ML20079C554
| ML20079C554 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 06/14/1991 |
| From: | Beck G PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20079C555 | List: |
| References | |
| NUDOCS 9106240008 | |
| Download: ML20079C554 (7) | |
Text
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l'Illt,AI)El.I'llIA El.ECTRIC COMi%NY NUCLEAR GROUP liEADQUARTERS 955-65 C11ESTERI1 ROOK llLVD.
WAYNE, PA 19087 5691 (215) 640-6000 NUCLEAR ENGINI1RINO & St RVICES DEPARTMENT June 14, 1991 Docket No. 50-278 License No. DPR-56 U.
S.
Nuclear Regulatory Commission ATTH: Document Control Desk Washington, DC 20555
Subject:
Peach Bottom Atomic Power Station, Unit 3 Technical Specification Change Request Dear Sirt Philadelphia Electric Company (PEco) hereby submits Technical Change Request 91-05, in accordance with 10 CFR 50.59, requesting an amendment to the Technical Specifications (Appendix A) of Operating Licence No. DPR-36 Information supporting this change request is contained in Attachment 1 of this letter and the proposed replacement pages are contained in Attachment 2.
This submittal requests a temporary change to the Technical Specifications to ensure that the maximum generating capacity is available for the peak electrical demand months of July and August and to avoid an economic penalty of 1.3 million dollars in replacement generation costs.
Specifically, Technical Specification 3.3.B.1 states: "Each control rod shall be coupled to its drive or completely inserted and the control directional valves disarmed electrically Philadelphia Electric Company by this request is seeking to have this Technical Specification changed for Unit 3 until the unit is shutdown for its 8th refuel outage.
This is scheduled for September 7,
1991.
Philadelphia Electric Company is coquesting that the proposed changes be processed in an exigent basis in accordance with 10 CFR 50.91, paragraph (a) (6).
The exigent circumstances are related to the loss of approximately 4 percent of the electrical output of Unit 3 and the resulting need for high cost replacement power during
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the summer months.
This cost is estimated at 1.3 million dollars.
This derating is especially critical during the summer months when
% l electrical demand is at its highest.
This May the PJM electrical grid
(
was in a maximum emergency generating situation for two consecutivo days.
While such an emergency is presently not occurring the potential for such as situation is high.
Accordingly, we are
)
9106240003 910614 i'
PDR ADOCK 05000278
- i P
i U. S *. Nuclear Regulatory Commission June 14, 1991 PBAPS, Units 2 and 3 Page 2
- TSCR 91-05 requesting that this TS Change Request be processed on an exigent basis.
If you have any questions regarding this matter, please contact us.
Very truly yours, l
/,..- ( (/
G. J.
Beck, Manager Licensing Section Nuclear Engineering and Services Attachments cc:
T.
T. Martin, Administrator, Region I, USHRC J.
J. Lyash, USNRC Senior Resident Inspector, PB J
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4 4
COMM011WEALTil OF PE!1118YINAlli A:
nn.
COU!1TY OF CllESTER D.
R. llelwig, being first duly swotn, deponen and nnyn:
That19 in Vice President of Philadelphin El ec t.ric Company; that he has rond the foregoing Application for Amnndment of Facility Operating Licenne 11o. DPR-56 (Technical Specification Change Requent flo. 91-05-0) to eliminate the seguirement of verifying coupling of a control rod, and Itnown the contentn thereof; and that the statements and mnttern net forth therein are true and correct to the bent of hin itnowledge, information and belief.
I D1 L
_.____/a
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Vice Presideit Subncribed and nworn to thin /'bday before me of
,Ju s c, 1991.
Ltbun e euk_])!Jmal-tiotary Public NOTARAL CE.AL CATHEM.E A VCNDEZ NNyy Puu c T e34'n Tep. Chew C:wf My Ccyst en Espees De d FM3
ATTAi"iMENT 1 PEACil BOTTOM ATOMIC POWER STATION UNIT 3 Docket No 50-278 Licenso No. DPR-56
" TEMPORARY ELIMINATION OF Tile REQUIREMENT TO VERIFY COUPLING OF A CONTROL ROD"
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--4+-G-4*aoMASM--G1M JAA > & M mn--MW" TSCR 91-05 Juno 14, 1991 Page 2 Introduction control Rod Drive (CRD) 38-23 was identified as being uncoupled when it was withdrawn to the fully withdrawn position "48" while performing the wookly CRD exerciso that is required by Technical Specification 4.3.A.2.a.
CRD 38-23 became uncoupled under normal drive water pressure.
Soveral attempts to recouple woro made by giving notch insert signals from position 48 to 46 using normal drivo pressure as directed under the control Rod Uncoupled Proceduro.
Those coupling attempts were unsuccessful.
The rod was able to bo inserted with no difficulty to position 00.
While withdrawing the rod to verify coupling integrity, it was observed that the rod would withdraw normally and settle at any position from 00 to 46 without incident.
The only valid method of control rod to control rod drive coupling verification is to initiato a withdraw from position "48" and verify that the CRD does not go to the over travel position.
Because coupling of the control rod and the drivo cannot be confirmed it must be assumed that the rod is uncoupled.
The rod was the isolated in i
the full-in position and declared inoperable por Technical l
Specification 3.3.B.l.
Philadelphia Electric Company is sooking a technical specification chango to allow control rod 38-23 to be withdrawn to position 46 for the romainder of fuel cycle 8.
This will eliminato a power dorato of approximately 4% by allowing the withdraw of rod 38-23 and its symmetric partners.
pgscription of Channes Licenseo proposes the following changes:
1)
Limiting Condition for Operation 3.3.B.1 currently states:
"Each control rod shall be coupled to its drive or completely inserted and the control rod directional control valvos disarmed electrically.
This requirement does not apply in the refuel condition when the reactor is vented.
Two control rod drives may be removed as j
long as Specification 3.3.A.1 is mot."
l The licenseo proposes to modify the above LCO to stato:
"Each control rod shall be coupled'to its drive or completely inserted and the control rod directional control valves disarmed electrically, except as in 3.3.B.l.a2 This requirement does not apply in the refuel condition when the reactor is vented.
Two control rod drives may be removed as long as specification 3.3.A.1 is mot."
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TSCR 91-05 Junn 14, 1991 Page 3 2)
The licensco proposes to add the following statement, immediately after 3.3.B.l.
This new statement should be numbered 3.3.B.l.a:
"For control rod 38-23, fo-tha remainder of cycla 3 (to be completed before 19, 31.,1).
If couoling c.annot be accomplished, the uncou vet control rod ray be withdrawn when 2 10% of r 'ei :hetual power only if all the following condition. a,e sat.4 itielt 1) no other uncouplec cont 101 roc'. 's withara-r, 2) the uncoupled control rod may..St be wi thdracn past notch position /S.
3)
The licenseo proposes the add: n w. of the t.J1owing statement to be numbered 4.3.9.
When repositioning the uncoupled control rod, per Specification 3.3.B.1.a the uncoupled control rod's position shall be verified to have followed the control rod drive by neutron instrumentati7n (LPRM or TIP).
If the control blade cannot be verified to have followed the drive out to its final position, then the rod shall be completely inserted and the control rod directional control valves disarmed electrically.
Safety Discussion Change request (1) is an administrative change to create and exception for control rod 38-23.
Change request (2).dotails the specific request the Philadelphia Electric company is making.
Control rod coupling integrity is required to ensure compliance with the analysis of the control rod drop accident (CRDA).
For the CRDA, the faulty control rod is assumed to not be coupled to the CRD and that it sticks in the fully inserted position.
The rod then is presumed to become unstuck and drop to the position of the withdrawn CRD.
Above 10% of the rated thermal power the consequences of a CRDA are negligible, therefore the control rod coupling integrity will not increase the consequences of the CRDA accident.
The control rod will not be withdrawn-past position "46",
This limitaition will minimize the mechanical loads on the control rod to CRD spud / socket coupling mechanism.
Change request (3) details the additional compensatory actions PECo is proposing to initiate if rod 38-23 is withdrawn.
The neutron monitoring systems of the Local Power Range Monitoring (LPRM) systems and the Transversing Incore Probe (TIP) will verify the position of the control rod blade within the reactor. Verifing the control rod position with the neutron montioring system will be an indication that the control rod is l
f
TSCR 91-05 June 14, 1991 Pags 4 following the CRD and that the control rod is not stuck in the reactor.
Eo Significant llazards Consideration The following evaluation is provided in accordance with the requirements of 10 CFR 50.92:
1:
D2Rs the orocosed amendment involve a sianificant increase in the probability of consecuences of an accident previousiv evalualed No, the proposed amendment does not involve a significant increase in the probability or consequences of an accident previcusly evaluated.
This amendment incorporates compensatory actions in the Technical Specifications to assure that even with an uncoupled rod the rod position is known, that no other uncoupled rods are withdrawn, and that scram performance remains intact.
2:
Does the proposed amendment create the lossibility of a new or different kind of accident from any accident orsyiously evelaated?
No, the proposed amendment does not create the possibility of a new or different kind of accident from any previously evaluated.
The compensatory measures included in the Tachnical Specification changes assure that no new or
- fferent kind of accident is possible.
3:
Dggs the oronosed amendment involve a sianificant reduction in the marain of safety?
No, the proposed amendment does not involve a significant i
reduction in the margin-of safety as the limiting event is the CRDA and all fuel limits stipulated in that analysis will be met when the compensatory measures included in Technical specification changes are implemented.
Environmental Considerations This change will have no impact on the environment, either onsite or offsite.
Conclusion The Plant Operations Review Committee and the Nuclear Review Board have reviewed these proposed changes to the Technical Specifications and have concluded that they do not involve unreviewed safety questions, significant hazards consilerations or a environmental consideration and they will not endanger the health and safety of the public.
2160a2. doc
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