ML18046A972

Forwards Operating Reactors Branch Evaluation of SEP Topic IV-2 on Reactivity Control Sys.Response,Discussing as-built Facility Differences from Licensing Basis,Requested within 30 Days

ML18046A972
Person / Time
Site:	Palisades
Issue date:	10/16/1981
From:	Crutchfield D Office of Nuclear Reactor Regulation
To:	Hoffman D CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
References
TASK-04-02, TASK-4-2, TASK-RR NUDOCS 8110210192
Download: ML18046A972 (7)

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Docket No. 50-255 LS05-81* 10-026 Mr. David P. Hoffman Nuclear Licensing Administrator Consumers Power Company 1945 W. Parnall Road Jacksoo, Michigan 49201

Dear Mr. Hof1inan:

October 16, 1981

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SUBJECT:

SEP TOPIC IV-2, REACTIVITY CONTROL S.YSTEMS

• PALISADES NUCLEAR POWER PLANT Enclosed is a copy of our evaluation of SEP Topic IV*2, Reactivity Control Systems for Palisades. This assessment compares your facil fty, as described in Docket No. 50-255, with the criteria currently used by the regulatory staff for licensing new fac11 ities. Please fnfonn us ff your as-built

• factl tty differs from the 1 icensing basts assumed in our assessment.

Your response within 30 days of the date you.{;rece1ve this letter ts requested~

If no response is received within that t1me 9 we will assume that you have no comments or correction. This evaluation will be a basic input *to the inte-grated safety assessment for your fac11 i ty unless you tdenti fy changes needed to reflect the as-built conditions at your facility. This assessment.m~y be revised fn the future ff' your facility design is changed or ff NRC criteria relating to thf s subject are modified before the integrated assessment is

.comp_l eted.

• 1n future correspriridence regarding this topic, pl ease refer to the topic num-ber 1n your cover letter.

Sincerely, s-60' Dennis M. Crutchfield,. Chief

1; Operating Reactor*s Branch 115

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Enclosure:

As stated NRC FORM 318 (10-80) NRCM 0240 OFFICIAL RECORD COPY USGPO: 1981-335-960

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SAFETY EVALUATION REPORT SEP TOPIIV-2, REACTIVITY CONTROL SYST~

IN DING FUNCTIONAL DESIGN AND

- PROTECTION AGAINST SINGLE FAILURES.*

PALISADES NUCLEAR POWER PLANT DOCKET.NO. 50-255

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I * *lNTRODUCTI ON

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The purpose of. this-evaluation.is to insure. that the.design.basis for th_e_..-Palisades

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reactivity control systems is consistent with analyses performed to verify that the protection system meets General Design Criterion 25.

General Design Criterion 25

-~equires' that the reactor protection. ~ystem fre" design~'d *to as~~ure :tt1-~t :s,pecff~i*~-d :;_->:.

acceptable fuel design limits are not exceeded for any single malfunction of the reactivity control systems, _St,1ch ~s accidental withdr(iw_alof control.rod

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• Reacti.vitY cont~oT systems need not be single failure proof.

However, the protection system must be capable of assuring that acceptable fuel design limits are not*

e~ceeded in-the *event. ot' a '.'s fng.l e fa fi'ure r in °'the".,reacti vit/ "co~tro_l. s:ys te'ms*:.* -~ "rfr~

revre~* cr1terion*,-:co.vered -~i ~ ~tflts* :e.va-fuati'tiri',- is" ~ddress.-ed'- -fitsectio'ri0.1f:. =::*Re'vikw areas* that" are' not covered~ '.bu"t. are. rel'ateci.a~d essential to the' completion of this topic, are covered by other SEP topics addressed in Section III. The scope of the SEP to pi cs is defined in the 11Report on the Systematic Evaluation of Operating Facilities" dated November 25, 1977.

Thi~ repqrt is limited to the identification of inadvertent control rod withdrawals and malpositioning of.control rods which may occur as a result of single failure in the electrical circuits of the control-rod system.

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REVIEW CRITERION The review criterion for this topic is based upon Section 7.7, Part II of the NRC Standard Review Plan *. In. the specific case of the reactivity control systems a single failure shall not cause plant conditions more severe than those for which the reactor protection system is designed.

III.

..; 2 RELATED SAFETY TOPICS The following listed review areas_ are not covered in this report, but are related and* essential to the completion of this topic. These review areas are covered by other SEP topics as indicated below.

L. *Analyses of the consequences of control rod withdrawals and the malpositioning of control rods which may occur as ca result of single failures-in-the electrical

. circuits of the reactivity control systems are covered by SEP Topic XV-8, 11Control Rod Mi soperation (System Mal function or Operator Error) II

2. Analyses of reactivity insertions occurring as a result of inadvertent boron dilutions are covered in SEP Topic XV-10, 11Chemical and Volume Control System

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Mal function that Results in a Decrease in Boron Concentration in the Reactor.

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IV; REVIEW GUIDELINES The purpose of this evaluation is to identify inadvertent control rod withdrawals and malpositioning of control rods which may occur as a result of single failures in the electri~al circuits of the control rod system for the_Pali$ades Nuclear Power Pl ant.

V.". EVALUATION*

Information was provided in* Consumer~ Power Company° letter dated July *31,* 1981,.

describing des_ign features which limit control rod withdrawals and malpositioning

. of control rods caused by failures within the control rod system at the Palisades Nuclear Power Plant.

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.e Based upon an audit review of the informatjon provided by the licensee, we conclude that the following may occur as a result of single failures:

1)

• Any rod or group of rods rnay be inadvertently withdrawn or inserted until the*

plant trips.

(Shutdown rods must be withdrawn for a. group of regulating rods *to be moved).

2)

Regul ati.ng rods may he i~advertently withdrawn or inserted in ~equence until the plant trips.

(Shutdow~ rods must be withdra~n).

3) *A rod or. group Qf*rod~ cannot be withdrawn or.inserted when commande~ to be

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In AUTO, the reg~lati.ng rod sequence.(multiple rod motions)_ may be* activa~ed

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control.. switch* is released o~:.'plant trips). "

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Two rods. in different groups, a rod and part length group, or a rod and a group may be with drawn or inserted when a rod is manually wi_ ~~_dra\\'ln or inserted. ; "**.

. (Mo*v~ment *c*~;-tfnues *u~til manuaJ':confror.swittl1'1s**tele_a.sed ~~;*plant tr5ps).* --. * *-'1*

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Two" rods in same group, a 11 part l_ength rods or an entire group move when* an.

individual rod in that group is moved manually.

(Movement continues until manual.control switch _is released o_r plant.trips).

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A rod does not trove. when* its. group rroves.

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Two groups of rods may move instead of one group of rods.

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Regulating rods may be controlled automatically when under manual control.

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A shutdown rod may exceed exercise limits with regulating rods withdrawn.

11) A rod may drift into core.

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A rod or rods may ~rop into core.

The evaluation of rod motion is based upon the availabi~ity of interlock circuits associated with the rod control system such that certain con-sequential effects of single failures within the rod control system are precluded by the operability of these interlocks.

The basis for the assump-tion that these interlocks will be operable is that a failure in the inter-lock circuits will be identified and corrected during routine maintenance or as a result of system fault investigation. The effects of single failures occurring after an undetected failure has occurred in the inter-lock system are not included in the evaluation. This is consistent with the basis used for plants currently under operating license review.

The 12 types of control rod misop~ration can be characterized as rod in-sertion or withdrawal of one or more rods or group of rods (full or part-length).

SEP Topic XV-8 considered the effects of rod withdt~wals and rod drops of full length rods.

The range of reactivity worths analyzed bounds the potential reactivity changes from the single failures 6f full length'rods described above.

The staff concluded.in Topic.X_V-8 that cur-rent criteria were satisfied for these events.

Since part-length rods are no longer used for reactor control, topic XV~8 did not specifically address any malfunctions associated with them.

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~ause one or more part-length rods to be inserted into the reactor.

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drawal would not be of concern since the rods are positioned above the core.)

• Accordingly, the staff evaluated the effects on the reactor of insertion of

~part-length rods.. The part-length control rod design is. presented in Section J.3 of the FSAR.

The effects of malpositioning these rods is described in

• .Section 14.6 of the FSAR.

The part-length rods are similar in: design to the

full 1 engths -rods* over* the: fiYst *qua*r.ter of.:the:*~ro:d.1 ength*,- 'tnerefore the effects of inserti.ng.. a.part-length rod will be comparable to inserting a

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~:JuJ 1 _1 ength rod until the non~poi son. area *of the rod enters the core.. :.,. _:_;~;

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• power reductton with 'local. changes in power peaking: The'.rod drop analysis..

-*-. has demonstra_ted that the protection system can,accommodate these effects.

A core power reduction without a corresponding change in turbine demand would drop the co.ld leg temperature.

The operator would also have ind.ivi-dual-to-group deviatio~ alarms available if sbme part-length rods move, and rod position indications.

The upper and lower sections of the ex-core nuclear detectors provide addit~onal information concernJng axial flux distribution to the operator.

Thus there is reasonable assurance that part-lenth *rod malfunctions (rod drop;, undetected malpositioning) would not cause an event that would exceed fuel damage limits.

VI. CONCLUSION The staff has reviewed the reactivity control systems to identify control rod mis~peratio,n that may occur as *a ~esu1-t.of a sing1e:*ta.i1ure*. "'we.

concl~de that the licensee ha~ *adequ~tely asses~ed ~h~ ~ifects bf such*

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failures and that with respect to this topic; the Palisades* plant design

. meets applicable criteria.

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