NUDOCS 9710090321
Download: ML20217G307 (10)

Text

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• l JAFNPP l

3.3.A.2 (cont'd) 4.3.A.2 (cont'd)

b. The control rod directional control valves for e. The scram discharge volume drain and vent valves shall be inoperable control rods shall be disarmed ' full-travel cycled at least once per quarter to verify that the electrically. valves close iniess than 30 seconds and to assure proper valve stroke and operation.

c. Control rods with scram times greater than thors permitted by Specification 3.3.C.3 are inoperable, f. An instrument check of control rod position indication shall be but if they ce be inserted with control rod drive performed once/ day.

pressure they need not be disarmed electrically.

d. Control rods with inoperable accumulators or those whose position cannot be positively determined shall be considered inoperable.

e. Inoperable control rods shall be positioned such that Specification 3.3.A.1 is met.

(1) When operating with two or more inoperable control rods in the Startup/ Hot Standby or Run modes at $_10% rated thermal power, control rod patterns shall be equivalent to those prescribed by the Banked Position Withdrawal Sequence (BPWS) or else the inoperable control rods shall be separated by two or more operable control rods. If this condition is not met, restore compliance with the condition within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

Otherwise be in hot shutdown within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

(?) If nine or more control rods are inoperable, be in hot shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

9710090321 971008 PDR ADOCK 05000333 P PDR Amendment No. 5,131,152,155,181, 90

JAFNPP l

3.3 and 4.3 BASES (cont'd) the control cell geometry and local k.. Therefore, an Also if damage within the control rod drive mechanism additional margin is included in the shutdown margin test and in particular, cracks in drivo internal housings, to account for the fact that the rod used for the cannot be ruled out, then a generic problem affecting a demonstration (the analytically strongest) is not necessarily number of drives cannot be ruled out. Circumferential the strongest rod in the core. Studies have been made cracks resulting from stress assisted intergranu!ar which compare experimental criticals with calculated corrosion have occurred in the collet housing of drives at criticals. These studies have shown that actual criticals several BtNRs. This type of cracking could occur in a can be predicted within a given tolerance band. For number of drives and if the cascks propagated until gadolinia cores the additional margin required due to severance of the collet housing occurred, scram could control cell material manufacturing tolerances and be prevented in the affected rods. Limiting the period of calculational uncertainties has experimentally been operation with a potentially severed collet housing will determined to be 0.38% Ak. When this additional margin assure that the reactor w.ll not be operated with a large .,

is demonstrated, it assures that the reactivity control number of rods with failed collet housings.

requirement in met.

B. Control Rods

2. Reactivity Margin - Inoperable Control Rods

1. Coupling verification is perforrred to ensure the control Specification 3.3.A.2 requires that a rod be taken out of rod is connected to the Control Rod Drive (CRD). The service if it cannot be moved with drive pressure. If the Surveillance reqses demonstrating a CRD does not go rod is fully inserted, it is in a safe position of maximum to the overtravel position whea it is fully withdrawn.

contribution to shutdown reactivity. If it is in a non-fully The overtravel position feature provides a positive check inserted position, that position shall be consisten'. with the on the coupling integrity since only an uncoupled CRD shutdown reactivity limitation s%ted in Specification can reach the overtravel position. The verification is 3.3.A.1. This assures that the core can be shut down at required to be performed any time a control rod is all times with the remaining control rods assuming the withdrawn to the " full out' (notch position 48) position strongest operable control rod does not insert. or prior to declaring the control rod to be OPERABLE after work on the control rod or CRD System that could Distribution of inoperable bypassed control rods will be affect coupling. This includes control rods inserted one limited to ensure the consequences of a Control Rod Drop notch and then returned to the Accident are acceptable.

Amendment No. 15,155,103, 99

, Attachm:nt 11 to JPN 97-032 SAFETY EVALUATION FOR PROPOSED TECHNICAL SPECIFICATION CHANGES REGARDING DISTRIBUTION OF INOPERABLE CONTROL RODS New York Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 CPR-59

, , Att:chment 11 to JPN 97 032 Distribution of Inoperable Control Rods Safety Evaluation Page 1 of 4

1. Df40RIPTION OF THE PROPOSED CHANGES The iollowing proposed change to ths James A. FitzPatrick Technical Specifications revises the actions to be taken in the event multiple control rods are inoperable. The asso lated bases statement is revised to reflect this change in the Limiting Conc!ition for Operation.

ER29.1Q Revise Specification 3.3.A.2.e to road:

Inoperable cor. trol rods shall be positioned such that Specification 3.3.A.1 is met. ,

(1) When operating with two or more inoperable control rods in the Startup/ Hot Standby or Run modes at .110% rated thermal power, control rod patterns shall be equivalent to those prescribed by the Banked Position Withdrawal Sequence (BPWS) or else the inoperable control rods shall be separated by two or more operable control rods, if this condition is not met, restore compliance with the condition within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Otherwise be in hot shutdown within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. '

(2) If nine or more control rods are inoperable, be in hot shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Paae SR Replace the paragraph within Bases Section 3.3 and 4.3 A.2.e that states, Inoperable bypassed rods will be limited within any group to not more than one control rod of a (5x5) twenty-five control rod array, with:

Distribution of inoperable bypassed control rods will be limited to ensure the consequences of a Control Rod Drop Accident are acceptable.

11. PURPOSE OF THE PROPOSED CHANGES The proposed changes revise the iimit on inoperable control rods to be consistent with the assumptions used in evaluation of the control rod drop accident (CRDA) and associated bases.

l

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Att%hment 11 ts JPN 97-032 Distribution of inoperable Control Rods Safety Evaluation

. Page 2 of 4 111. SAFETY IMPLICATIONS OF THE PROPOSED CHANGES Currently Technical Specifications provide no basis for limiting inoperable control rods to not more than one in any 5x5 array. The specification controls the distribution of inoperable control rods and effectively limits the total number to no more than seven. The proposed change to Specification 3.3.A.2.e ties the allowable number and distribution of inoperable control rods to the assumptions used in analysis of the CRDA. The generic banked position withdrawal sequence (BPWS) analysis (reference 1) has evaluated the effect of up to eight fully inserted, out of sequer.ce control rods on control rod reactivity. Provided the out of sequence, inserted control rods are separated from each other by at least two rods in all directions, including the diagonal, the consequences of a Control Rod Drop Accident (CRDA) will continue to be acceptable. Technical Specification 3.3.8.3.e requires that FitzPatrick be operated with control rod patterns that are equivalent to those prescribed by the BPWS. Therefore, this change ensures consistency between rod patterns allowed by the Technical Specifications, those that have been evaluated in the core reload analysis (reference 2), and supporting accident analyses.

Explicitly limiting the number of inoperable control rods to eight, rather than the present implicit limit of seven has no significant effect on the level of safety of the l plant. This number is consistent with the bases of the CRDA analysis and is n.lficiently low that the plant will be shutdown if there is a potential systematic problem with the Control Rod Drive system. The compensatory action of placing the plant in hot shutdown in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is more limiting than the present requirement to reach the cold condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The important parameter with respect to controi rod operability is reaching the shutdown condition (v h O necessitates all moveable rods being inserted, wit;i a rod withdrawal block applied), rather than temperature of the Reactor Coolant System (i.e., hot or cold).

The proposed change also revises the power level at which limitations are placed on inoperable control rod distribution from "during reactor power operation" (> 1 %

rated thermal power) to "s 10% rated thermal power". Distribution controls for inserted inoperable control rods are only important for analysis of the CRDA since inserted rods are airaady performing their safety function of providing scram reactivity. The CRDA only has significant effects at low power (s 10%). Extending the range of application of incperable rod distribution controls to below 1 % rated thermal power improves the level of safety of the plant. Removing controls on inoperable rod distribution above 10% rated thermal power has no safety consequence, since the CRDA has no significant effect at these power levels.

The proposed change establishes time limits for restoring BPWS equivalent rod patterns. Currently, there is no limit on the time the reactor may be operated wnh non-complying rod ptterns provided no rod movement occurs (except by scram).

This change will increase the level of safety, since it assures returning the plant to an analyzed condition or taking appropriate compensatory action. The compensatory action of placing the plant in hot shutdown in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is more limiting than the present requirement to reach the cold condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Att:chment il to JPN-97 032 Distribution of inoperable Control Rods Safety Evaluation Page 3 of 4 IV. EVAU)ATION OF SIGNIFICANT HAZARDS CONSIDERATION Operation of the FitzPatrick plant in accordance with the proposed Amendment -

would not involvo a significant hazards consideration as defined in 10 CFR 50.92, since it would not:

1. involve a significant increase in the probability or consequences of an accident previously evaluated because:

The number and distribution of inoperable control rods is not a precursor to any accident, therefore the probability of an accident is not affected. The proposed changes assure the assumptions used in evaluation of accidents are satisfied, therefore there will be no increase in the consequences of an accident previously evaluated.

2. create the possibility of a new or different kind of accident from any accident previously evaluated because:

Changing the allowable number and distribution of inoperable control rods and the powei level at which these limits apply to be consistent with the accident analyses does not create the possibility of a new or different kind of accident.

3. involve a significant reduction in a roargin of safety because:

The proposed changes assure the assumptions used in the accident analyses are satisfied, therefore there will be no affect on the margin of safety as a result of these changes.

V. IMPLEMENTATION OF THE PROPOSED CHANGES This amendment request meets the eligibility cr.teria for categorical exclusion set forth in 10 CFR 51.22(c)(9) as follows:

(i) the amendment involves no significant hazards consideration.

As demonstrated in Section IV of this evaluation, the proposed change involves no significant hazards consideration.

(ii) there is no significant change in the types or significant increase in the amounts of any effluents that may be released offsite.

The proposed change ensures consistency between the plant's Technical Specifications and the supporting accident analyses. Therefore, there will be no change in the types or amounts of effluents that may be released offsite.

o

Attachment il t2 JPN 97 032 Distribution of inoperable Control Rods Safety Evaluation Page 4 of 4 (iii) there is no significant increase in individual or cumulative occupational radiation exposure.

The proposed change ensures consistency between the plant's Technical Specifications and the supporting accident analyses. Therefore, there will be no effect on individual or cumulative radiation exposure.

Based on the above, it is concluded that there will be no impact on the environment l

- resulting from the proposed change and the proposed change meets the criteria specified in 10 CFR 51.22 for a categorical exclusion from the requirements of 10

- CFR 51.21 relative to reiluiring a specific environmental assesstnent by the Commission.

Additionaliy, implementation of the proposed change will not adversely afiect the Fire Protection Program at the FitzPatrick Plant.

VI. CONCLUSIONS Based on the discussions above, implementation of Technical Specification changes governing the distribution of inoperable control rods does not inutve a significant hazards consideration, or an unreviewed safety question, and ><ill not endanger the health and safety of the public. The Plant Operating Review Comm ttee and Safety Review Committee have reviewed this proposed Technical Specification change and agree with this conclusion.

Vll. REFERENCES

1. NEDO-21231, " Banked Position Withdrawa! Sequence," January 1977

2. NEDE-24011-P A-13, " General Electric Standard Application for Reactor Fuel," August 1996

, Attachment til to JPN 97-032

• MARKUP OF TECHNICAL SPECIFICATION PAGE CHANGES PROPOSED TF.CHNICAL SPECIFICATION CHANGES REGARDING Dl3TRIBUTION OF INOPERABLt" CONTROL RODS l

I New York Power Authority

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~ JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 DPR-59 s

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l JAFNPP 3.3 and 4.3 BASES icent'd) the control cell pometry and local k . Therefore, an Also if damage within the control rod drive mechanism and additional margin is included in the shutdown margin test to in porticular, c<acks in drive internal housings, cannot be account for the fact that the rod unsd for the demonstration ruled out, then a gerr,ric problem affecting a number of (the analytically strongest) is not r-marily the strongest drives cannot be ruled out. Circumferential cracks resulting rod in the core. Studies have been made which compare from stress assisted interg,ranular corrosson have occurred expenmental criticale with emiendated criticals. These in the collet houssng of drives at several BWRs. This type studies have shown that actual criticals can be predicted of cracking could occur in a number of dnves,and if the within a given tolerance bond. For gedalense cores the cracks propagated until severance of the catet housmo additional energin required due to control cell material occurred, scram could be prevented in the effected rods.

manufacturingtolerancesandemac dationoluncertaintieshes Limiting the penod of operation with a potentially severed expenmentaNy been determined to be 0.38% Ak. When collet housing will assure that the toector will not in this additional margin is demonstrated, ist soeures that the operated with a large number of rods with failed comet reactivity control requirement is mot. houesngs.

2. Reactivity hierge - Inoperable Control Reds B. Control Rods Specefication 3.3.A.2 requires that a rod be taken out of 1. Coupleng verification is performed to ensure the control rod service if it cannot be moved with drive pressure. If the rod is connected to the Control Rod Drive ICRD). The is fully inserted, it is in a safe poestion of mexemum Surveellence requires demonstrating a CRD does not go to contribution to shutdown reactewity. If it is in a non-fully the overtravel poestson when it is fully withdrawn. The inserted pesP.icr., that position shen be consistent with the overtravel poesteon foeture provides a positive check on the shutdown reactewity hmetation stated in Specification couping integrity smce <wdy an uncoupled CRD can ree.ch 3.3.A.1. This assures tiset the core can be shut down at all the overtravel posetion. The verification is required to be times with the remesning control rods assumeng the performed any time a control rod is withdrawn to the "fuE strongest operable control rod does not inee-t. out* Inotch position 48) position or prior to declaring the control rod to be OPERABLE after work on the control rod

. :;- --r: ";i;:::: f ;;i " 5: 1--2:;d ..:C. s,

c; or CRD System that could affect coupleng. This includes l5:5;: ..;i S control rods inserted one notch and then retumed to the l :s .;{.];].; . e. eut;;' ;;d O' Dide;bloo of icopeerdle Q % e d cc M d 5 wil 4: tim ked to cosure Oe cc,nsepence5 td a coAwt M %g seeideA ue acccfchte.

Amendment No. ,1 , 44+-

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