ML20206G378
| ML20206G378 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 11/10/1988 |
| From: | Johnson I COMMONWEALTH EDISON CO. |
| To: | Murley T Office of Nuclear Reactor Regulation |
| References | |
| RTR-REGGD-01.097, RTR-REGGD-1.097 5326K, NUDOCS 8811220254 | |
| Download: ML20206G378 (4) | |
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Chicago, Ilhnois 60690 November 10, 1988 Mr. Thomas E. Murley, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555
Subject:
Quad Cities Station Units 1 and 2
Tech Spec Changes in Support of HRC Regulatory Guide 1.97 and Detailed Control Room Design Review (DCRDR)
Modification - Clarification of Regulatory 1.97 Comitment for Reactor Water Level'*
H RC_Do citLHo si.JiDelli_ sad _iQ:2 tt5 Reference (a):
Letter f rom J.R. Wojnarowksi to D.B. Vassallo dated August 1, 1985
Dear Mr. Murley:
The purpose of this letter is to provide your staff with an update regarding items which are a result of Connonwealth Edison Company's (Ceco)
Regulatory Guide (R.O.) 1.97 Modification Program and Petailed Control Room Design Review (DCRDR) effort at its Quad Cities Station.
The modifications associated with k.G. 1.97 vere completed at Quad Cities Station. Work is continuing regarding the DCRDR program. CECO has performed a review of the Quad Cities 1 and 2 Technical Specifications (DPR-29 and 30), and the review has indicatsd that several items will require changes to the Technical Specifications in order to reflect the Regulatory Guide and DCRDR modifications. These changes are identified in Attachment A to this lotter.
A schedule will be developed for the sutweittal of these Technical Specifications in future discussions with the Quad Cities Licensing Project Manager.
Our review for R.G.
1.97 compliance has ident1.* led a comitment which requires clarification.
Reference (a) transmitted Ceco's R.G. 1.97 Compliance Final Supenary Report.
Specifically,Section IV, '* Variably Summary Table",
pages 23 and 24 of this report indicate that the following 1::?trument loops comply with R.G. 1.97 requirements for a B1 variable for reactor water level 003 8811220254 881110 PDR ADOCK 05000254 P
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T.E. Murley November 10, 1988 a
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57" to -243" 46" to -334" 358" to -42" r
only the -57" to -243" instrument loop complies with R.G. 1.97 requirements.
Ceco had not intunded on taking credit for the other two indicators and the reason the one loop is adequate is described below.
r A R.G. 1.97 8 variable is defined as an indication hacessary to allow the operator to determine whether a safety function is being accomplished.
R.3. 1.97 speelfles a reactor water 1evel indication range of 117" to -290" as the required range for this B1 variable. CECC has concluded that that existing redundant qualified instrument loops, with a e age of 57" to -243", meet the intent of this requirement. That is, the existjug quellfled instrument loops provide adequate indication to allow the operator to determine if a safety function is being accomplished, since no safety function is required above 57" or below -243".
The following paragraphs provide a bewis for this conclusion.
Reactor water level indication is not required above 57" to determine j
whether a safety function is being accomplished.
Existing trip setpoints enaure that high pressure injection systems will trip prior to exceeding the upper range of the existing qualified instrument (e.g., 57").
Reactor water level can be raised 57" utilising low pressure emergency core cooling systems.
However, reactor water level is normally s'intained at 30".
If reactor water level is raised to above the estating qual; fled range (e.g., 57"), adequate core cooling is maintained which would be indicated by the qualified level instruments (i.e., pegged upscale) and on scale flow indication which is also qualified.
ruthermore, reactor water level indication is not required below
-243" to determine ' nether a safety function is being acceeplished.
Luisting emergency operatinct procedures require the operator to initiate steam cooling j
when above 700 psig reactor pressure and at a reactor water level of -240" or below. While in the steam cooling mode, reactor pressure is monitored to ensure that adequate core cooling is being accomplished.
!! reactor pressure 7
1s reduced to below 700 psig while reactor water level is below -240",
then the operato-will initiate emerger.cy depressarisation. While in the emergency l
depressurization mode, seactor pressure is monitored to determine if a safety function is being accomplished. However, if the reactor is depressurised and l
reactor water level is still below -243", then the operator will continue to attempt restoration of reactor water level to 30" as quickly as possible i
l regardless of reactor water level.
As a result reactor water level indleation j
below -243" would not assist the operator in determining if a safety function is being accomplished.
1 1
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.g T.E. Murley November 10, 1988 Even though the estating configuration meets the Intent of k.G. 1.97, Ceco has initiated a modifier.tlon to respan the quallfled instrument loop to a range of 60" to.340".
Tht', modification would allow removal of the redundant unquallfled instrument m*Ltioned previously
'e.g.,
66" to 334") and provide for a greater reactor water level monitor ing range '.4Jer post accident conditions.
This modification was completed during the Unit 2 Spring 1988 outage and will be completed during the upcoming Summer 1989 outage.
Please direct any questions you may have regarding this matter, to this office.
Very truly yoars, gh j *kkitW I. M. Johnsoy Nuclear Licensing AM nist**ior im Attachment cci A.B. Davis - Regional A&ministrator, RI!!
T. Ross. NRR, Project Manager Quad Cities Resident Inspector 5326K
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ATTACIMarLA TECint1. CAL SPICIIICATICt!_ChA:sE6 AS SOC I ATED-.MITIL H RC_J EGU LATORLGUI D Ll. 91 AND_DCADR_H001 EICATICtiS.ALOU AILCIIIIS_ SIATICtl 1.
Unit 1/2 Table 3.2-4 Reretor water level range needs to be changed from -243 inches + 57 inches to
-340 inches + 60 inches.
Drywell temperature instrument readout 2.
Unit 2 Table 3.2-4 location change has been submitted to NRC to reflect moving the instruments from the 902-21 to 902-3 panel per Modificat.on H4-2-87-62.
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3.
Unit 1/2 Table 3.2-4 Torus water temperature readout location and tempe.ture range must be c'aanged from 90.;2)-21 to 901(2)-36 and fron 0-200'r to 0-300*r to eflect post-accident monitoring requisemente.
Torus water temperature readout 4.
Unit 1/2 Table 4.2-2 location must be changed from 901(2)-21 to 901(2)-36 to reflect post-accident monitoting requitements.
5.
Unit 1 Table 3.2-4 Drywell temperature instrument readout location must be changed from 901-21 to 901-3 when Modification H4-1-87-62 is completed during June 1989 refuel outage.
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