Text

Monthly Operating Repts for Mar 1989 for Quad Cities Nuclear Power Station
	ML20245A231
Person / Time
Site:	Quad Cities
Issue date:	03/31/1989
From:	Deelsnyder L, Robey R; COMMONWEALTH EDISON CO.
To:	; NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
	RAR-89-18, NUDOCS 8904250108
	Download: ML20245A231 (54)
	v • d • e

. -.

~}-

. =

-.j.

QUAD-CITIES NUCLEAR POWER STATION UNITS 1 AND 2 MONTHLY PERFORMANCE REPORT MARCH, 1989 COMMONWEALTH.-EDISON COMPANY AND IOWA-ILLIN0IS GAS & ELECTRIC COMPANY NRC DOCKET NOS. 50-254 AND 50-265.

LICENSE NOS. OPR-29 AND DPR-30 i

8904250108 890331

?"

PDR ADOCK 05000245 &

R PNV F

j k

0027H/0061Z

/%

i I

TABLE OF CONTENTS I.

Introduction II.

Summary of Operating Experience A.

Unit One B.

Unit Two i

III.

Plant or Procedure Changes, Tests, Experiments, and Safety Related Maintenance A.

Amendments to Facility License or Technical Specifications B.

Facility or Procedure Changes Requiring NRC Approval C.

Tests and Experiments Requiring NRC Approval D.

Corrective Maintenance of Safety Related Equipment IV.

Licensee Event Reports V.

Data Tabulations A.

Operating Data Report B.

Average Daily Unit Power Level C.

Unit Shutdowns and Power Reductions VI.

Unique Reporting Requirements A.

Main Steam Relief Valve Operations B.

Control Rod Drive Scram Timing Data VII.

Refueling Information VIII.

Glossary 0027H/00612

I.

INTRODUCTION Quad-Cities Nuclear Power Station is composed of two Boiling Water Reactors, each with a Maximum Dependable Capacity of 769 MWe Net, located in Cordova Illinois.

The Station is jointly owned by Commonwealth Edison Company and Iowa-Illinois Gas & Electric Company. The Nuclear Steam Supply Systems are General Electric Company Boiling Water Reactors.

The Architect / Engineer was Sargent & Lundy, Incorporated, and the primary construction contractor was United Engineers & Constructors.

The Mississippi River is the condenser cooling water source.

The plant is subject to license numbers DPR-29 and DPR-30, issued October 1, 1971, and March 21, 1972, respectively; pursuant to Docket Numbers 50-254 and 50-265. The date of initial Reactor criticalities for Units One and Two, respectively were October 18, 1971, and April 26, 1972. Commercial generation of power began on February 18, 1973 for Unit One and March 10, 1973 for Unit Two.

This report was compiled by Lynne Deelsnyder and Verna Koselka, telephone number 309-654-2241, extensions 2185 and 2240.

i I

0027H/0061Z

)

II.

SUMMARY

OF OPERATING EXPERIENCE l

A.

Unit One Unit One began the month of Mcrch operating at full power. On March 2, j

at 2122 hours0.0246 days 0.589 hours 0.00351 weeks 8.07421e-4 months , the unit was placed in Economic Generation Control (EGC).

On March 3, at 0455 hours0.00527 days 0.126 hours 7.523148e-4 weeks 1.731275e-4 months , the unit was taken off EGC and an increase to full load was taken with recirculation pumps. At 0610 hours0.00706 days 0.169 hours 0.00101 weeks 2.32105e-4 months , 820 MWe was achieved.

Until March 4, power levels were held constant with control rod maneuvers. At 0535 hours0.00619 days 0.149 hours 8.845899e-4 weeks 2.035675e-4 months , load was reduced to 750 MWe, then increased to 825 MWe, and at 2055 hours0.0238 days 0.571 hours 0.0034 weeks 7.819275e-4 months , the unit was placed in EGC.

On March 5, at 1530 hours0.0177 days 0.425 hours 0.00253 weeks 5.82165e-4 months , the unit was taken off EGC to perform the Control Rod Drive Monthly Surveillance. At 1805 hours0.0209 days 0.501 hours 0.00298 weeks 6.868025e-4 months , the surveillance was completed, and at 1819 hours0.0211 days 0.505 hours 0.00301 weeks 6.921295e-4 months , the unit was placed in EGC. The unit remained in EGC until March 6.

At 0147 hours0.0017 days 0.0408 hours 2.430556e-4 weeks 5.59335e-5 months , the unit was taken off EGC to perform routine weekly surveillance. At 1904 hours0.022 days 0.529 hours 0.00315 weeks 7.24472e-4 months , all surveillance were completed and the unit was placed in EGC.

On March 7, at 0510 hours0.0059 days 0.142 hours 8.43254e-4 weeks 1.94055e-4 months , the unit was taken off EGC and an increase to full load was taken.

Power levels were held constant until 1906 hours0.0221 days 0.529 hours 0.00315 weeks 7.25233e-4 months when power levels were adjusted and the unit was placed in EGC.

From March 8 thru March 18, the unit operated near full power or remained in ECC. On March 18, at 2205 hours0.0255 days 0.613 hours 0.00365 weeks 8.390025e-4 months , load was reduced to 600 MWe at the request of the Chicago Load Dispatcher.

Power levels were held constant while the Main Steam Isolation Valve Closure Time Quarterly Surveillance was performed.

At 0030 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months , on March 19, the testing was completed, power levels were adjusted, and at 0241 hours0.00279 days 0.0669 hours 3.984788e-4 weeks 9.17005e-5 months , the unit was placed in EGC.

At 1310 hours0.0152 days 0.364 hours 0.00217 weeks 4.98455e-4 months , the unit was taken off EGC at the request of the Load Dispatcher and a load reduction to 600 MWe was taken.

At 1725 hours0.02 days 0.479 hours 0.00285 weeks 6.563625e-4 months , a power increase to 750 MWe was taken with recirculation pumps and at 2010 hours0.0233 days 0.558 hours 0.00332 weeks 7.64805e-4 months , the unit was placed in EGC.

For the remainder of the month, the unit remained in ECC or operated near full power with minor interruptions to perform routine surveillance.

B.

Unit Two Unit Two began the month of March operating in Economic Generation Control (EGC).

On March 2, at 0010 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months , the unit was taken off EGC and a load reduction to 745 MWe was taken in preparation for a Reactor Feed Pump swap.

At 0222 hours0.00257 days 0.0617 hours 3.670635e-4 weeks 8.4471e-5 months , the feed pump swap and all surveillance were completed and the unit was placed in EGC. At 0327 hours0.00378 days 0.0908 hours 5.406746e-4 weeks 1.244235e-4 months , a drywell floor drain sump (DWFDS) hi level alarm was received in the control room and it was noted that the DWFDS pump total for the last four hours had been

800 gallons. At 0835 hours0.00966 days 0.232 hours 0.00138 weeks 3.177175e-4 months , EGC was tripped and a power reduction to 300 MWe was taken in preparation for an entry into the drywell. At 1241 hours0.0144 days 0.345 hours 0.00205 weeks 4.722005e-4 months , 299 MWe was reached and at 1305 hours0.0151 days 0.363 hours 0.00216 weeks 4.965525e-4 months , after deinerting, drywell entry was made. The 3C Electromatic Relief Valve vacuum breaker was 1 discovered open. The vacuum breaker was closed manually and subsequently opened intermittently due to steam leakage past the 3C Electromatic Relief Valve..At 1918 hours0.0222 days 0.533 hours 0.00317 weeks 7.29799e-4 months , unit shutdown procedures were initiated. At 2120 hours0.0245 days 0.589 hours 0.00351 weeks 8.0666e-4 months , the generator was taken off-line and at 2137 hours0.0247 days 0.594 hours 0.00353 weeks 8.131285e-4 months , the reactor was scrammed by normal shutdown procedure.

From March 3 thru March 5, maintenance was performed on the. failed 3C Electromatic Relief Valve and the vacuum breaker that hung up.

On March 5, at 0251 hours0.00291 days 0.0697 hours 4.150132e-4 weeks 9.55055e-5 months , the mode switch was placed in STARTUP and at 0642 hours0.00743 days 0.178 hours 0.00106 weeks 2.44281e-4 months , the reactor was made critical. At 1447 hours0.0167 days 0.402 hours 0.00239 weeks 5.505835e-4 months , the generator was synchronized to the grid. An ascent to full power was begun with control rods and recirculation pumps. On March 6, at 0727 hours0.00841 days 0.202 hours 0.0012 weeks 2.766235e-4 months , full power was achieved.

Power levels were held constant while Reactor Core Isolation Cooling and High Pressure Coolant Injection testing was performed. On March 7 all surveillance were completed and at 1430 hours0.0166 days 0.397 hours 0.00236 weeks 5.44115e-4 months , power levels were adjusted and the unit was placed in EGC.

From March 7 thru March 22, normal operational activities occurred with the unit operating near full power or remaining in EGC with minor interruptions to perform routine surveillance. On March 22, at 2205 hours0.0255 days 0.613 hours 0.00365 weeks 8.390025e-4 months , a power reduction was taken to less than 45% power with control rods and recirculation pumps because of problems discovered while performing turbine control valve fast closure scram testing. A problem was corrected in the testing circuit by replacing a limit switch and the testing was successfully completed. At 0506 hours0.00586 days 0.141 hours 8.366402e-4 weeks 1.92533e-4 months , on March 23, an ascent to full power was taken and at 0810, 823 MWe was achieved. At 1507 hours0.0174 days 0.419 hours 0.00249 weeks 5.734135e-4 months , a load reduction was taken to 650 MWe at the request of the Chicago Load Dispatcher.

On March 24, at 1145 hours0.0133 days 0.318 hours 0.00189 weeks 4.356725e-4 months , power levels were adjusted and the unit was placed in EGC.

The unit remained in ECC or operated near full power until March 27.

At 1900 hours0.022 days 0.528 hours 0.00314 weeks 7.2295e-4 months , a power reduction was taken i

to 350 MWe per the Load Dispatcher. Power levels were held constant until March 28.

At 0410 hours0.00475 days 0.114 hours 6.779101e-4 weeks 1.56005e-4 months , an ascent to full power was made. At

]

0717 hours0.0083 days 0.199 hours 0.00119 weeks 2.728185e-4 months , 819 MWe was achieved.

Full load was held until March 30 j

while LPRM's were being calibrated.

On March 30, at 1710 hours0.0198 days 0.475 hours 0.00283 weeks 6.50655e-4 months , power l

levels were adjusted and the unit was placed in EGC.

The unit remained in EGC for the remainder of the month.

I I

1 I

I

r.

III.

PLANT OR PROCEDURE CHANGES, TESTS, EXPERIMENTS, AND SAFETY RELATED MAINTENANCE A.

Amendments to Facility License or Technical Specifications Technical Specification Amendment Nos. 115 and 111 were issued on February 22, 1989, to Facility Operating License DPR-29 and DPR-30 for Quad Cities Nuclear Power Station, Units 1 and 2.

These amendments delete the snubber tables in Technical Speci-fications and also, as ren.ucsted, all references to hydraulic snubbers were removed and identified typographical errors were corrected.

B.

Facility or Procedure Changes Requiring NRC Approval There were no Facility or Procedure changes requiring NRC approval for_the reporting period.

As required by Quad Cities Technical Specifications, the following information supports Revision 0 of the Quad Cities ODCM Chapters 1-11 and Appendices A-F to Technical Specification 6.8.B.

Attached is a copy of Quad Cities Ot~ Site Review package 89-9 which is required by Tech. Spec. 6.8.B.1.c.

(See Attachment A)

Also includtd is detailed information te support the change entitled, " Principal Substantive Changes" which is required by Tech. Spec.

6.8.B.'.a.

(See Attachment B)

Since this revision is a rewrite of the entire ODCM, a complete copy of this manual may be obtained upon request.

C.

Tests and Experiments Requiring NRC Approval There were no Tests or Experiments requiring NRC approval for the reporting period.

D.

Corrective Maintenance of Safety Related Equipment The following represents a tabu ar nummary of the major safety t

related maintenance performed e ts One and Two during the reporting period. This summary -.ludes the following: Work Request Numbers, Licensee Event Report Numbers, Components, Cause of Malfunctions, Results and Effects on Safe Operation, and Action Tcken to Prevent Repetition.

0027H/0061Z f

i-4-

ATTACHMENT A QAP 1400-Ti Revision 2

.QUA0 CITIES STATI0ft OflSITE September:1985 l'

REVIEW = ASSIG!4MEtti -

0 ATE 3 -? 7-9 c7 REVIEW f10.

R9-h REVIEW PARTICIPAt4TS:

bD n bl20U[

'bf l

l& k NS$

^

4 l b&\\ G

[$i ?^>W' 3/>y D ~

~N'[~ke

' ASSIG!i!iENTi Revie0 h 00cm #eVD. chages l_ g #d jpcie g g 4Nid Eftcidic4 hon 6 7 8.

8 bi APPROVED OCT :.01985

" i o.c.o.s.n.

9/032h ij

__________m.__.__.___.m_

m..

l

L l

QAP 1 6 T2 QUAD-CITIES STATION Revision 4

)

ON-SITE REVIEW REPORT-November 1987 Reference Information:

OSR Request Originator:

')

OSR No:

99-9 Station AA Off-SiteReviewLN Review Date:

J-2 7'83

' Nt.A wA Other'

'!,ricem/None,f.

j O t Request'0 ate:

dhtltQ NFS WO BWR Engineering M

Subject:

000m 4v' R, compWe awaiM of occs.

Reason for Review:

Tech. Spec. 6.1.G.2.a

/

(On-Site) sTech. Spec. 6.1.G.I.a

/

(Off-Site)

Other:

NRC Bulletin d

Station KA l

On-Site Reference Materials (attach):

Safety Evaluation uA Procedures Affected OCf 400 29 Tech Spec Pages aA FSAR Pages

~A AIR Number W4 Other DD(M Chdd5 l.Il Agindis A-F 01sposition:

W Q}Q 4 gq.694q I

/

Routine Report Off-Site Review for Concurrence (T.S. 6.1.G.2.a.(5))

/

AIR Issued (#

)

ua NRC Submittal Needed

/

Technical Specification Change uA Unreviewed Safety Question aA Other

/ M e la oot g p gt, No Further Action O

.d Other ova APPROVED 9/0322a, NOV 281987 Q.C.O.S.R.

7

~

QAP 1400-T2 Revision 4 1

L l-QUAD-CITIES STATI(*

ON-SITE REVIEW REPORT N'h OSR NO ON-SITE REVIEW

SUMMARY

h bW hl?,lh.Mck h

10 I

(O Vf fTO

$11 M

ufik y4*M

M L4 pf*/

9 ON-SITE REVIEW RECOMMENDATION:

b ch1" d

086 @

p cwcl N yui/6

/

MW /[Lza

/

A PARTICIPANTS:

Ap d:

M Stattoif Manage 7 / ~

3/77/h7 Oate:

APPROVED NOV 281987

(""* 1 )

0. c. O. S. R.

,fo322, 2_

e.

e Rev. 1 4-1-88 4

i i

1 OSR FORM - 1 l

RECORD OF REOUEST FOR OFFSITE REVIEW 1

Subject 00 W

\\1 (nnAD fhf (fb)lskf 0

00( VV} '

Station 0,iori

(,4,~ 6 Onsite R.vi.w Ro. M-09 Submitted by K

b]I/

h4A Date 1ltl f$

0 1

Test or experiment not involving an unreviewed safety question.

Proposed test or experiment involving an unreviewed safety d

question l

Proposed change to procedure, equipment or system involving an unreviewed safety question.

Proposed change to Tech. Spec. or license.

Unanticipated deficiency of design or operation of safety related structures, systems, or components.

Proposed change to GSEP.

A Referral by T. S. Supervisor Station Manager, V-P BWR

/1 Operations, V-P PWR Operations, or Manager of Quality Assurance Additional subject description:

Supporting documents attached: 052 M-1,

/p#pt [ rom so/M lln$Dl 5/lr All@b. banCf5 M/b TYk OanlP A ll) tuwmuu, fo k b/

o 9

y Date required for offsite Review completion: f4 Reason for specified date: Oh Received by Date Senior Participant Offsite Review No.

IV-16

Changes made to Chapter 10 of the ODCM rev.

O, as referenced in Section C.l.

of letter from Golden to Bax dated Feb. 28, 1989.

1. The equations for calculating alarm and trip setpoints for the airborne and liquid releases were not changed.

2. Section 10.2.3.1 Radwaste discharge monitor to read that the monitor is installed and deleted the LCD condition that was there.

3. Section 10.2.4.1 Service Water Header Releases to read that the monitor is installed and deleted the LCD condition that was there.

4. Changed Figure 10-1 Chimney Rad. Monitor Equipment,.part numbers.

5. Changed Figure 10-2 to reflect the actual system design.

6. Changed Figure 10-3 to reflect the actual system design.

7. Changed Figure 10-4 to reflect the actual system design.

i

RECEIVED February 28, 1989 N 29 Rouit y

---. _^ LAsuus.

1 f

~

derich E.D. Eenigenburg

_,,_,, _c p.,,on J.P. Joyce R. Pleniewicz R.E. Querto

Subject:

Offsite Dose Calculation Manual (ODCM), Onsite Review The ODCM has been completely rewritten to become a " readable" document for those who reference it. Attached is a summary of

-the technical changes that evolved from clarifying the' text and

'recent regulatory concerns.

Five copies of the Generic ODCM and Station Specific annexes, Revision 0, have been sent to the cognizant station person for ODCH.

The immediate initiation of the onsite approval process will be greatly appreciated.

I recommend a single onsite review meeting be held. Mary Ellen Di Ponzio'and Berline Ferguson of my staff will be in attendance to answer any questions.

To stay on schedule, I would appreciate having all onsite reviews in March, pq After completion'of the onsite review, please forward for offsite review and send a copy of the onsite review to Berline Ferguson.

For

/

additional information or clarification, you can contact Berline Ferguson on Q

extension 8154.

4N'

. John C. Golden Supervisor of Emergency Planning Attachment l

cc:

B. Ferguson M. Di Ponzio 1

R. Aker/K. Aleshire L. Aldrich/K. Klotz S. Barrett/S. Bell D. Saccomando/J. Hallace J. Sirovy/R. Hiebenga V. H1111ams/T. Sharp EPG-00-RE-0DCM

}

3234h/2 L

I

t. :o - -

Page 1 of 19 ATTACHMENT B 1!.7 l

PRINCIPAL SUBSTANTIVE CHANGE 8 -

IN THE NARCH 1989 REVISION OF THE OFF8ITE DO8E CALCULATION NANUAL l'

Prepared by:

Gerald R. Davidson Sargent & Lundy February 24, 1989 I

Pcgo 2 of 19-PRINCIPAL SUBSTANTIVE CHANGES IN THE MARCH 1989 REVISION OF THE OFF8ITE DOSE CALCULATION MANUAL j

TABLE OF CONTENTS A.

MAJOR CHANGES 5

A.1.

10 CFR 20 Organ Dose Rate Limit 5

A.2.

N-16 Skyshine Dose Calculation at BWRs 5

B.

DOCUMENTATION IMPROVEMENTS 7

B.1.

Documentation of Bases 7

B.2.

Calculation of Meteorological Dispersion Factors and Dose Factors 7

B.2.1.

Plume Depletion 7

B.2.2.

Use of Terrain Correction Factors 7

B.2.3.

Restriction of Maximum Plume Height 7

B.2.4.

Neglect of Heat Content in Determining Plume Rise 8

B.2.5.

Energy Spectra for Dose Factor Calculations 8

B.2.6.

Vertical Dispersion Parameter 8

B.2.7.

Change in Dose Factor Designation from

" Maximum Offsite" to " Site Boundary" 8

B.3.

Maximum Radioactivity in Tanks 9

B.4.

Content of Special Report on Radiological Impact on Drinking Water 9

B.S.

Calculation of Dose Due to the Uranium Fuel Cycle 9

B.5.1.

Viewing Results as Initial Estimates 9

B.S.2.

Doses from Interim Radwaste Storage Facilities 10 C.

STATION-SPECIFIC TEXT 11 l

Col.

Chapter 10 of the Revision 11 C.2.

Chapter 11 of the Revision 11 C.3.

Appendix F of the Revision 11 C.3.1.

Braicwood 11 C.3.1.1.

Aquatic Environment Dose Parameters (Table 7.2-1 of Current ODCM; Table F-1 of Revised ODCM) 11 C.3.1.2.

Station Characteristics (Table 7.2-3 of the Current ODCM; Table F-2 of the Revised ODCM) 11 C.3.1.3.

Critical Ranges (Table 7.2-4 of the Current ODCM; Table F-3 of the Revised ODCM) 12 C.3.1.4.

Terrain Correction Factors (Table 7.2-5 of the Current ODCM) 12 i

Paga 3 of 19

.C.3.2.

Byron 12 C.3.2.1.

Aquatic Environment Dose Parameters (Table.7.2-1 of Current ODCM; Table F-1 of Revised ODCM) 12

.C.3.2.2.

Station Characteristics (Table 7.2-3 of the Current ODCM; Table F-2 of the Revised ODCM) 12 C.3.2.3.

Critical Ranges 1

(Table 7.2-4 of the Current ODCM; Table F-3 of the Revised ODCM) 12 C.3.2.4.

Terrain Correction Factors (Table 7.2-5 of the Current ODCM) 12 C.3.3.

Dresden 13 C.3.3.1.

Station Characteristics (Table 7.2-3 of the Current ODCM; Table F-2 of the Revised ODCM) 13 C.3.3.2.

Critical Ranges (Table 7.2-4 of-the Current ODCM; Table F-3 of the Revised ODCM) 13 C.3.3.3.

Terrain Correction Factors (Table 7.2-5 of the current ODCM) 13 C.3.3.4.

D/Q at the Nearest Milk Cow and Meat Animal Locations Within 5 Miles (Tables 7.2-7 and 7.2-10 of the current ODCM; Table F-6 of the Revised ODCM) 13 C.3.4.

La Salle 14 C.3.4.1.

Station Characteristics j

(Table 7.2-3 of the Current ODCM; Table F-2 of the Revised ODCM) 14 C.3.4.2.

Critical Ranges (Table 7.2-4 of the Current ODCM; Table F-3 of the Revised ODCM) 14 C.3.4.3.

Terrain Correction Factors (Table 7.2-5 of the Current ODCM) 15 C.3.5.

Quad Cities 15 C.3.5.1.

Aquatic Environment Dose Parameters (Table 7.2-1 of current ODCM; Table F-1 of Revised ODCM) 15 g

C.3.5.-

Station Characteristics I

(Table 7.2-3 of the Current ODCM; Table F-2 of the Revised ODCM) 15 C.3.5.3.

Critical Ranges l

(Table 7.2-4 of the Current ODCM; i

Table F-3 of the Revised ODCM) 15 1

4

~ _ - - - - _... - - - -... - -.. - _ - _ _. -. _ -. _ _. _. - - _

'Page 4 of'19 C.3.5~.4.

Terrain Correction Factors (Table'7.2-5 of the Current ODCM) 16 C.3.5.5 LD/Q at=the Nearest Milk Cow and Meat Animal Locations Within 5 Miles (Table 7.2-7 of the Current ODCM Table F-6 of the Revised ODCM) 16 p

C.3.6.

Zion 17 l

C.3.6.1.

Aquatic Environment Dose Parameters (Table 7.2-1 of Current ODCM; L

Table F-1 of Revised ODCM) 17 i

C.3.6.2.

Station Characteristics (Table 7.2-3 of the Current ODCM; Table F-2 of the Revised ODCM) 17 C.3.6.3.

' Critical Ranges (Table 7.2-4 of the Current ODCM; I

Table.F-3 of the Revised ODCM).

17 C.3.6.4.

Terrain Correction Factors

' (Table 7.2-5 of the Current ODCM) 17 D.

REFERENCES 19 i

e 4

wmm _ _ _

.._m..m.__

Page 5 of 19

~

PRINCIPAL SUBSTANTIVE CEANGE8 I

IN THE MARCH 1989 REVISION

)

OF THE OFFSITE DOSE CALCULATION MANUAL l

A major revision of the Commonwealth Edison Company Offsite Dose j

Calculation Manual (ODCM) is scheduled to be issued'in March 1989.

It will be designated Revision O.

Most of the changes from the current ODCM are editorial in nature.

The principal

-substantive changes are discussed in this document.

{

A.

MAJOR CEANGES l

A.1.

10 CFR 20 Organ Dose Rate Limit

)

For three stations (Dresden, La Salle, and Quad Cities), the child thyroid has been substituted for the irfant thyroid for purposes of evaluating compliance with the 10 CFR 20 limit of 1500 mrem /yr on organ dose rate.

Compare Section A.1.5 of the revision with Section 2.1.2.2 of the current ODCM.

This change brings the ODCM into compliance with revisions in Technical Specifications at Dresden, La Salle, and Quad Cities that are ex-pected to be completed by March 1989.

Note that Braidwood, Byron, and Zion already use the child thyroid for this evalua-tion.

The change was requested by the NRC (References 1 - 3) because it j

is the present understanding of the NRC staff that the calculated dose rate to a child thyroid is larger than the calculated dose rate to an infant thyroid.

This arises from the use of the cur-rently accepted values of breathing rate and dose factor for a child and an infant.

Therefore, comparison of child thyroid dose rate with the organ dose rate limit is more conservative than comparison of infant thyroid dose rate with the limit.

A.2.

N-16 Skyshine Dose Calculation at BWRs The formula and procedures for calculating offsite dose due to nitrogen-16 skyshine from turbines and turbine piping at boiling water reactors have been modified (compare Section A.3.2 and

. Equation A-35 of the revision with Section 2.3.1.2.5 and Equation 2.38 of the current ODCM).

The foll,owing changes have been made:

a.

An enhancement factor of 5 has been added for use during periods when power is being generated with hydrogen ad-dition to primary coolant.

This factor is denoted by the symbol Mh in Equation A-35 of Appendix A of the re-vised ODCM.

At a given power level, Mh is the ratio be-tween calculated offsite dose rate with hydrogen addi-tion and calculated offsite dose rate without hydrogen addition.

The value of 5 for M is based on measure-b ments made at Dresden (see Section E.11.1 and Reference 73 of the revised ODCM).

l

[

Paga 6 of 19' E

b.

In.the current ODCM (see Section 2.3.1.2.5), the

<skyshine formula is considered applicable only.at distances of.up to 1100 meters.

No guidance is provided on what value to take for skyshine dose at greater

' distances.

In.the revised ODCM, the formula is considered applicable at all distances.

The region of applicability has been extended-in the revised ODCM in order to enable skyshine dose estimates at distances greater than or equal to 1100 meters'in case such estimates should be needed.

The definitions of the energy parameters (E c.

E) in the h

O, formula (Equation A-35 of'the revision) have been modi-fled to stipulate use of cross energy generated in MW,-

Previously, it was not stated whether to use gross hrs.

or not energy.

This change has been made so stations will not have to guess about whether to use' net or gross energy.

Although the difference is not large, the use of gross.rather than net energy is conservative.

d.

The location and occupancy parameters for calculating skyshine dose have been modified for Quad Cities (compare Table F-8 of' Quad Cities Appendix F of the.re-vision with Equation 2.40 of the current ODCM).

. Location and occupancy parameters for calculating e.

skyshine dose have been provided for the first time-for La Salle (see Table F-8 of La Salle Appendix F of the revision).

f.-

The: dose obtained by use'of the skyshine formula (Equation A-35) is characterized in the text of the re-vised ODCM as an " initial estimate."

This will allow the station to calculate a more realistic value should such a step be desirable.

See Section A.3.2, Section E.11.1, and Table F-8 of the revised ODCM for details and discussion of these changes.

1

]

Page 7 of 19

{

p

'B.

_ DOCUMENTATION. IMPROVEMENTS In'the' revision, various improvements in the documentation'of the

'ODCM methodology have been made.

These include the following:

1 Correction of inaccuracies and inconsistencies in the j

o current text.

./

Elimination of sections and equations not needed because c

there is no call for their use in current station Technical Specifications.

~The principal documentation changes are discussed below.

.B.1.

Documentation of Bases The bases of-some of the material in the current ODCM are not stated.. Appendix E of the revision provides supplementary docu-mentation of bases.

L B.2.

Calculation of Meteorological Dispersion Factors and Dose Factors B.2.1.

Plume Depletion The current ODCM is somewhat unclear on whether or not plume de-plation was included in determining the tabulated station-spe-cific values.of X/Q (see Sections 3.2.3.2 and 3.4 of the current ODCM).

In. fact, no credit was taken for plume depletion in cal-culating X/Q.

This is stated clearly in section B.3.4 of the-re-vision and also is. discussed in Section E.9.6 of the revision.

The quantitative significance of neglecting plume depletion is

' discussed in Reference 4.

-B.2.2.

Use of Terrain Correction Factors The current ODCM states that terrain correction factors were used in calculating station-specific meteorological and dose factors

-(see Section 3.2.2.1 ano Table 7.2-5).

In fact, terrain correc-tion factors were not used.

This is. stated in Section B.3.1.1.2 of the revision.

The quantitative significance of not correcting l

for terrain height is discussed in Reference 4.

l B.2.3.

Restriction of Maximum Plume Reight In the calculation of the station-specific meteorological and l-dose factors which are in both the current and revised versions of the ODCM, the maximum plume height was restricted to 100 me-ters, a conservative approximation.

This restriction is not noted in the current ODCM, but it is stated in the revision (see Equation B-15 of the revision).

Pago 8 of 19

(

B.2.4.

Neglect of Heat Content in Determining Plume Rise The current ODCM provides nonzero values of heat content for re-leases from the elevated (stack) release points at Dresden, Quad Cities and La Salle (see Table 7.2-3 of the current ODCM).

This might be interpreted as implying that credit was taken for heat

{

content in determining plume rise.

In fact, no credit was taken for heat content. In the revision, plume heat content is stated

{

l to be zero for these three stations (see Table F-2 of the revi-l sion).

l j

B.2.5.

Energy Spectra for Dose Factor Calculations The description in the current ODCM of the methodology for calcu-lating station-specific dose factors is inaccurate in two re-spects.

First, in the formulas of Section 3.3.1.1 of the current ODCM, each radionuclides is assumed to emit moneenergetic gamma radiation with an energy equal to the average gamma ray energy per disintegration.

Second, corrections for tissue absorption are made using a tissue energy absorption coefficient M i appro-t priate to this average energy (see Section 3.3.1.2 of the current ODCM).

A corrected description of the methodology is provided in Sections B.5 and B.6 of the revised ODCM.

The corrected description states that the dose fcctors were calculated by using the photon energy emission spectra of the radionuclides rather than their average energies.

The tabulations in the current ODCM of average gamma emission en-ergies and associated tissue energy absorption coefficients (see Table 7.1-9) are omitted from the revision because the data are not used.

B.2.6.

Vertical Dispersion Parameter

'In the current ODCM, Table 7.1-7 provides equations and parame-ters for obtaining the vertical dispersion parameter o used to g

calculate meteorological and dose factors.

In the review of the ODCM, the information in this table was found not to describe the procedure actually used to obtain the values of a for certain s

stability classes and distance ranges.

Correct information is provided in Table D-10 of the revision.

B.2.7.

Change in Dose Factor Designation from " Maximum Offsite" to " Site Boundary" In the dose factor table of the current ODCM (Table 7.2-8), the dose factors are described as " maximum" by the table title,

" Maximum Offsite Finite Plume Gamma Dose Factors...."

In fact, all of the dose factors were calculated at the site boundary.

The site boundary dose factors are very good approximations to the maximum offsite dose factors, but they are not necessarily equal to the maximum dose factors.

This is made explicit in the

Page 9 of 19 revision by modifying the title to " Site Boundary Finite Plume Dose Factors...."

This issue is discussed in Section E.9.8.3 of the revision.

B.3.

Maximum Radioactivity in Tanks The current ODCM provides formulas and parameters for determining

'the maximum permissible concentration of radioactivity in an un-protected outdoor tank (see Section 2.2.3 and the lower half of.

Table 7.2-1 of the current ODCM).

At present, the Technical Specifications of four CECO stations have specifications re-stricting the radioactivity in such tanks'.

For three of the sta-tions (Braidwood, Byron, and Zion), the restriction is stated in the Technical-Specifications as a specific Curie limit (e.g.,

10 Curies).

The Technical Specifications for La Salle specify a limit " calculated in the ODCM."

In the current ODCM, the limit for La Salle~is specified as 10 Curies (see Table 7.2-1 of the current ODCM).

Since the current ODCM provides a numerical limit on tank activ-ity for La Salle and since the Technical Specifications of the other affected CECO stations provide numerical limits for those stations, there is-no need for the outdoor tank concentration formulas and parameters in Section 2.2.3 and Table 7.2-1 of the current ODCM.

Therefore, these formulas and parameters have been omitted in the revision (see Section A.2.4 of the revision).

For each station, the applicable Curie limits have been specified'in Table F-1 of the revision.

B.4.

Content of Special Report on Radiological Impact on Drinking Water Under certain special circumstances, a station is required by'its Technical Specifications to prepare a special report on the radi-l ological impact of radioactivity in its liquid affluents on the i

nearest community drinking water supply (see section A.4 of the revision).

Section A.4.3 of the revision spells out the content of this special report in somewhat greater detail than it is spelled out in Section 5.4 of the current ODCM.

Section A.4.3 of the revision calls for the station to provide information on amounts of radioactivity released, flow rates, and dilution val-ues.

The current ODCM does not call for providing this informa-tion in the report.

B.S.

Calculation of r,ose Due to the Uranium Fuel Cycle B.5.1.

Viewing Results as Initial Estimates Procedures for calculating dose due to the uranium fuel cycle for purposes of assessment of compliance with 40 CFR 190 are described in Section 2.3 of the current ODCM and in Section A.3 of the revised ODCM.

The procedures are in general similar., but the discussion in the revised ODCM emphasizes that the calculated dose is viewed as an initial estimate of the dose to the

Page 10 of 19 maximally exposed member of the public.

This allows re-evaluation of the assumptions in the calculations for realism should this be desirable.

B.5.2.

Doses from Interim Radwaste Storage Facilities Radiation from Interim Radwaste Storage Facilities should be taken into account in assessing compliance with the uranium fuel cycle dose limits of 40 CFR 190 (see Section A.3 of the revi-sion).

These facilities are discussed in Section E.3.3 of the revision.

The radiation dose from these facilities is judged to be negligible in comparison with 40 CFR 190 limits.

This is done on the grounds of design basis calculations which are based on the assumption that radwaste containers have a contact dose rate of 5 R/hr.

i i

I l

l

Paga 11 of 19 C.

STATION-SPECIFIC TEXT C.1.

Chapter 10 of the Revision Chapter 10 of the revised ODCM contains the station-specific ma-terial~that is in Sections 0.1, 8.2, and 8.3 of Chapter 8 of the current ODCM.

There have been some substantive changes in this material based on comments by station personnel.

The material has also been edited for style.

C.2.

Chapter 11 of the Revision Chapter 11 of the revised ODCM contains the station-specific ma-terial that is in Section 8.4 of Chapter 8 of the current ODCM.

The changes in this material hhve concerned only format, riot sub-stance.

C.3.

Appendix F of the Revision Appendix F of the revised ODCM contains the station-specific ma-terial that is in Section 7.2 of Chapter 7 of the current ODCM.

The changes in this material have concerned only format, not sub-stance, except as follows:

C.3.1.

Praidwood C.3.1.1.

Aquatic Environment Dose Parameters (Table 7.2-1 of current ODCM; Table F-1 of Revised ODCM)

The value of t", the time of travel from the station liquid a.

discharge to the nearest community water suppif, is 120 hours0.00139 days 0.0333 hours 1.984127e-4 weeks 4.566e-5 months in the current ODCM and 121 hours0.0014 days 0.0336 hours 2.000661e-4 weeks 4.60405e-5 months in the revised 0DCM.

The change results from revising the distance between the discharge and the water supply from 120 miles to 121 miles.

The value 121 miles is provided in Table E-2 of the revised ODCM and is based on data in the Braidwood Environmental Report.

b.

In the revised ODCM, the table states numerical limits (in Curies) on radioactivity in unprotected outdoor tanks.

The val-ues are from.the station Technical Specifications.

The current ODCM states that there are no unprotected outdoor tanks.

C.3.1.2.

Station Characteristics (Table 7.2-3 of the current ODCM; Table F-2 of the Revised ODCM)

The value of the Building Factor D is stated as 60.5 m in the current ODCM and as 51.96 m in the revised ODCM.

The value pro-vided for D in the revised ODCM is the value actually used in calculating the meteorological and dose factors that are tabu-lated in Section 7.2 of the current ODCM and in Appendix F of the revised ODCM.

Page 12 of 19 I

(

C.3.1.3.

Critical Ranges (Table 7.2-4 of the current ODCM; Table F-3 of the Revised ODCM)

)

I In Table 7.2-4, the distances to the nearest resident are speci-i fled to the nearest meter, whereas in Table F-3, the distances l

are rounded to the nearest 100 meters.

The rounding in Table F-3 l

more accurately reflects the precision of the 1988 nearest resident census data (distances are reported to the nearest 0.1 mile).

C.3.1.4.

Terrain Corgschion Factors (Tabla 7.2-5 of the Current ODCM)

This table is not included in the revised ODCM because a value of

)

zero was assumed for all terrain correction factors in calculat-1 ing the meteorological and dose factors tabulated in the ODCM (see Item B.2 of this document).

C.3.2.

Byron C.3.2.1.

Aquatic Environment Dose Parameters (Table 7.2-1 of current ODCM; Table F-1 of Revised ODCM)

In the revised ODCM, the table states numerical limits (in Curies) on radioactivity in unprotected outdoor tanks.

The val-4 ues are from the station Technical Specifications.

The carrent ODCM states that there are no unprotected outdoor tanks.

C.3.2.2.

Station Characteristics (Table 7.2-3 of the current ODCM; Table F-2 of the Revised ODCM)

The value of the Building Factor D is stated as 60.6 m in the current ODCM and as 51.96 m in the revised ODCM.

The value pro-vided for D in the revised ODCM is the value actually used in calculating the meteorological and dose factors that are tabu-lated in Section 7.2 of the current ODCM and in Appendix F of the revised ODCM.

C.3.2.3.

Critical Ranges (Table 7.2-4 of the current ODCM; Table F-3 of the Revised ODCM)

The distances to the nearest resident in Table F-3 differ from those in Table 7.2-4.

The distances in Table F-3 are based on the 1988 nearest resident census.

C.3.2.4.

Terrain Correction Factors (Table 7.2-5 of the Current ODCM)

This table is not included in the revised ODCM becatise a value of zero was assumed for all terrain correction facte a in calculat-

Page 13 of 19 ing the meteorological and dose factors tabulated in the ODCM (see Item B.2 of this document).

C.3.3.

Dresden C.3.3 1.

Station Characteristics (Table 7.2-3 of the Current ODCM; Table F-2 of the Revised ODCM)

I The values of several of the parameters in this table are differ-ent in the current and revised versions of the ODCM.

The values in the revised version are those which were actually used in cal-culating the meteorological and dose factors that are tabulated in Section 7.2 of the current ODCM and in Appendix F of the re-vised ODCM.

C.3.3.2.

Critical Ranges (Table 7.2-4 of the Current ODCM; Table F-3 of the Revised ODCM)

The distances to the nearest resident in Table F-3 differ from those in Table 7.2-4.

The distances in Table F-3 are based on the 1988 nearest resident census, which identified a closer resident in the ENE sector.

Also, the distances in Table F-3 are rounded to the nearest 100 meters instead of specified to the nearest meter.

The rounding in Table F-3 more accurately re-flects the precision of the 1988 nearest resident census data (distances are reported to the nearest 0.1 mile).

Table 7.2-4 states that there is one sector having a dairy farm within 5 miles, the NNE sector with a farm at 8000 meters.

Table F-3 states that there are no dairy farms within 5 miles.

This is based on the 1988 milch animal census.

C.3.3.3.

Terrain Correction Factors (Table 7.2-5 of the current ODCM)

This table is not included in the revised ODCM because a value of zero was assumed for all terrain correction factors in calculat-ing the meteorological and dose factors tabulated in the ODCM (see Item B.2 of this document).

C.3.3.4.

D/Q at the Nearest Milk Cow and Meat Animal Locations Within 5 Miles (Tables 7.2-7 and 7.2-10 of the current ODCM; Table F-6 of the Revised ODCM)

In Table F-6, the distance to the nearest milk cow has been re-vised for one sector, as follows:

['

y-Paga 14 of'19

. Distance (meters)

Former-

.New Sector Value value_.

NNW-l 7725 8047 The D/Q valuer in the table have been adjusted accordingly:

)

i D/O F1/(meters)2y Release.

Former New Classification Value Value Elevated 9.445E-11 8.854E-11 Mixed Mode-1.070E-11 1.000E-10 Ground 2.855E-10 2.663E-10 The new values were obtained from data in the " nearest meat ani-Emal" section:of.the table.

The' change'was made to bring the nearest milk cow distances-in Table F-6 into agreement with those in Table F-3 And with results of the'1988 census of nearest milch animals, which found no dairy farms within a-5-mile-radius.cf Dresden.

C.3.4.

La Salle C.3.4.1.

Station Characteristics j

(Table 7.2-3 of the current.ODCM; Table F-2 of the~ Revised ODCM)

The values of several of the parameters in this table are differ-ent in the current and' revised versions of.the ODCM.

The values in the revised version are those which were actually used in cal-culating the meteorological and dose factors that are tabulated in Section 7.2 of the current ODCM and in Appendix F of the re-vised ODCM.

C.3.4.2.

Critical Ranges (Table 7.2-4 of the current ODCM; Table F-3 of the Revised ODCM)

The distances to the nearest resident in Table F-3 differ from those in Table 7.2-4.

The distances in Table F-3 are based on i

the 1988 nearest resident census, which identified a closer resident in the NNW sector.

Also, the distances in Table F-3 are rounded to the nearest 100 meters'instead of specified to the nearest meter.

The rounding in Table F-3 more accurately re-flects the precision of the 1988 nearest resident census data (distances are reported to the nearest 0.1 mile).

Paga 15 of 19 K

C.3.4.3.

Terrain Correction Factors (Table 7.2-5'of the Current ODCM)

This table.iis not included in the revised ODCM'because a value of zero was assumed for all terrain correction factors in calculat--

ing-the meteorological and dose factors tabulated in the ODCM (see--Item-B.2 of this document).

C.3.5.

Quad Cities C.3.5.1.

Aquatic Environment Dose Parameters

.(Table 7.2-1 of Current ODCM; Table F-1 of Revised ODCM)

-The value of t',

the time of travel from the station liquid dis-charge to the. nearest community water supply, is 7.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months in the current ODCM and 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months in the-revised ODCM.

The change results from revising the distance between the discharge and the water supply from 15 miles to 16 miles.

The value 16 miles is provided in Table E-2 of the revised ODCM and is based on information ob-tained by CECO's B.

S. Ferguson from the Army Corps of Engineers in January 1989.

C.3.5.2.

station Characteristics (Table 7.2-3 of the Current ODCM; Table F-2 of the Revised ODCM)

The values of several of the parameters in this table are differ-ent in the current and revised versions of the ODCM.

The values in the revised version are those which were actually used in cal-culating the meteorological and dcas factors that are tabulated in Section 7.2 of the current ODCM and in Appendix F of the re-vised ODCM.

C.3.5.3.

Critical Ranges (Table 7.2-4 of the Current ODCM; Table F-3 of the Revised ODCM)

In Table 7.2-4, the distances to the nearest resident are speci-fled to the nearest meter, while others appear to have been rounded off to the nearest 100 meters.

In Table F-3, all of the nearest resident distances are rounded to the nearest 100 meters.

The rounding in Table F-3 more accurately reflects the precision

~

of the 1988 nearest resident census data (distances are reported to the nearest 0.1 mile).

The results in Table F-3, which are based on the 1988 nearest resident census data, indicate that the nearest residents in the ENE and SW sectors are 100 meters closer to the station than stated in Table 7.2-4.

?.

Paga 16 of 19 C.3.5.4.

Terrain Correction Factors (Table 7.2-5 of the current ODCM)

This table is not included in the revised ODCM because a value of zero was assumed for all terrain correction factors in calculat-ing the meteorological and dose factors tabulated in the ODCM (see Item B.2 of this document).

C.3.5.5 D/Q at the Nearest Milk Cow and Meat Animal Locations Within 5 Miles (Table 7.2-7 of the current ODCM Table F-6 of the Revised ODCM)

In Table F-6, the distance to the nearest milk cow has been re-vised for two sectors, as follows:

Distance (meters)

Former New Sector Value Value

. East 3219 8047 South 1694 8047 The D/Q values in the - table have been adjusted accordingly based on data taken from Reference 1 of Quad Cities Appendix F of the revised ODCM:

East Sector D/O f1/(meters)21 Release Former New Classification Value Value Elevated 3.212E-10 8.744E-11 Mixad Mode 7.066E-10 1.497E-10 Ground 1.055E-09 2.144E-10 South Sector D/O f1/(meters)2y Release Former New Classification Value Value Elevated 5.681E-10 7.839E-11 Mixed Mode 1.348E-09 1.106E-10 Ground 2.783E-09 1.826E-10 The change was made to bring the nearest milk cow distances in Table F-6 into agreement with those in Table F-3 and with results of the 1988 census of nearest milch animals, which found no dairy farms within a 5-mile radius of Quad Cities.

l Page 17 of 19 C.3.6.

Bion C.3.6.1.

Aquatic Environment Dose Parameters (Table 7.2-1 of current ODCM; Table F-1 of Revised ODCM)

Table F-1 of the revised ODCM provides a numerical limit (in Curies) on radioactivity in an outside Temporary Radioactive Liquid Storage Tank.

The limit is the value in the station Technical Specifications.

In contrast, Table 7.2-1 of the cur-rent ODCM provides data for calculating the limits on radioactiv-ity in the Primary Water Storage Tank and the Secondary Water Storage Tank.

This information has been omitted from Table F-1 because the station Technical Specifications do not limit the ra-bioactivity in these two storage tanks (see Technical Specification 3.11.5).

C.3.6.2.

station Characteristics (Table 7.2-3 of the Current ODCM; Table F-2 of the Revised ODCM)

The values of several of the parameters in this table are differ-ent in the current and revised versions of the ODCM.

The values in the revised version are those which were actually used in cal-cular.ing the meteorological and dose factors that are tabulated in Section 7.2 of the current ODCM and in Appendix F of the re-vised ODCM.

1 C.3.6.3.

Critical Ranges (Table 7.2-4 of the current ODCM; Table F-3 of the Revised ODCM)

The distances to the nearest resident in Table F-3 differ from those in Table 7.2-4.

The distances in Table F-3 are based on the 1988 nearest resident census.

Also, the distances in Table F-3 are rounded to the nearest 100 meters instead of specified to the nearest meter.

The rounding in Table F-3 more accurately re-flects the precision of the 1988 nearest resident census data (distances are reported to the nearest 0.1 mile).

C.3.6.4.

Terrain correction Factors (Table 7.2-5 of the current ODCM)

This table is not included in the revised ODCM because a value of zero was assumed for all terrain correction factors in calculat-ing the meteorological.and dose factors tabulated in the ODCM (see Item B.2 of this document).

C.3.6.5 D/Q at the Nearest Milk Cow and Meat Animal Locations Within 5 Miles In Table F-6, the distance to the nearest milk cow has been re-vised for one sector, as follows:

i-

.Paga 18 of 19 Distance (meters)

Former New Sector Value Value North 6437 8047 The D/Q values in the table have been adjusted accordingly:

D/O f1/(meters)2J.

Release Former New Classification Value Value Elevated 6.698E-11 4.634E-11 Mixed Mode 1.827E-10 1.246E-10 Ground 4.820E-10 3.26 2-10 The new values were obtained from data in the " nearest meat ani-mal" section of the table.

The change was made to bring the nearest milk cow distances in Table F-6 into agreement with those in Table F-3 and with results of the 1988 census of nearest milch animals, which found no dairy farms within a 5-mile radius of Zion.

1

Page 19 of 19 D.

REFERENCES 1.

" Acceptance of the offsite Dose calculation Manual, Revisions 11 and 11A, Updated to April 30, 1987 (TACS 62097 and 62099)," letter from D. R. Muller (NRC) to D.

L.

Farrar (CECO) related to Dresden Units 2 and 3, June 25, 1987.

2.

" Acceptance of the Offsite Dose Calculation Manual Updated Through Revision 12 for La Salle County Station, Units 1 and 2," letter from E. G. Adensam (NRC) to D. L.

Farrar (Ceco),

October 21, 1986.

3.

" Quad Cities Nuclear Power Station, Units 1 and 2 - Offsite Dose Calculation Manual, Revision 11 (TAC 62101/62103)," let-ter from T. Ross (NRC) to L.

D.

Butterfield, Jr. (Ceco), July 27, 1987.

4.

" Discrepancies in the offsite Dose Calculation Manual (ODCM)," letter from G. R. Davidson (Sargent & Lundy) to M.

E. DiPonzio (CECO), October 10, 1988.

e

UNIT 1 MAINTENANCE

SUMMARY

WORK REQUEST No.: Q63017 LER NUMBER:

87-026 COMPONENT: System 1600 - Piping support identified on M-1603-06/c is to be relocated for FSAR compliance.

Reference drawing M-992A, Line 1-1318-6-LX.

CAUSE OF MALFUNCTION: Unit I was in RUN mode at 15% thermal power. The station was notified by BWRED that piping supports on drawing number M-1603-05 and M-1603-06 for RCIC pump suction line 1-1318-6"-LX did not comply with FSAR criteria for allowable stress. The apparent cause of the event was design error involving A/E personnel.

RESULTS & EFFECTS ON SAFE OPERATION: The safety of the plant and personnel l

were not affected during this event.

Analysis of the model has shown system-operability even though FSAR criteria was not met.

ACTION TAKEN TO PREVENT REPETITION: Corrective action was to remove the incorrect support to comply with FSAR criteria.

WORK REQUEST NO.:

Q63909 LER NUMBER:

87-03 COMPONENT:

System 2300 - Inspected flow controller for loose solder joints.

Found that all joints looked good.

Generic problem found on NPRDS.

CAUSE OF MALFUNCTION: Unit 1 was in RUN mode at 100% thermal power. While performing QOS 1300-52, it was discovered that the flow controller did not respond to automatic flow control signals.

The system worked in MANUAL, but RCIC was declared inoperable because it was unclear whether the flow controllers would work properly in the event of an auto-start.

NRC was notified.

The cause of the f ailure was determined to be a loose cold solder joint in the setpoint tape chassis of the electronic controller. A replacement controller was installed.

The original controller was repaired and re-installed.

RCIC was verified operable during Unit 1 startup.

RESULTS & EFFECTS ON SAFE OPERATION: The safety of the public and plant personnel was never affected during this event since HPCI was proven operable per Tech. Spec. 3.5.E.2.

ACTION TAKEN TO PREVENT REPETITION: Work Request Q55263 was written to repair the RCIC controller. Work Requests Q63908, Q63910, and Q63911 were written to inspect, and repair, if necessary, the remaining flow controllers on the RCIC and HPCI systems.

o.

WORK REQUEST NO.:

Q66762 LER NUMBER:

NA COMPONENT:

System 203 - The acoustic monitor for relief valve 1-203-3A did not function when valve was opened. Tests were done and monitor was fixed.

CAUSE OF MALFUNCTION: Unit I was in RUN mode at 18% thermal power. Procedure QOS 201-1 was in progress.

When valve 1-203-3A was opened for surveillance, the acoustic monitor failed to indicate an opening of the valve. Valve operation was verified by exhaust temperature increase and turbine bypass valve closure.

The thermocouple for the valve functioned when the valve was cpened. This justified continued operation of Unit 1.

A poor connection on the main control room chassis connector of the acoustic monitor power module was at fault.

RESULTS & EFFECTS ON SAFE OPERATION: Plant safety was not affected because the acoustic monitor can be inoperable as long as the thermocouple is operable.

ACTION TAKEN TO PREVENT REPETITION: Work Request Q66762 was initiated to investigate and repair the problem with the monitor. The Instrument Maintenance Department cleaned the control chassis back plain connector and checked the local pre-amp terminals for tightness.

To alleviate problems with edge connectors in the acoustic monitor power modules, they are going to be replaced with new and rebuilt modules.

WORK REQUEST NO.:

Q67029, Q67045, Q67080 LER NUMBER:

NA COMPONENT:

System 6600 - The 1/2 diesel tripped on high crankcase pressure.

Tests revealed no problem.

A baffle was installed in 1/2 Emergency Diesel Generator to prevent spurious trips of engine due to oil splashing on pressure switch.

Engine crankcase pressure detector baffle plate was fabricated.

EPN 1/2-6641-CPS CAUSE OF MALFUNCTION: Unit I was in RUN mode at 93% thermal power and Unit 2 was in REFUEL mode. While performing operability test, the 1/2 Emergency Diesel Generator tripped on high crankcase pressure after starting. Another start attempt was made and no trips occurred.

It is suspected that the false trips were caused by oil splashing on the pressure switch diaphragm.

RESULTS & EFFECTS ON SAFE OPERATION: The safety implications of the event are minimal since the required surveillance specified by Tech. Specs. 4.9.E.1 were performed. The Unit 1 Emergency Diesel Generator and Division II low pressure core cooling systems were demonstrated fully operabic, and two offsite power sources were available. The 1/2 Emergency Diesel Generator crankcase pressure switch is bypassed on an automatic initiation and the 1/2 Emergency Diesel Generator was available to fulfill its design function.

ACTION TAKEN TO PREVENT REPETITION: An oil deflector assembly, recommended by the manufacturer, was installed between the crankcase pressure switch and the accessory gear train housing.

Similar changes were made for the Unit 1 and Unit 2 Emergency Diesel Generators per Work Request Q67145 and Q67146. Work Request Q67045 was initiated to perform the corrective work on the 1/2 Diesel.

1

'e WORK REQUEST NO.: Q67996, Q67997 LER NUMBER: 87-026 COMPONENT:

System 2300 - Piping support on lines 1-2325-6" and 1-2302-16" do not meet FSAR compliance, also LER compliance.

CAUSE OF MALFUNCTION: BWR Engineering notified Quad Cities Station that five piping supports had failed to meet FSAR compliance. They were supports M-1604-03, M-1604-09 and M-1604-10 for the HPCI pump suction line 1-2325-6",

and M-1604-05 and M-1604-13 for HPCI pump suction line 1-2302-16".

The apparent cause of the event was design error on the part of A/E personnel.

RESULTS & EFFECTS ON SAFE OPERATION:

The safety of the plant and personnel were not affected during this event. Analysis of the-model has shown system operability even though FSAR criteria was not met.

ACTION TAKEN TO PREVENT REPETITION:

Corrective action was to remove the incorrect support to comply with FSAR criteria. Work Requests Q67996 and Q67997 were written to remove the supports.

WORK REQUEST No.: Q68577 LER NUMBER: NA COMPONENT:

System 0700 - Number 2 tip machine ball valve was stuck open.

Installed new valve - old valve failed leak test.

CAUSE OF MALFUNCTION: Unit One was in RUN mode at 100% thermal power. After completion of a Local Power Range Monitor System, using the Traversing Incore Probe, the ball valve on machine two would not close after the TIP was retracted into its shield. The ball valve was closed from the control room. The cause of the failure was a weak closure spring and/or crud buildup in the ball i

valve internals.

RESULTS & EFFECTS ON SAFE OPERATION:

The safety of the plant and personnel were not jeopardized because the ball valve was closed in eight minutes.

Therefore, containment isolation was.not jeopardized. This event was not considered to be potentially significant.

ACTION TAKEN TO PREVENT REPETITION: The ball valve was removed under Work Request Q68577. Work Request Q68693 was written to rebufid the ball valve.

The valve was replaced like-for-like.

The remaining TIP ball valves were to be rebuilt in an outage of sufficient duration.

c=

n-~-w__--__-__.__._____

c, WORK' REQUEST No.: Q68613 LER NUMBER: NA COMPONENT:

System 1700 - Repaired and cycle calibrated ID MSL Rad Monitor S/N 304A3700G2 and reinstalled.

CAUSE OF MALFUNCTION: Unit I was in RUN mode at 92% thermal power. A half

. scram and a half Group I isolation actuation occurred.

It was found that the ID main steamline radiation monitor displayed several failure messages.

The monitor was replaced under Work Request Q68612. The half scram and isolation were reset. The suspected cause is a poor contact between an IC and its socket.

RESULTS & EFFECTS ON SAFE OPERATION:

The safety consequences were considered minimal because the three other monitors were operable and the monitor failed in the conservative direction.

ACTION TAKEN TO PREVENT REPETITION:

Immediate corrective action was to replace the MSL radiation monitor chassis with a spare. During troubleshooting, under Work Request Q68613, the monitor started performing properly and would not fail. This failure was considered to be an isolated event.

' WORK REQUEST NO.:

Q68634 LER NUMBER: NA COMPONENT:

System 1600 - Unit 1 torus temperature RCDR "B" read incorrect.

Recorder was cleaned and calibrated.

CAUSE OF MALFUNCTION: The Unit 1 NSO on panel checks discovered the 1-1602-8 torus recorder "B" channel reading 100*F and "A" channel reading 920*F.

Channel "B" was declared inoperable.

The cause of the problem was found to be a dirty connector on the back of the recorder.

RESULTS & EFFECTS ON SAFE OPERATION: The safety consequences were considered minimal since the problem was found to be a connection problem and not an instrument failure.

ACTION TAKEN TO PREVENT REPETITION: The connectors on the back of the recorder were cleaned and then both channels read properly.

l J

e 4

WORK REQUEST No.: Q68920 LER NUMBER:

88-013 COMPONENT: System 5700 - One belt on lA core spray room cooler was broken.

Belt was replaced.

CAUSE OF MALFUNCTION: Unit 1 was in RUN mode et 60% thermal power. The unit EA found the Unit 1 "A" core spray room cooler to be off, and it was unable to be started. The 1A core spray system and Unit 1 RCIC system were declared inoperable.

It was found that one cooler belt was broken and the other was off the pulley. NRC notification was completed. Mechanical Maintenance replaced the V-belts and the systems were declared operable.

The cause was determined to be insufficient preventative maintenance.

RESULTS & EFFECTS ON SAFE OPERATION: The safety of the plant was not affected during this event because Unit 1 HPCI and the LPCI mode of RHR were successfully tested after the 1A core spray was declared inoperable. Therefore, all ECC systems including the "B" loop at core spray were operable throughout the event.

ACTION TAKEN TO PREVENT REPETITICN:

Immediate corrective action was to reinstall the loose fan belt and operate the room cooler temporarily on one fan belt.

After surveillance were completed, the room cooler was taken out of service and the fan belts replaced. Work was donc under WR Q68920.

WORK REQUEST No.: Q68952 LER NUMBER: NA l

COMPONENT:

System 0300 - Scram outlet valve A0 127 on 30-31 had a broken l

diaphragm. Diaphragm was replaced and air tested.

l l

CAUSE OF MALFUNCTION: Unit I was in RUN mode at 99% power.

The H-8 control rod scrammed due to a failure of the scram outlet valve diaphragm (A0 305-127).

l The rod scrammed due to mechanical failure only.

L RESULTS & EFFECTS ON SAFE OPERATION:

Shfety consequences were minimal. The I

situation was analyzed by a Qualified Nuclear Engineer with H-8 at 00 position.

No further rod movements were considered necessary during the repair of the diaphragm.

ACTION TAKEN TO PREVENT REPETITION: Work Request Q68952 was written to repair the diaphragm. The diaphragm was replaced, the valve setpoint adjusted and piping checked for leaks. The rod successfully scrammed. The Station has been sampling the diaphragm on a refuel outage basis as a result of end-of-life failures of the diaphragms identified by General Electric in SIL #457.

1

WORK REQUEST No.: Q69054 LER NUMBER: NA COMPONENT: System 6600 - 1/2 Diesel Generator fuel oil transfer pump would not auto-transfer on loss of normal feed. A bad connection on coil was found.

I CAUSE OF MALFUNCTION: Unit 1 and 2 were in RUN mode at 100% thermal power when an attempt to perform special test 1-114 was made.

It was found that the 1/2 Diesel Generator fuel oil transfer pump would not auto-transfer to its alternate feed after losing its normal feed. The 1/2 Diesel Generator-was declared inoperable. The cause of the failure was found to be less than adequate contact between one of the coil contact retainer spring clips and one of the coil contacts.

RESULTS 6 EFFECTS ON SAFE OPERATION: The safety significance of this event was considered minimal. After the 1/2 Diesel Generator was declared inoperable, all surveillance testing required by Technical Specification 3.9.E was satis-factorilly completed.

ACTIGN TAKEN TO PREVENT REPETITION:

Immediate corrective action was to burnish the connection between the contactor coil and the coil contact retainer spring clip, as documented in Work Request Q69054.

Special Test 1-114 was completed without discrepancy. A program was to be established to periodically test and verify the operability of the alternate feeds on all safety-related equipment.

I j

UNIT 2 MAINTENANCE

SUMMARY

WORK REQUEST NO.:

Q65831 LER NUMBER:

88-007 COMPONENT:

System 220 - Outboard feedwater check valve 2-220-62B failed LLRT.

CAUSE OF MALFUNCTION: Unit 2 was in shutdown for the end of cycle nine refueling and maintenance outage. While performing local leak rate testing, the 2-220-58B and 2-220-62B valves volume was shown to have a leakage of 890.1 standard cubic feet per hour. This was excessive leakage. The 2-220-62B valve would not seat with air, and therefore, would not pressurize. The cause of the leak was not known at the time of the writing of the repair.

RESULTS & EFFECTS ON SAFE OPERATION: The safety of the plant was not jeopardized since local leak rate testing is a conservative method of measuring containment leakage. During an accident, the actual leak rate would be less than that determined by local leak rate testing.

ACTION TAKEN TO PREVENT REPETITION: No corrective action was taken at the time of the original report.

However, a supplemental report indicated that galling and wear were found on the nuts for the bonnet bolts, and galling on the studs for the bonnet seal ring. Action taken to insure the containment integrity of the valve was to install a new seat / disc assembly, o-ring, and pressure seal ring.

WORK REQUEST NO.:

Q65865, Q65866 LER NUMBER: 88-013 COMPONENT:

System 5400 - During an offgas isolation, the A0 5401B and 5401A valve demonstrated signs of leakage.

Installed a new valve and new gaskets.

CAUSE OF MALFUNCTION: Unit 2 was in RUN mode at 89% thermal power. The l

Steam Jet Air Ejector suction valves 2-5401A and B closed due to an isolation signal _from low pressure switch 2-3041-21B. During the event, condenser backpressure did not increase, indicating the valves were leaking. Unit 2 was in REFUEL mode at 0% thermal power when it was discovered that valve 2-5401B was installed incorrectly. The cause of the event is insufficient work instruction.

RESULTS & EFFECTS ON SAFE OPERATION: The safety significance of this event is minimal.

In the event of a high radiation signal, the environment would be isolated by MSIV closure even if the SJAE suction valves were not able j

to close.

ACTION TAKEN TO PREVENT REPETlTION:

Immediate corrective action was to replace l

the valve with a properly oriented valve.

Further corrective action was taken

]

to identify all safety-related butterfly valves on both units to determine that they are in proper orientation.

l

l l

l 1

WORK REQUEST NO.: Q65967 LER NUMBER: 88-007 COMPONENT:

System 2000 - Drywell flool drain sump discharge valve (AO-2001-3) failed LLRT. Valve was cleaned and replaced. Passed next LLRT.

CAUSE OF MALFUNCTION: The failure of the valve was found primarily to be the result of material buildup on the valve seats and in the valve grids, and damage to the seats.

Improper application of the valve was also found to be a cause of the problem.

RESULTS & EFFECTS ON SAFE OPERATION:

The safety of the plant was not jeopardized since local leak rate testing is a conservative method of measuring containment leakage. During an accident, the actual leak rate would be less than that determined by local leak rate testing.

ACTION TAKEN TO PREVENT REPETITION:

Corrective action was to clean and polish the valve wedge and seat. An "as left" LLRT measured a leakage rate of 1.5 SCFA.

A station review felt that this valve may not be suitable for this application.

An Action Item Report was written to investigate replacement of the valves.

Long range corrective action will be based on the results of this AIR (88-29).

WORK REQUEST No.: Q65969, Q65970, Q65972, Q65973, Q67083 LER NUMBER: NA COMPONENT:

System 0030 - Condensate pump room vault MK-490, MK-488, MK-478, MK-1028 and MK-479 (Vault C) failed LLRT. The penetration secl was tightened.

CAUSE OF MALFUNCTION: Unit 2 was in REFUEL outage. While performing local leak rate testing on the Unit 2 RHR SW vault, penetrations MK-1028, MK-478, MK-1031, MK-488, MK-490 and MK-479 showed signs of leakage.

The cause was determined to be slightly loose link seals around the penetrations.

RESULTS & EFFECTS ON SAFE OPERATION: The safety consequences of the event were minimal since the leaks could only be detected using the soap bubble solution, which is extremely sensitive to even the slightest leakage.

ACTION TAKEN TO PREVENT REPETITION: The link seals were tigFtened and adjusted on both sides to ensure integrity of the penetrations. Work Requests Q65969, Q67083, Q65970, Q65973 and Q65972 were issued to correct penetrations MK-490, MK-479, MK-488, MK-1028 and MK-478, respectively. The penetrations were successfully leak tested after the corrections were made.

i..

p-IV.

LICENSEE EVENT REPORTS' There were no Licensee Event Reports for Unit One and Two for the month of March.

4 V.

DATA TABULATIONS The following data tabulations are presented in this report:

A.

Operating Data Report B.

Average Daily Unit Power Level C.

Unit Shutdowns and Power Reductions 0027H/00612

APPENDIX C OPERATING DATA REPORT DOCKET NO.

50-254 UNIT One OATE April 7, 1989 COMPLETED 8Y Lynne Deelsnyder l

TELEPHONg 309-654-2241 OMMATING STATU$ 0000 030189 2400 033189 GROSE HOURS IN REPORTING PERICO:

744

1. REPORTIN0 PERICO:

2511 max. DEPENo. CAPACITY (WWo. Net): 769

2. CURRENTLY AUTHORIZED POWER LEVEL (g9 DEElGN ELECTRICAL RATING (MWo.Ned:

N/A

3. POWER LEVEL TO WHICH REETRICTED (IP ANY) (MWo.Neti:

4. REASONE POR RESTRICTION llP ANY):

THIS MONTH YM TO DATE CUMULATIVE 2160.0 119702.2__

5. NUMSER OP HOURS REACTOR WAS CRITICAL.............. 744. 0 0.0 0.0 3421.9 E. REACTOR RESERVE SHUTDOWN NOURE...................

744.0 2160.0 115819.2

. 7. MOURE GENERATOR ON LINE.........................

0.0 0.0 909.2 E. UNIT RESERVE SHUTDOWN HOURE.....................

i

............. 1756121 5102376 246792455

9. GROEE THERMAL ENERGY GENERATED (MWH) 571878 1667000 80024613

10. GROSE ELECTRICAL ENERGY GENERATED (MWM)...........

.. 548217 1598073 75163347

11. NET ELECTRICAL ENERGY GENERATED (MWM) 0 100 80.9 l

12. r EACTOR SE RVICE P ACTOR.......................

100 100

_ 83.2

13. REACTOR AV AILABILITY P ACTOR,..............

14. UNIT SERVICE P ACTOR............................. 100 100 78.2 100 100 78.9

15. UNIT AVAILABILITY P ACTOR.............

it. UNIT CAPACITY PACTOR (UelnE MOC)...................

95.8 96.2 66.0 93.4 93.8 64.4

17. UNIT CAPACITY P ACTOR (Uunt Desip MWel................

0.0 0.0 5.3

18. UNIT PORCED OUTAGE RATE.....

18. SHUTDOWNS SCHEDULEO OV ER NExT E MONTHS ITYPE DATE. AND OURATION OF EACN):

20. IP SMUT DOWN AT END OP REPORT PERIOO. ESTIMATED OATE OP STARTUP:

21. UNITS IN TEST STATUS (PRIOR TO COMMERCt AL OPERAT10N):

PORECAST ACHIEVED INITIAL CRITICALITY INITIAL ELECTRICITY COMMERCIAL OPERATION 1.l(>9

b-9 APPENOlX C 3

OPERATING DATA REPORT OOCKET NO.

50-265 1

UNIT Tun DATg April 7, 1989 COMPLETED gy I.vnne Deelsnyder l

TELEPHONg 309-654-2241

'l OMRArmesTAfus 0000 030189 744 2400 033189 GROSE HOURS IN REPORTING PERIGO:

1. REPORTim0 PEm00:

2511 max. OEPENO. CAPACITY (ARMe.80sel: 769

2. CURRENTLY AUTHORISE0 POWER LEVEL (y09 OESIGN ELECTRICAL RAflNG (MW>Nett:

N/A

3. POWSR LEVEL TO WHICH REETRICTED llP ANYI IMWpNetj;

4. REASONS POR RESTRICTION llP ANYl TMis MONTN YR TO DATE CUtdWLATIVS

8. NUMBER OF MOURS REACTOR WAS CRITICAL.............. 686. 9 2102.9 1110s?.A i

0.0 0.0 2985.8.

S. REACTOR RESERVE SHUT 00WN MOURE...................

f. MOURE GENERATOR ON UNE.......................,. 6 78. 6 2094.6 109826.3 0.0 0.0 702.9 E. UNIT RESERVE SMUTOCWN MOURE.....................

9. GROSE THERMAL ENERGY GENERATED (MWMD............ 1556638 4832552 235742825

.. 506414 123764 7551721s

10. GROSE ELECTRICAL ENERGY GENERATED (MWM).........

485277 1518492 71255069

11. NET ELECTRICAL ENERGY GENERATED (MWM) 92.3 97.4 76 R

12. r EACTOR SERVICE P ACTOR.......................

92.3 97.4 7 8. 8__

13. REACTOR AVAILABILITY PACTOR................

91.2 97.0 74.2

14. UNIT SERVICE P ACT OR........................

91.2 97.0 74.A

19. UNIT AV AILASI LITY P ACTOR........................

84.8 91.4 63.0

18. UNIT CAPACITY P ACTOR (UsinE MOCl.....................

82.7 89.1 61.4 I

17. UNif CAPACITY PACTOR (Umne Design WWel.............

8.8 3.0 8.4

}

18. UNIT PORCEO OUT AGE RATE.........................

19. SMUT 00WNS SCHEDULEO CVER NEXT E MONTHS ITYPE. DATE. AND OURAflON OF EACMl:

20. IF SMUT DOWN AT END OF REPORT PERIOD ESTIMATED DATE OP STARTUP:

21. UNITE IN TEST STATUS (PRIOR TO COMMERCI AL OPERATION):

PORECAST ACHIEVED INITI AL CRITICALITY INITIAL ELECTRICITY COMMERCIAL OPERAT1000 1.16 9 l

4 APPENDlX 8

-j

. AVERAGE DAILY UNIT POWER LEVEL DOCKET NO.

50-254 UNIT one DATE April 4, 1989 COMPLETED BY Lvnna Deelsnyder TELEPHONE 309-654-2241 MONTH Mnrch. 1989

. DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe Net)

(MWe Net) 1-745 765 37 762-751-2 3,

3 760 675 j,

794 741 4

g.

5 723 21 717 8

768 717 22 772 721 y_

23 757 735 g

24 g

765 735 5

10-745 g

702 11 721 27 709

- 12 722 742 g

719 726 13 3

14 718 3

706 l.

15 725 gi 731 gg 732 INSTRUCTIONS On this form, list the average daily unit power level in MWe-Net for each day in the reporting month. Compute to the nemiest whole megawatt.

These figures will be used to plot a graph for cach reporting month. Note that when nuximum dependable capacityis ugd for the net electrical rating of the unit, there may be occasions when the daily average power level exceeds the 1001 line (or the rntricted power level line). In such cases, the average daily unit power output sheet should be footnoted to explana the apparent anomaly.

1.16 8

.S g

9 i_____._______

2 APPENDlX 5 AVERAGE DAILY UNIT POWER LEVEL DOCKET NO.

50-265 UNIT Two DATE ' April 4, 1989.

COMPLETED BY Lynne Deelsnyder TELEPHONE 309-654-2241 MONTH March, 1989 9

DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe Net)

(MWe Net) 1 746 747 jy 420 2

gg 730

~9 708 3

13

-6 732 4

20

[

148 751 5

21 g

755 22 748 y

761 625 23 g

750 24 738 9

731 3

736 10' 740' 3

668 715 638

,39 27 12 687 3

701 13 756 779 3

748 763 14 30 15 749 31 738 q

ig 755 INSTRUCTIONS On this form, list the average daily unit power level in MWe Net for each day in the reporting month. Compute to the pensest whole megawatt.

These figures will be used to plot a graph for cach reporting month. Note that when maximum dependable capacity is ugd for the net electrical rating of the unit, there may be occasions when the daily average power level exceeds the 100'# line (or the restricted power level line). In such cases, the average daily unit power output sheet should be footnoted to explain the apparent anomaly.

1.16 8

~

l

1l!l D 28 g

E 9 1

4 V

3 1

2 S

O 6 g

2 T

R N

R 1 E

4 E

P g

G D

5 M

3 2

Y 6

M P U y

1 68 N

O A A S

9 C

S 9. L 0

/

- n1 0 o E

3 S

0it E

N 3 ss D

O i u I

Pvg T

Teu L

C QRA A

YB EV

. D E

I E

N T

T O

C E

H E

L P

R P

E R

M L

O O

E C

C T

eno

.S N

NO ITC w@

U 9

D 8

g$goU E

9 R

1 R

E h

DW c

)

O r

l w@0 a

XP a

I M

n DD hm$

i NN f

EA

(

PPS E

O 1

AN H

E N

W T

ST O

N NNT D

O EER T

M CVO U

I EP H

T L

E S

R R

O T

P I

E N

R U

{g" wo

$gR N)

OS E

I R N

TU O

AO RH T

U(

I D

N 9

U 8

S

"@w 9

1 4

E wg 5

I 2

T 5

I 0

C l

5 i

E D

r T

A p

A U

A D

Q

.o E

N M

A T

N A

E O

5 K

T E

N

/

C I

T D

O N

A I

D U

D

,I l

l

i' l

D 28 k

1 e

E 9 4

mc V

S 1

2 S

rn R

2 T

oa O G U

E N

f n R 1 D

4 E

re C

P Y

5 M

et G

y 3

2 N

6 M

P n P U 1 68 S

O i

S 9

L 9

C oa A

A

- n1 E

0

/

t M 0 o E

3 S

0it D

N d e 3 ss O

ev i u I

ml P vg L

T ma Teu C

aV QRA A

r Y

cf B

E S e V

i D

E I

yl E

N T

l e T

O C

lR E

H E

a L

P R

uc P

E R

ni M

L O

at O

E C

Ma C

T m

ro or t t cc ae el S

RE NO ITC X

U m@o E

D V

9 e $ g 0u L

ER 8

A 9

V 1

R E

DW

)

O h

l XP c

m8u a

r A

n IDD a

gm$

C i

NN M

f EA

(

P PS E

O 1

AN H

E N

W T

ST O

N NNT D

O EER T

M CVO U

I EP H

T L

E S

R R

O T

P I

E N

R U

2 N$h 5QM A

N)

OS O

I R 4

W TU T

AO 5

RH 6

T U(

I D

NU 9

8 S

9

"$w E

1 F

mh I

5 T

6 I

5 2

C l

E 2

0 D

i T

0 5

A r

A 3

U p

D 0

Q A

9 8

O E

N M

A 3

T N

A E

O 5

K T

E N

9

/

C I

T 8

D O

N A

I D

U D

flfl

l l'

VI. UNIQUE REPORTING RE0UIREMENTS The following items are included in this report based on prior commitments to the commission:

A.

Main Steam Relief Valve Operations Relief valve operations during the reporting period are summarized in the following table. The table includes information as to which relief valve was actuated, how it was actuated, and the circumstances resulting in its actuation.

Unit: Two Date: March 5, 1989 Valves Actuated No. & Type of Actuation 2-203-3A 1 Manual 2-203-3B 1 Manual 2-203-3C 1 Manual 2-203-3D 1 Manual 2-203-3 E 1 Manual Plant Conditions:

100 MWe Description of Events:

Surveillance Technical Specification 4.5.D.l.a.

B.

Control Rod Drive Scram Timing Data for Units One and Two The basis for reporting this data to the Nuclear Regulatory Commission are specified in the surveillance requirements of Technical Specifications 4.3.C.1 and 4.3.C.2.

The following table is a complete summary of Units One and Two Control Rod Drive Scram Timing for the reporting period. All scram timing was performed with Reactor pressure greater than 800 PSIG.

0027H/0061Z

u C. t e

m r u u

e q c

3. s e

c 3n S A

I

' n Nom g g

Oia n n

I t r i i

T ac m m

PcS i i

I i T

T Rf e Cig m m

Sca a a

Eer r r

Dpe c c

Sv S S

A l (

t t

a o

o c2 ll H

i n C.

2 1

hc3 e

t tk T3 i ir n

no U

UW 9

8 S

T 1

N L

3 O

E N

EM R

2 M

O E

1 I

%I T

R N

T0T c

U O

9R e

)

S C

O E

s 1

8 A

T XRS 99 86 E

AON 7

M 9

MFI F2 E2 2

8

(

G N

1 1

I M

1 I

T T

I 9

8 NU M

5 4 6

M O

N 0

A N

R

%W 93 2 2

R O

F A

C TR S

D AO S

H E

F E

ST O

W V

DI F0 NW 0 3 4

I S

F R

O 00 4 5

T R

D CY 5

L EL 2

1 1

E U

P 0

SL S

0 U

E R

NF R

I M

EO 0

MR 00 7 9

I F 29 6 T

6, D

0 0 0

EE GT AR RE 5

ES 7 0 8

VN 53 3 2

AI 0 0 0

S R O EO BR 8 M

8 1

UF NO 9

9 E 8 8

T A

4 4

O 2

3

en.

l VII.

REFUELING INFORMATION The following information about future reloads at Quad-Cities Station was requested in a January 26, 1978, licensing memorandum (78-24) from D. E.

O'8rien to C. Reed, et al., titled."Dresden, Quad-Cities, and Zion Station--NRC Request for Refueling Information", dated January 18, 1978.

0027H/00612 l

i

..es

'a o

QTP 300-S32 Revision 1 QUAD-CITIES REFUELING March 1978 INFORMATION REQUEST 1.

Unit:

01 Reload:

9 Cycle:

10 2.

Scheduled date for next refueling shutdown:

9-9-89 3

Scheduled date for restart following refueling:

12-11 __

4.

Will refueling or resumption of operation thereaf ter require a technical specification change or other license amendment:

NOT AS YET DETERMINED.

5.

Scheduled date(s) for submitting proposed licensing action and supporting information:

JUNE 10, 1989 6.

Important licensing considerations associated with refueling, e.g., new or

' different fuel design or supplier, unreviewed design or performance analysis methods, significant changes in fuel design, new operating procedures:

NONE AT PRESENT TIME.

l 7

The number of fuel assemblies.

a.

Number of assemblies in core:

724

)

j b.

Number of assemblies in spent fuel pool:

1773 1

8.

The present licensed spent fuel pool storage capacity and the size of any increase in licensed storage capacity that has been requested or is planned in number of fuel assemblies:

a.

Licensed storage capacity for spent fuel:

3657 b.

Planned increase in licensed storage:

0 l

9.

The projected date of the last refueling that can be discharged to the spent fuel pool assuming the present licensed capacity: 2008 XPPROVED APR 2 01970 i

Q.C.O.S.R.

1A--

=-

v'

~

QTP 300-S32 Revision 1 QUAD-CITIES REFUELING March 1978

{

INFORMATION REQUEST i

1.

Unit:

02 Reload:

9 Cycle:

10 2.

Scheduled date for next refueling shutdown:

2-3-90 3

Scheduled date for restart following refueling:

5-7-90 4.

Will refueling or resumption of operation thereaf ter require a technical specification change or other license amendment:

NOT AS YET DETERMINED.

5.

Scheduled date(s) for submitting proposed IIcensing action and supporting information:

NOVEHBER 2, 1990 6.

Important licensing considerations associated with refue!!ng, e.g., new or

'different fuel design or supplier, unreviewed design or performance analysis methods, significant changes in fuel design, new operating procedures:

NONE AT PRESENT TIME.

7 The number of fuel assemblies.

a.

Number of assemblies in core:

724 1

b.

Number of assemblies in spent fuel pool:

1475 8.

The present Ilcensed spent fuel pool storage capacity and the size of any increase in licensed storage capacity that has been requested or is planned in number of fuel assemblies:

a.

Licensed storage capacity for spent fuel:

3897 b.

Planned increase in licensed storage:

0 9

The projected date of the last refueling that can be discharged to the spent fuel pool assumirg the present licensed capacity: 2008 34' FD P R C) T/ E E) APIR 2 01978 C3. CL C). S. R.

a i

lE___

a.

.q--

i

)

VIII. GLOSSARY i

The following' abbreviations which may have been used in the Monthly Report, are defined below:

ACAD/ CAM -

Atmospheric Containment Atmospheric Dilution / Containment Atmospheric Monitoring ANSI American National Standards Institute APRM Average Power Range Monitor.

ATHS Anticipated Transient Without Scram Boiling Water Reactor BHR CRD Control Rod Drive Electro-Hydraulic Control System EHC

- EOF Emergency Operations Facility GSEP Generating Stations Emergency Plan High-Efficiency Particulate Filter HEPA HPCI High Pressure Coolant Injection System HRSS-High Radiation Sampling System IPCLRT Integrated Prirary Containment Leak Rate. Test IRM Intermediate Range Monitor ISI Inservice Inspection Licensee Event Report LER LLRT Local Leak Rate Test Low Pressure Coolant Injection Mode of RHRS LPCI LPRM Local Power Range Monitor MAPLHGR Maximum Average Planar Linear Heat Generation Rate MCPR Minimum Critical Power Ratio MFLCPR Maximum Fraction Lihitting Critical Power Ratio MPC Maximum Permissible. Concentration MSIV Main Steam Isolation Valve NIOSH National Institute for Occupational Safety and Health PCI Primary Containment Isolation PCIOMR Preconditioning Interim Operating Management Recommendations Reactor Building Closed Cooling Water System RBCCH RBM Rod Block Monitor Reactor Core Isolation Cooling System RCIC RHRS Residual Heat Removal System RPS Reactor Protection System RHM' Rod Horth Minimizer-SBGTS Standby Gas Treatment System SBLC Standby Liquid Control SDC Shutdown Cooling Mode of RHRS SDV Scram Discharge Volume SRM Source Range Monitor TBCCH Turbine Building Closed Cooling Hater System TIP Traversing Incore Probe TSC Technical Support Center 0027H/0061Z

u m

n-c.

. 'f^'N :C::mm::nw alth Edison

). Quad Cities Nuclear Power Station

-[

n

.g eg 22710 206 Avenue North

. Cordova, Illinois 61242 9740..

Telephone 309/654-2241-RAR-89-18 April 3,1989 Director of Nuclear Reactor Regulations r

U. S. Nuclear Regulatory Commission Mail Station P1-137 Washington, D. C.

20555-

' Enclosed.for your information.is the Monthly Performance Report covering the operation of Quad-Cities Nuclear Power Station, Units One and Two, during the. month of March, 1989.

L Respectfully, COMMONWEALTH EDISON COMPANY

-QUAD-CITIES NUCLEAR POWER STATION

.(g R. A'.'Ro ey Services Superintendent RAR/vmk/eb Enclosure

{$

l \\

i 0027H/0061Z

- - - - _ _ _ _ - _ - _ _ -. -

ML20245A231

Contents

Text

SUMMARY

Subject:

SUMMARY

Subject:

SUMMARY

SUMMARY

Navigation menu

ML20245A231

Text

SUMMARY

Subject:

SUMMARY

Subject:

SUMMARY

SUMMARY

Navigation menu

Search