ML17334B659

Petition Per 10CFR2.206 Requesting That OLs Be Modified, Revoked or Suspended Until Reasonable Assurance That Sys in Conformance W/Design & Licensing Bases Requirements

ML17334B659
Person / Time
Site:	Cook
Issue date:	10/09/1997
From:	LOCHBAUM D A UNION OF CONCERNED SCIENTISTS
To:	CALLAN L J NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO)
References
2.206, DD-99-03, DD-99-3, NUDOCS 9711180014
Download: ML17334B659 (30)

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NOTES: RECIPIENT ID CODE/NAME INTERNA FILE CENTER EXTERNAL: NOAC COPIES LTTR ENCL 1 1 1 1 RECIPIENT ID CODE/NAME NUDOCS-ABSTRACT NRC PDR COPIES LTTR ENCL 1 1 1 1 CATEGORY 2 REGULA~.Y INFORMATION.DISTRIBUTIO YSTEM (RIDS)ACCESSXON NBR:9711180014 DOC.DATE: 97/10/09 NOTARIZED:

NO DOCKET FACIL:50-315 Donald C.Cook Nuclear Power Plant, Unit 1, Indiana M 05000315 50-316 Donald C.Cook Nuclear Power Plant, Unit 2, Indiana M 05000316 AUTH.NAME AUTHOR AFFILIATION LOCHBAUM,D.A.

Union of Concerned Scientists RECIP.NAME RECIPIENT AFFILIATION

'CALLAN,L.J.

Ofc of the Executive Director for Operations

SUBJECT:

Submits petition per 10CFR2.206 recgxesting that operating licenses be modified, revoked or suspended until reasonable assurance that sys in conformance w/design s licensing recgxirements.

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50 Docket (PDR Avail)E Q 0 D U N NOTE TO ALL"RIDS" RECIPIENTS:

PLEASE HELP US TO REDUCE WASTE.TO HAVE YOUR NAME OR ORGANIZATION REMOVED FROM DISTRIBUTION LISTS OR REDUCE THE NUMBER OF COPIES RECEIVED BY YOU OR YOUR ORGANIZATION, CONTACT THE DOCUMENT CONTROL DESK (DCD)ON EXTENSION 415-2083 TOTAL NUMBER OF COPIES REQUIRED: LTTR 4 ENCL 4

~\UNION OF CONCERNED SCIENTISTS October 9, 1997 Mr.L.Joseph Callan Executive Director for Operations United States Nuclear Regulatory Commission Washington, DC 20555-0001 SUB JECTi PETITION PURSUANT TO 10 CFR'2.206, DONALD C.COOK NUCLEAR PLANTS UNITS 1 AND 2, DOCKET NOS.50-315 AND 50-316

Dear Mr.Callan:

The Union of Concerned Scientists submits this petition pursuant to 10 CFR 2.206 requesting that the operating licenses for Donald C.Cook Units 1 and 2 be modified, revoked, or suspended until there is reasonable assurance that their systems are in conformance with design and licensing bases requirements.

A process comparable to the system certifications recently used by the Salem and Millstone licensees would provide this necessary level of assurance.

UCS additionally requests that a public hearing into this matter'e held in the Washington, DC area prior to the first unit at D C Cook being authorized to restart.At this hearing, we will present information supporting the contentions in this petition.BBack Bround h On October 9, 1996, the NRC requested that its power reactor licensees provide information pursuant to 10 CFR 50.54(f)regarding the adequacy and availability of design bases information.

The NRC's issued this request as a result of its investigations at the Millstone Power Station.The licensee for the D C Cook plant responded with a letter dated February 6, 1997, describing the administrative controls it uses to provide assurance that the Cook Nuclear Plant is operated and maintained within the established design bases.An NRC team recently conducted an architect/engineer design inspection at D C Cook.According to the NRC's Project Manager for 9 C Cook, this NRC team examined two safety systems and their supporting systems.The team's findings forced the licensee to shut down both units on September 10, 1997.The NRC issued a confirmatory action letter to the licensee dated September 19, 1997, specifying issues arising from the design inspection that must be resolved prior to restarting the units.These issues (listed in Attachment 1)include physical modifications to the plants and revisions to the plants', operating licenses.Numerous NRC Daily Event Reports (listed in Attachment 2)described the findings&om design inspection as reported by the licensee.The NRC has not yet released the design inspection report and we have been told that it will not be issued until next week at the earliest.PDR noOCi 0Sao03XS H PDR llllllliilllllllllllllllllllllllllllliilI.,['.~~~Washington Office: 1616 P Street NW Suite 310~Washington, DC 20036-1495

~202-332-0900

~FAX: 202-332-0905 Cambridge Headquarters:

Two Brattle Square~Cambridge, MA 02238-9105

~617-547-5552

~FAX: 617-864-9405 California Office: 2397 Shattuck Avenue Suite 203~Berkeley, CA 94704-1567

~510-843-1872

~FAX: 51 0-843-3785

'I I~~'

i~~~~October 9, 1997 Page2of4 Basis for R uested Action The NRC conducte'd architect/engineer design inspections at only six of its nearly 70 operating power reactor licensee sites.These design inspections examined only one or hvo safety systems along with their supporting systems at each site.The NRC Project Manager reported that the design inspection at D C Cook examined the residual heat removal and component cooling water systems along with their supporting systems.These design inspections focused on the facilities'riginal design and the licensees'onformance with the safety analysis reports.The systems examined by the NRC at D C Cook had already been covered by the licensee's design basis documentation reconstitution pr'ogram.Design basis documents (DBDs)for the containment, containment structure, containment spray, emergency core cooling, component cooling water, and residual heat removal systems had been approved by the licensee prior to the NRC team's arrival.The licensee informed the NRC that its'BD program had not identified any deficiencie involving equipment operability.

The findings by the NRC design inspection team prompted the licensee to declare both trains of the emergency core cooling systems and the containment spray system inoperable.

The units were shut down on September 8 and 9, 1997.The licensee reported making physical changes to the plant to correct some of the problems and indicated that additional physical changes may be required.The licensee has proposed fixing the specific operability issues identified during the NRC design inspection and then restarting the units.Confining the scope of the restart activities in this way would be treating the symptoms rather than the cause of the problems.The NRC design inspection revealed serious deficiencies in the licensee's design control programs.These deficiencie crcatcd the specific problems that forced the plants to be shut down.These deficiencies

>ay also be responsible for similar problems in other safety systems which were not examined by the NRC.It is important to note that the NRC identified significant operability problems in systems that the licensee had covered in recently approved DBDs.The licensee stated in its February 6, 1997, submittal that.it verifies and validates the information in its DBDs via reviews and physical plant walkdowns prior to their approval.Thus, the NRC discovered significant problems in systems which had been closely scrutinized by the licensee.Had the NRC's findings involved systems which have not yet been covered under the licensees'BD program, it might be reasonable to assume that the licensee would have identifiicd them at that later date.However, there is little reason to believe that these problems would have.been resolved unless the NRC had identified them.Attachment 2 lists NRC Daily Event Reports (DERs)involving issues identified by the NRC design inspection at D C Cook.DER Nos.32740, 32806, 32822, 32839, 32843, 32875, 32890, 32904, 32914, 32915, 32921, 32948, and 329S8 describe potential deficiencies that appear to have existed at D C Cook prior to the initiation of its design basis documentation reconstitution effort in 1992.That effort was therefore apparently unable to detect these potential deficiencies.

DER Nos.32823, 32824, 32903, 32939, and 3294S describe potential deficiencies that appear to have been introduced since 1992.Thus, the licensee's design control and quality assurance programs are apparently unable to ensure that the facility is maintained within its design bases.

0 0 I~r ,f h't ,~

~~~Q October 9, 1997 Page3 of4 I UCS feels that the design basis documentation reconstitution and Updated Final Safety Analysis Report (UFSAR)validation programs as described in the licensee's response to the NRC's 50.54(f)letter lack the rigor and focus necessary to identify potential design-related operability issues.Our conviction is supported by the findings&om the NRC design inspection.

Since the corrections to the NRC's findings were not limited to mere paperwork fixes but included actual changes to the plant's physical configuration, the safety significance of these and potentially other undetected problems cannot be understated.

I~The fiaws in the licensee's design control programs must be corrected.

The systems at D C Cook, at least those with a safety function, must be certified to be capable of performing their required actions under all design conditions.

Then, and only then, can the units bc restarted with reasonable assurance that public safety will be adequately protected.

It would be irresponsible to restart these units knowing that the programmatic failures that caused the safety problems identified by the NRC team may have produced comparable problems affecting the operability of other safety'systems.

The legal precedent for our position is stated by the NRC's Atomic Safety and Licensing Appeal Board in'he Matter of Vermont Yankee Nuclear Power Corporation, Memorandum and Order (ALAB-138), dated July 31, 1973: "As a general rulc, the Commission's regulations preclude a challenge to applicable regulations in an individual licensing proceeding.

10 CFR 2.758.This rule has been frequently applied in such proceedings to preclude challenges by intervenors to Commission regulations.

Generally, then, an intervenor cannot validly argue on safety grounds that a reactor which meets applicable standards should not be licensed.By the same token, neither the applicant nor the staff should be permitted to=-challenge applicable regulations, either directly or indirectly..

Thus, those parties should not generally be permitted to seek or justify the licensing of a reactor which does not comply with applicable standards.

Nor can they avoid compliance by arguing that, although an applicable, regulation is not met, the public health and safety will still be'protected.

For, once a regulation is adopted, the standards it embodies represent the Commission's definition of what is required to protect the public health and safety."[emphasis added]"In short, in order'for a facility to be licensed to operate, the applicant must establish that the'acility complics with all applicable regulations.

If the facility does not comply, oi if there has been no showing that it does comply, it may not be licensed."[emphasis added]The NRC design inspection at D C Cook identified significant issues which caused both units to be shut down.These issues were caused by programmatic deficiencies in the licensee's design control pi'ograms.

A contributing factor for these issues is the failure of the licensee's quality assurance and self-assessment programs to detect these problems.Nothing in the reported findings from the design inspection supports a conclusion that these findings are isolated consequences.

The NRC's design inspection invalidates any showing that this facility complies with all applicable'regulations.

Therefore, the design control deficiencies must be corrected to prevent future non-compliances with safety regulations.

And just as importantly, a.thorough review of all systems with safety functions must be'completed prior to restart to detect and correct past non-compliances.

P J~P t l t'I V October 9, 1997 Page 4 of 4 1 h r UCS.is not advocating that the NRC apply-a higher standard at D C Cook.-Instead, we are requesting that the NRC ensure that the D C Cook facility is in accordance with the minimum safety standards which constitute the legal grounds fo'r allowing the units to operate.Our request is consistent with the measures required by the NRC when other sampling inspections find problems.We ask the NRC to expand the inspection scope.based upon the identified problems just as would be required when snubber (e.g., pipe restraint) and reactor vessel internals inspections found problems: 'uested Actions*~I'CS petitions the NRC to protect public health and safety by preventing the units at D C Cook from operating until such time that there is reasonable assurance that all significant non-compliances have been identified and corrected.

The system certification process recently used at the Salem Generating Station and the Millstone Power Station would provide such reasonable assurance.

We request a public hearing on this matter be held in the Washington, DC area before any unit at D C Cook is authorized to restart.Sincerely, auiug'David A.Loch aum Nuclear Safety Engineer CC: Chairman Shirley Ann Jackson.United States Nuclear Regulatory Commission Washington, DC 20555-0001.

Honorable Spencer Abraham United States Sen'ate Washington, DC 20510-2203 Mr.A.B.Beach, Regional Administrator United States Nuclear Regulatory Commission

.801 Warrenville Road Lisle, IL 60532-4351 Honorable Carl Levin United States Senate.Washington, DC 20510-2202

.Honorable Fred Upton United States House of Representatives Washington, DC,20515-2206

~Attachments:

1)Design Inspection Issues, That Will Be Resolved Prior to D C Cook Restart 2)NRC Daily Event Reports on D C Cook De'sign Inspection Findings=V I g 0 J I~S I n'l/J v s/l P, I'I I r p hr II I

~~Py Attachment 1 Design Inspection Issues That Will Be Resolved Prior to D C Cook Restart 1 The following issues, quoted verbatim, were specified on the NRC's Confirmatory Action Letter dated , September 19, 1997, as requiring resolution prior to restart of any D C Cook unit:.1.Recirculation Sump Inventor'y/Containment Dead Ended Compartnients Issue Analyses will be performed to demonstrate that the recirculation sump level is adequate to prevent'ortexing," or appropriate modifications will be made.[See also Attachment 2-Power Reactor Event Number 32890], Recirculation Sump-Venting Issue'r r!II Venting will be re-installed in the recirculation sump cover.The design will-incorporate foreign material exclusion requirements for the sump.[See'also Attachment 2-Power Reactor Event Numbers 32875 and 32903]Thirty-six Hour Cooldown, with One Train of Cooling 4 Analyses will be performed that will demonstrate the capability to cool down the units consistent with design basis requirements and necessary changes to procedures will be completed; ES-1.3 (Switchover to Recirculation Sump)Procedure 6.Changes to the emergency procedure used for switchover of the emergency core cooling-and containment spray pumps to the recirculation sump will be implemented.

These changes will provide assurance there will be adequate sump volume, with pr'oper corisideration of instrument bias and single failure criteria.[See also Att.2-Power Reactor Event Numbers 32806 and 32904]h ll , Compressed Air Overpressure Issue Overpressure protection will be provided downstream of the 20 psig, 50 psig, and S5 psig control air regulators to mitigate the effects of a postulated failed regulator.

[See also Attachment 2-Power Reactor Event Numbers 32939 and 32988]I Residual Heat Removal (RHR)Suction Valve Interlock Issue'technical specification change to allow operation in mode 4 with the RHR suction valves open and power removed is being processed.

Approval.of this change by the NRC will be required prior'o restart.[See also Attachment 2-Power Reactor Event Numbers 32914 and 32921]-Fibrous Material in Containment Removal of fibrous material from containment that could clog.the recirculation sump will be corn'pleted.

[See also Attachment 2-Power Reactor Event Number 3294S]

0 l I~'

Attachment 2 NRC Daily Event Reports on'C Cook Design Inspection Findings The following summaries were taken from the daily event reports available on the NRC's wcbsite (mvw.nrc.gov).

~The only editing involved dclction of unnecessary detail, such as who was notified about the events,'and the addition of clarification for acronyms.Othcrivisc, these narratives are verbatim.1 POWER REACTOR EVENT NUMBER: 32890 UNUSUAL EVENT, DECLARED 4 TECHNICAL SPECIFlCAITON REQUIRED SHUTDOWN ON BOTH UNITS DUE TO INOPERABLE CONTAINMENTS

+a result of issues raised during the ongoing architect/engi'neer design inspection, the liccnsce was reviewing the design aspects of the containmcnts (both units, have similar containmcnts).

Aflcr consulting with the nuclear steam supply system supplier (Westinghouse) the licensee determined that concerns existed about whether adequate communication (flow paths)exists between the active and inactive portioris of the containmcnt sump.During certain scenario, the volume of water flow back to the containment recirculation sump may not be adequate to support long-term emergency core cooling (ECC)systems (RHR[residual heat removal]system, safety injection system, charging system)or containment spray pump operation during thc recirculation phase of a large or small.brcak LOCA.Thc containmcnt drainage system is designed to ensure that water en'tering thc containment from the breach in tlie reactor coolant system, ECC systems-injection, and ice condenser melt flows back into the.containment recirculation sump via drains.Licensee analysis was unable to confirm that suflicient communication

-'xisted bctwccn inactive and active volumes of the containment to ensure adequate drainage to the recirculation sump.-Without adequate drainage into the sump, a low sump level will result, which jeopardizes long term operation of the ECC Systems and containment spray pumps due to vortcxing and air entrainment.

As a conservative measure because of these concerns, the licenscc declared both trains of thc ECC Systems and the, containment spray system inoperable for both units and entered Tcchnical SpcciTication limiting condition for operation action statement 3.0.3 to shut down both units.The liccnsce commenced shutting Unit 1 down from 100%power at 1655 aild Uill't 2 down fi'om 100%power at 1728.At 2000, the licensee dcclarcd an unusual event on both units due to the'potential loss of containmcnt barrier, on both units.The licenscc plans to perform further analysis to determine the extent of thc existing communication between the portions of the sumps and whether plant modifications will be necessary.

~~~Update0311 EDT on 09/10/97 by Tilly taken by MacKinnon*~~

I Thc unusual cvcnt was tcrminatcd and exited at 0303 EDT when.Unit 1 cntcrcd mode 5 (cold shutdown).

Unit 2 cntcred mode 5 at 0015 EDT (cold shutdown).

f POWER REACTOR EVENT.NUMBER: 32875 FAILURE TO MAINTAIN THE CONTAINMENT RECIRCULATION SUMP 1/4" PARTICULATE RETENTION REQUIREMENT (HISTORICAL ISSUE)I k.A 1/4" particulate retention requirement for the containment recirculation sump was not properly established in 1979 following sump modifications.

The containment recirculation sump rcquircment to retain 1/4" particles is to ensure that containment spray, nozzles do not become plugged.Thc containmcnt spray system takes suction from the containment recirculation sump following injection of the refueling'water storage tank supply during a loss of coolant accident.

~~l l r 1 I Attachment 2 (continued)

NRC Daily Event Reports on D C Cook Design Inspection Findings II In 1979, modifications were performed on the containment recirculation sump.One of the modifications involved moving a 1/4" rctcntion element from inside the recirculation sump to the entrance of the sump.When the'retention element was moved, thc 1/4" retention requirement ivas not fully addressed, and pathways excccding the 1/4" requirement, were inadvertently established.

Thc inadvcrtcnt pathways established included: 3/4",vents in the roof of the recirculation sump entrance, the containment sump drain line from the recirculation sump, and small gaps around the sump entrance.These pathways have since been elimi'nated or the 1/4" requirement has been established.'

.Thc licensee is reporting the fact that since 1979,until the 1/4" requirement was established or the pathway was eliminated, the containment recirculation sump did not meet its design rcquircment.

Thc containment recirculation sump currently meets the 1/4" requircmcnt.

A condition rcport has,been written to initiate investigation into tlus event and determine appropriate preventive actions.This event was dctcrmined to bc reportable at 0856 on September 5, 1997.l~~~Update at 1905 on 09/10/97 by Randy Ptacck entered by JolliQ'e~**'I Mer further review of the above condition, the licensee concluded that thc emergency core cooling (ECC)system was outside its design basis as a result of the 1/4" rcquircmcnt not being mct following thc 1979 plant modifications.

By not adequately covering the 1/4" particulate retention requirement, larger particles had the potential to enter thc recirculation sump.Thc ECC System has not been analyzed for these larger particles nor is it within tlie design of the ECC System to handle these larger particles.

The licensee has concluded that this event is also rcportablc to thc NRC in accordance with the requirements of'0CFR50.72(b)(1)(ii)(a) unanalyzed condition, and 10CF50.72(b)(2)(iii)(d)accident mitigation.

I POWER REACTOR EVENT NUMBER: 32903'ONTAINMENT RECIRCULATION SUMP VENT HOLES HAVE SEEN FILLED WITH CONCRETE r I As a result of questions posed by the NRC architect/engineer design inspection team, the licensee detcrmincd that~thc inlet venting requirement for thc containmcnt recirculation sumps was not properly maintained following modifications to thc Unit 2.sump in 1996 and the Unit 1 sump in 1997 (both units have similar containmcnts).

'he containment recirculation sump venting rcquircment was cstablishcd in 1979 as part of the original sump design to reduce the potential for air entrainment through the sump.The venting requirement wIas met through the, addition of five 3/4-inch diameter holes drilled in thc roof of thc sump inlet.(The holes did not meet the 1/4-inch~diamctcr requirement as reported in Event¹32875.)When these holes werc discovered during the Unit 2 1996 refueling outage and the Unit 1 1997 refueling outage, they were classified as abandoned equipment holes that exceeded the 1/4-inch particulate retention rcquiremcnt for thc sumps and they were filled with concrete.

0'I/II I Attachment 2 (c'ontinued)

NRC Daily Xvent Reports on D C Cook Design Inspection Findings POWER REACTOR EVENT NUMBER;32806~INSTRUMENTATION INDICATIONS USED TO DETERMINE WHEN REFUELING WATER'TORAGE TANK TO CONTAINMENT SWITCHOVER IS'REQUIRED MAY NOT HAVE BEEN CORRECT TO PREVENT VORTEXING IN THE CONTAINMENT RECIRCULATION SUMP.*I During the evaluation of a proposed procedure change that aGects mvitchover from the refueling water storage tank (RWST)to the containment sump during a loss-of-coolant, accident (LOCA), it was dctermincd that the instrumentation indications used to determine when the switchover is required may not have bccn correct.to prevent vortcxing in the containmcnt recirculation sump.I To address this situation, procedures associated with the mvitchover (on both units)have been conservatively changed to accommodate the related instrument inaccuracies.

These changes assure adequate RWST water is in containment before mvitchover to eliminate concerns that vorteung would occur in thc containmcnt sump after switchover.

The problem is that the RWST water level indicators are connected to tlic suction linc that goes to the residual heat rcmov'al (RHR)pumps.Due to thc flow in these lines, the indicated water level at winch the switchover would be initiated would be less than the actual water level of the RWST (thc licensee would bc putting less water into the containmcnt than er~ted).Also, thc licenscc said that they liad some inaccuracies associated witli their containmcnt sump instrumentation.

The licensee adjusted thc containmcnt sump indication to assure that they have an adequate volume in the containment to prevent vortexing.

The licensee relies upon two indications for mvitchover, RWST water level and containmcnt water level.POWER REACTOR EVENT NUMBER: 32904.SINGLE FAILURE DURING RECIRC SUMP SWITCHOVER COULD BE UNANALYZED CONDITION I'I As a result of questions posed by the NRC arclutect/engineer design inspection team, the licensee determined that thc possibility of a single failure during an accident wlulc performing switchover of the emergency core cooling~system pumps from the refueling water storage tank (RWST)suction to the recirculation sump suction could have resulted in thc plant being in an unanalyzed condition.

Tlus condition is outside thc plant design basis, and it potentially could have prevented the fulfillment of a safety function of structures or systems.The plant emergency, operating procedures (EOPs)as currently written require that the west residual heat removal (BHR)pump bc the first pump mvitchcd from thc RWST suction to the rccirc sump suction.Once this is*accomplished, the centrifugal charging (CC)pumps'uctions and the safety injection (Sl)pumps'uctions arc then swapped'from the RWST supply to the discharge of the west RHR pump.If thc west RHR pump werc to fail at this., point when all CC and SI pumps were being supplied from its discharge, prior to thc east RHR pump suction being transferred from the RWST to thc rccirc sump, all CC and SI pumps could also fail duc to thc loss of suction flow.This would result in the loss of all high and medium head injection with only the flow from the east RHR pump available for injection into the reactor coolant system.Thc liccnsec is currently reviewing thc EOPs to determine an alternate mvitchovcr sequence that would eliminate the condition as described above.l

~~I A t 1 Attachment 2 (continued)

NRC Daily Event Reports.on D C Cook Design Inspection Findings POWER REACTOR EVENT NUMBER: 32939 INSTALLED PLANT.MODIFICATION INTRODUCED THE POSSIBILITY.

OF A SINGLE FAILURE WHICH COULD RESULT IN THE LOSS OF BOTH TRAINS OF THE ESF VENTILATION SYSTEM.At 1620 on 09/16/97, the licensee determined that a plant modiTication installed behveen December 1996 and August 1997 introduced the possibility of a single failure which could result in the loss of both trains of the engincercd safety features (ESF)ventilation system if'the 85-psi air header was to be lost.Prior to thc installation of the'plant modification, the ESF ventilation system charcoal inlet and bypass dampers both utilized a 20-'psi air header and werc positioned such that the charcoal bypass dampcrs werc normally open and would fail closed;and thc charcoal inlet dampers were normally closed and would fail open.The plant modification installed ncw bypass dampers which required higher air prcssure to operate and were, thercforc, transferred to thc 85-psi header.If the,~85-psi air header was lost, it would result in thc,rcpositioning of the normally open bypass dampers ivithout the,, opening of the charcoal inlet dampers on both trains.This would result in dead heading of thc filter train fans and-loss of cooling to emergency core cooling system (ECCS)equipment.

" POWER REACTOR EVENT NUMBER: 32988 NON-SAFETY~RELATED AIR HEADERS LACK OVERPRESSURE PROTECTION f.'uring an arclutectural engineering inspection a question was raised regarding the lack of ovcrpressurc protcction-on thc 20, 50 and 85 psig control air headers.Thc specific concern is the potential for common mode failure of both trains of safety related equipmcnt served by thc,air hcadcrs'.The ovcrprcssurc condition is'ostulated'to be caused by regulator failure.Although system rcvicws have found no,component failure mode which would result in the devices being incapable of going to their fail-safe position, a design change package has been prepared to provide ovcrprcssure protection, on the 20, 50 and 85 psig headers.I f-POWER REACTOR EVENT NUMBER: 32914 LICENSEE IDENTIFIED THAT BOTH UNITS HAD OPERATED THEIR RHR SYSTEM CONTRARY TO THE DESCRIPTION IN THE FSAR.'C At 1615 EDT, with Units 1 and 2 shutdown in mode 5, it was dctcrmined that both units have operated contrary to thc design basis for the residual heat removal (RHR)system as described in the FinaL Safety Analysis rcport (FSAR).FSAR Chapter 9, Section 9.3, describes the interlocks associated with the residual licat removal (RHR)suction valves from thc reactor coolant system (RCS).The suction linc valves arc interlocked through separate, channels of the RCS system prcssure signals to provide automatic closure of both valves whenever RCS prcssure cxcceds RHR design prcssure..Thc FSAR states that the interlock may be dcfcatcd when thc RCS is open to atmosphere.

However, for a number of years this interlock has been procedurally'defeated on both units to prevent inadvertent closure and loss of RHR suction during shutdown cooling operation by opening the valves and racking'ut their.breakers in mode 4.'Thc ovcrpressurc protection afforded by thc automatic closure function dcscribcd in the FSAR was defeated without a safety evaluation being pcrformcd.

This loss of automatic closure function represents an unanalyzed condition and is, thcrcforc, reportable.

0 J 4 1

~~~~'ttachment 2 (continued)

, NRC Daily Event Reports on D C Cook Design Inspection.Fin'dings I Plans are to degas, dcpressurizc, and open the RCS on both units to atmosphere.

Degas will start on Unit 1, and when completed, thc unit will proceed to depressurize while Unit 2 starts degas procedures.

When the RCS is open to atmosphere on both units, the plant will be in compliance with the FSAR.This condition was identified by the1iccnsee during an ongoing NRC architect/engineer inspection.

II*~~Update at 2130 EDT on 9/13/97 fiom Robert Blyth to S.Sandin~**The licensee has completed its safety evaluation for mode 5 operation and concluded that thcrc was no unreviewcd safety question or change of operation as described in thc FSAR.Conscqucntly, degas of Unit 1 has been"'erminated, and neither unit will bc vented to atmosphere.

POWER REACTOR EVENT NUMBER: 32921 THE LICENSEE IDENTIFIED THAT BOTH RHR PUMPS HAD BEEN OPERATED WHEN THE RCS WAS DEPRESSURIZED, WHICH IS CONTRARY TO THE DESCRIPTION IN THE FSAR.Chapter 9 of the Final Safety'nalysis Rcport (FSAR)states: 'Only one residual heat removal'(RHR) pump will be" operated when the reactor coolant system is open to atmosphere to prevent damaging both pumps in the unlikely" event that suction should be lost.'perating proccdurcs for, the RHR system do not prevent operation of both RHR pumps when thc reactor coolant'system (RCS)is open to atmosphere, and in thc past, both RHR pumps have been run when the RCS was vcntcd to atmosphere.

I Plant operating proccdurcs are being reviewed to determine the impact.Procedure changes will be implemented as necessary to address the FSAR rcquircmcnt.

A condition rcport has bccn initiated to investigate'and determine appropriate preventative actions.H POWER REACTOR EVENT NUMBER: 32948 IT WAS DETERMINED THAT FIBROUS MATERIAL IS PRESENT IN BOTH UNIT 1 AND UNIT 2-CONTAINMENT IN ENOUGH QUANTITY TO POTENTIALLY CAUSE EXCESSIVE BLOCKAGE OF THE CONTAINMENT RECIRCULATION SUMP SCREEN DURING THE RECIRCULATION PHASE OF A LOSS OF COOLANT ACCIDENT.In 1985, 1986,,and 1995"Fiberfrax" refractory insulation materials in bulk, blanket or board form werc used as damming material when installing fire stops in cable trays in both containments.

Thc specification governing installation of the fire stops did not require removal of thc material, only.stating that it should be removed"if necessary." The material was not removed.The material is prcscnt in 12 cable, trays in Unit 1 and 15 cable trays in Unit 2.r When the Fiberfrax is exposed to water or steam/water environment it could potentially, break into small pieces', which could be transported to the recirculation sump by the-water, flow in containment during a loss of coolant accident.Once it reaches the recirculation sump it has the potential to clog the scrccns in excess of tlic design value.Excessive screen blockage could result in ECCS inoperability during the recirculation mode.The Fibcrfrax material,is currently being removed from the containmcnts, and removal will bc completed prior to , restart of the units.The possibility that the licensee's work e'ontrol"process allowed uncncapsulatcd fibrous material to be installed in other locations inside containment is being investigated.

A K'P h I J l r k t I I P I V 1, Attachment 2 (continued)

NRC Daily Event Reports on D C Cook Design Inspection Findings POWER REACTOR EVENT NUMBER: 32740 UNITS 1&2 OPERATED OUTSIDE THEDESIGN BASIS FOR SERVICE WATER INLET TEMP's a result of questions posed by members of the ongoing NRC design inspection team, thc licensee has'etermined that Units 1&2 have operated outside the plant design basis for service water inlet temperature.

/'t The Updated Final Safety Analysis Report (UFSAR), Table 9.5-3, lists service water inlet temperature design valueas 76'F.This value is used as input to analyses such as containment peak prcssure and control room habitability.

Although engineering analyses were performed in 1988'raising the temperature to 87.5'F as listed in the plant Tcdmical Specifications, a 10CFR50.59 safety evaluation was never performed, nor was thc UFSAR properly'evised:

Plant service water inlet temperature is thc same as Lake Michigan water tcmpcraturc.

A review of historical data indicates that during July'and August of any year, Lake Michigan"water temperature is likely to exceed thc 76'F value.Specific data for 1997 shows that Lake Michigan water temperature, and thcrcforc plant service water inlet'emperature, was greater than 76'F on July 17, July 18, and August 4, 1997.All plant systems which utilize service water as a cooling medium have bccn dctcrmincd to bc operable.A 10CFR50.59 safety evaluation will be-p'erformed and appropriate changes will be incorporated into thc UFSAR.This report is intended to cover any temperature exclusions above 76'F and below the 87.5'F value listed in the plant Technical Specifications that may occur prior to the completion of thc 10CFR50.59 safety evaluation.

POWER REACTOR EVENT NUMBER: 32822*DISCOVERY THAT A NORMAL OPERATING PROCEDURE ALLOWED PLANT OPERATION WITH COMPONENT COOLING WATER HEAT EXCHANGER OUTLET TEMPERATURES GREATER THAN THE DESIGN LIMIT SPECIFIED IN THE FINAL SAFETY ANALYSIS REPORT During the'ongoing NRC architect/engineer design inspection, a question,was asked relative to a statement used in the normal operating procedure for the component cooling water (CCW)system.The statement allows for a heat-exchanger outlet temperature for CCW to reach 120'F for a period of 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> during normal cooldown on the residual heat removal system.Investigation revealed that this statement divas in the original issue of the procedure in 1976.However, no 10 CFR 50.59 unreyicwed safety evaluation determination documentation could be found to suppoit tlus design parameter.

Thc licensee's Final Safety-Analysis Rcport (FSAR)states that thc CCW heat exchanger outlet design temperature

's 95'F.Based on the FSAR requiring the 93'F outlet temperature and the.lack of an unrcvicwed safety question dctcrmination to justify operation cxcceding 95'F, the units were in a condition that allowed operation outside the design basis because thc procedure allowed operation up to 120'F for a period of 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> during normal cooldown on the residual heat removal system.The units arc n'ot cuncntly in a Technical Specification limiting condition for operation as a result of tins issue.II'Procedure changes have been made to remove thc inappropriate statement.

A condition rcport has also been written to initiate an investigation into this event and determine appropriate preventive actions.

0 V/4 S 4 1 lI

~)~~I Attachment 2 (continued)

NRC Daily Event Reports on'D C-Cook Design Inspection Findings POWER REACTOR EVENT NUMBER: 32823 FAILURE OF A SAFETY REVIEW TO ADDRESS FINAL SAFETY ANALYSIS ATTRIBUTES ON, ASSOCIATED COMPONENT COOLING WATER COOLING REQUIREMENTS During, the ongoing NRC architect/engineer design inspection, a question was asked relative to dual train component cooling water (CCW)system outages.During dual train CCW outagcs, CCW cooling is supplied to thc spent fuel pool (SFP)heat exchanger only from the opposite unit.If that unit has a loss of coolant accident (LOCA), CCW to the SFP heat exchanger will isolate.Final Safety Analysis Rcport (FSAR)Table 9:5-2, footnote 3, indicates that the SFP heat exchanger is assumed to bc on the non-accident unit.h The licensee reported the following inspection questions:

r , 1)Does a dual train-CCW outage represent a condition outside thc plant design basis2 2)Was this reviewed as part of the process of allowing a dual train CCW outage2'ased on a review of FSAR Table 9.5-2, it was concluded that footnote 3 was established to clarify why no values for SFP heat exchanger flow for thc unit undergoing the LOCA are listed in the table.Footnote 3 reflect normal'FP cooling system design and operation.

I A review was performed of the safety evaluation pcrformcd for the Unit 2 full core oflload with one train of spent--'uel cooling.This safety review covered the Unit 2 refueling outage schedule which included a dual train CCW outage.1'ootnote 3 of Table 9.5-2 reprcscnts the normal design of thc SFP cooling system, that is, the SFP cooling system is designed to rcmove the.heat generated by stored spent fuel elements in tlic[SFP].The system incorporates two separate trains.The safety review for the Unit 2 full.core oflload xvith one train of spent fuel cooling addressed thc FSAR section 9.4 attribute of the SFP, cooling dealing with time to boil events and bulk pool tcmpcraturc requirements; however, the safety review failed to address ESAR section 9.5 attributes associated CCW cooling rcquiremcnts as given in Table 9.5-2.t'his issue impacts both units.However, the units are not currently in a Tcchnical Specification limiting condition for operation as a result of this issue., POWER REACTOR EVENT NUMBER: 32824 FAILURE TO PERFORM A 10 CFR 50.59 EVALUATION FOR A PROCEDURE CHANGE INVOLVING COMPONENT COOLING WATER HEAT.EXCHANGER OUTLET TEMPERATURE LIMITS~During the ongoing NRC architect/engineer'esign inspection, a question was asked relative to thc fact that during thc last Unit 2 refueling outage;an adininistrative limit of 90'F was placed on the component cooling water (CCW)system.The thermal analysis indicated that a maxiinum CCW tempcraturc of 90'F would eliminate all margin associated with thc spent fuel pool (SFP)design assuming a design flow of 3,000.gpm.

+w't't tl l I t Attachment 2 (continued),,".NRC Daily Event Reports on D C Cook Design Inspection Findings The following inspection question was asked: Since a change in CCW tcmpcraturc was required to meet the Final Safety Analysis Report (FSAR)value of 160'F for the SFP, was a 10 CFR 50.59 unrevicwcd safety evaluation perfoimed?

I'he licensee reviewed tlic cliange to the procedure to limit CCW tcmperaturc to 90'F.The licensee considered this~change to bc an adniinistrative change only to lower the allowable tcmperaturc to the SFP cooling heat exchanger.'

10 CFR 50.59 evaluation was not performed because it was not rccognizcd that the 95'F requirement was essentially being changed.I Without the completion of an unrevicwcd safety question determination, thc plant was i'n a condition outside the" design basis.The units arc not currently in a technical specification limiting condition for operation as a result of this issue.1 I l'condition rcport has been written to initiate actions to investigate this event and provide prcventivc actions.The 90'F limit is no longer in the operating proccdurcs.

'POWER REACTOR EVENT NUMBER: 32839 AVAILABLE WATER VOLUME IN RWST NOT ADEQUATE IN MODES 5 AND 6 During the ongoing NRC architect/engineer design.inspection, NRC inspectors asked a question about thc reactor, coolant makeup rcquircd aAcr a 10CFR50, Appendix R fire.To rcsporid to thc question, the licensee reviewed two associated design calculations.

The more restrictive calculation was determined to bc the calculation of rrccord to mcct the rcquircment.

This calculation.requires 87,000 gallons of avater to bc available in thc refueling water storage tank (RWST).The value of 87,000 gallons was approved,on 02/20/90: During modes 1 through 4, plant proccdurcs adequately ensure that this requirement is met.During modes 5 and 6, plant procedures arc not'dcquatc to ensure that this requirement is met.r I Thc plant has been in modes 5 and 6 many times since this rcquircmcnt became effective on 02/20/90.Based on tlus, the plant has been in an unanalyzed condition several times since 02/20/90.r Currently both units arc in mode 1.The licenscc is reviewing plant operating procedures to determine'impact and'ill implement procedure chances as needed prior to either unit entering modes 5 or 6.The licensee is continuing to evaluate the subject calculations and plans to submit a liccnsce event rcport to the NRC on tins subject.POWER REACTOR EVENT NUMBER:,32843 LAKE MICHIGAN TEMPERATURE EXCEEDED PLANT DESIGN BASIS LIMIT IN AUGUST 1988'As a result of questions posed by members of the ongoing NRC architect/enginccr design inspection team, the licensee has'dctcrmined that the water temperature of Lake Michigan, thc plant's ultimate heat sink, cxcccded the plant design basis lake temperature limit of 76'F for 22 days during August 1988.I 0~f l I 1 Attachment 2 (continued)

NRC Daily Event Reports on D C Cook Design Inspection Findings The control room is normally cooled by an air conditioning system which utilizes non-safety related clullers.The safety related portion of the control room air conditioning system utilizes water from Lake Michigan as the cooling medium.This water would bc supplied directly to the cooling coils following manual realignment.

At an average lake temperature of 81'F that existed during the 22 day period in>ugust 1988, the temperature inside'he control room could have reached 110.4'F had the non-safety related chillers not functioned.

At a temperature of 110.4'F, the lifetime of some instrumentation inside the control room, the solid state protection,system, and the nuclear instrumentation, is estimated to be at 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> or 6.25 days.The impact of this shortened instrument life span on plant operation had not been cvaluatcd.

/At the time of this event, the plant Technical Specifications allowed continuous operation with control room tcmpcraturcs up to 120'F.The Technical Specifications have since been rcviscd such that continued operation with control room temperatures in cxccss of 95'F is not permitted.

Operation of thc plant during thc time period when lake temperature exceeded the design basis limit, without analysis indicating acceptable control room cooling could bc maintained above this temperature limit,'and without procedures to alert personnel of the situation, is considered as operation in an unanalyzed condition.

Thc instrumentation was not adversely impacted by thc lugh lake tempcraturcs as the non-safety related chillcrs continued to function and maintain acceptable control, room temperatures.

POWER REACTOR EVENT NUMBER: 32915 OVERPRESSURE PROTECTION OF THE COMPONENT COOLING WATER SYSTEM PIPING NOT IN ACCORDANCE WITH THE ANSI CODE REQUIREMENTS i Chapter 9.5 of the FSAR sta'tes: Thc relief valve on the component[cooling water]surge tank is sized to relieve the maximum flow rate, of water that would enter thc surge tank following a rupture of a reactor coolant thermal bamer cooling coil.The set prcssure assures that thc design pressure of thc component cooling system is not exceeded.'he piping design code at thc Cook plant is B31.1.B31.1 states that an intercepting stop valve cannot be located between the source of pressure and the prcssure relief dcvicc credited for protecting the pipe.In this instance, the prcssure source is the ruptured thermal barrier, the prcssure relief device is a safety relief valve on the surge tank.Contrary to the code requirement, thcrc are manual valves maintained open behveen the two.These valves were not controlled in accordance with or exempted from B31.1, An evaluation is being performed to determine the most cQcctive method of establishing and maintaining thc code'cquircmcnt.

A condition report has been written to initiate an investigation into this event and determine the appropriate preventative actions." Tlus condition was identified in response to'an ongoing NRC architect/engineer design inspection.

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