ML20217C614

Discusses Mod of Tech Spec Requirements Re BWR Scram Accumulators for Facility.Change Re CRD Pump Operability Acceptable as line-item Tech Spec Improvement

ML20217C614
Person / Time
Site:	LaSalle
Issue date:	07/02/1991
From:	Barrett R Office of Nuclear Reactor Regulation
To:	Kovach T COMMONWEALTH EDISON CO.
References
NUDOCS 9107170306
Download: ML20217C614 (4)

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Docket Nos. 50-373 and 50-374 Mr. Thomas J. Kovach Nuclear Licensing Manager Commonwealth Edison Company-Suite 300 OPUS West 111 1400 OPUS Place Downers Grove, Illinois 60515

SUBJECT:

MODIFICATION OF TECHNICAL SPECIFICATION REQUIREMENTS RELATED TO BWR SCRAM ACCUMULATORS FOR LASALLE, UNITS 1 AND 2 The NRC staff recently approved a request to remove from the Technical Specifications (TS) a surveillance requirement for testing scram accumulator check valves of the control rod drive (CRD).

Philadelphia Electric Company proposed this change for the Limerick Generating Station, Units 1 and 2.

Du-ing its evaluation of the proposed change, the staff co..cluded that the removal of this TS surveillance requirement would be an acceptable line-item TS improvement for the eight other boiling water reactor (BWR) plants that were licensed between November 12, 1982 and August 8, 1985 with this requirement. This change does not alter the in-service testing requirements of Section XI of the ASME Boiler and Pressure Vessel code for these valves.

The staff has also concluded that the Limerick TS change modifying the action requirements related to the CRD pump operability is acceptable as a line-item TS improvement. This change relaxed the action requirements to allow 20 minutes to permit a CRD pump to be restored to service.

A copy of the TS change request for Limerick and the staff's safety evaluation are provided as Enclosures 1 and 2, respectively, to facilitate your preparation of a license amendment request to implement these TS changes as applicable to LaSalle, Units I and 2.

However, any action on these line-item TS improvements is voluntary.

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Mr. Thomas J. Kovach July 2,1991 You are encouraged to propose changes to the LaSalle TS that are consistent with the enclosed guidance. The NRC project manager will expeditiously review a proposed license amendment conforming to this guidance.

Please contact the project manager if you have any questions on this matter.

Sincerely, afiY Ri ar

. Barrett, Director Proje Directorate 111/2 Division of Reactor Projects - Ill/IV/V Office of Nuclear Reactor Regulation

':n:losures:

1.

Copy of Technical Specification Change Request for Limerick Station 2.

NRC staff Safety Evaluation for Limerick Technical Specification Change cc w/ enclosures:

See next page i

i 1

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Mr.-Thomas J. Kovach-July 2,;1.991 You are encouraged to propose changes to the LaSalle TS that are consistent with the enclosed guidance. The NRC project manager will expeditiously review a proposed license amendment conforming-to this guidance.

Please contact the project manager if you_have any questions on this matter.

Sincerely, Original Signad By:

Richard J. Barrett, Director Project Directorate III/2 Division of Reactor _ Projects - III/I'./V-

=0ffice of Nuvlear Reactor Regulation

Enclosures:

1.

Copy of Technical Specification Change Request for Limerick Station -

2.

NRC staff Safety Evaluation for Limerick Technical Specification Change cc w/ enclosures:

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f Mr. Thomas J. Kovach LaSalle County Station Commonwealth Edison Company Unit Nos. I and 2 cc:

Phillip P. Steptoe, Esquire Robert Cushing Sidley and Austin Chief, Public Utilities Division One First National Plaza lilinois Attorney General's Office Chicago, Illinois 60603 100 West Randolph Street Chicago, Illinois 60601 Assistant Attorney General 100 West Randolph Street Michael 1. Miller, Esq.

Suite 12 Sidley and Austin Chicago, Illinois 60601 One First National Plaza Chicago, Illinois 60690 Resident inspector /LaSalle, NPS U.S. Nuclear Regulatory Commission Rural Route No. 1 P. O. Box 224 Marseilles, Illinois 61341 Chairman LaSalle County Board of Supervisors LaSalle County Courthouse Ottawa, Illinois 61350 Attorney General 500 South 2nd Straet l

Springfield, Illinois 62701 Chairman Illinois Commerce Commission Leland Buildirig 527 East Capitol Avenue l

Springfield, Illinois 62706 Illinois Department of Nuclear Safety Office of Nuclear Facility Safetv 1035 Outer Park Drive Springfield, Illinois 62704 Regional Administrator, Region 111 U. S. Nuclear TFe~g0Tatory Commission 799 Roosevelt Road, Bldg. #4

- Glen Ellyn, Illinois 60137 '

Robert Neumann Office of Public Counsel State of-Illinois Center 100 W. Randolph Suite 11-300 Chicago, Illinois 60601

10 CFR 50.90 EMC.LoSURE I PHILADELPHIA ELECTRIC COMPANY NUCLEAR GROUP HEADQUARTERS 955 65 CHESTERBROOK BLVD.

WAYNE PA 19087 5691 (sist 6as sooo November 17, 1989 Docket Nos. 50-352 50-353 License Nos. NPF-39 NPF-85 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk i

Washington, D. C. 20555

SUBJECT:

Limerick Generating Station, Units 1 and 2 Technical Specifications Change Request

Dear Sir:

Philadelphia Electric Company hereby submits Technical Specifications Change Request No. 89-12, in accordance with 10 CFR l

50.90, requesting an amendment to the Technical Specifications-(TS)

(Appendix A) of Operating License Nos. NPF-39 and NPF-85.

Information supporting this Change Request, is contained in l to this letter, and the proposed replacement pages are contained in Attachment 2.

This submittal requests changes to TS Section-3/4.1.3, Limiting Condition for Operation (LCO) 3.1.3.5 and Surveillance Requirement (SR) 4.1.3.5, for the control rod sram accumulators.

A misapplication of the requirements-of TS Section 3/4.1.3 resulted in our request for a temporary waiver of compliance dated June 9, 1989 and an emergency TS Change Request dated June 10, 1989.

The NRC letter approving the temporary waiver of compliance, dated June 9, 1989, allove3' continued operation until the emergency-TS Change request was approved.

As committed to in our reemest for the emergency TS Change, dated June 10, 1989, we.re proposing TS changes to provide a final resolution of this issue.'

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The changes proposed herein were modeled after the corresponding TS approved for use at Hope Creek Generating Station, License No. NPF-57.

l If you have any questions regarding this matter, please contact us.

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very truly yours, bW

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. A. Hunger,-Jr.

Director Licensing Section I

Nuclear Services Department Attachmentt i

cc:

W. T. Russell, Administrator, Region I, USNRC T. J. Kenny, USNRC Senior Resident Inspector, LGS T. M. Gerusky, Director, PA Bureau of Radiological Protection 4

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TECHNICAL SPECIFICATIONS CHANGE RIQUEST NO 8942 COMMONWEALTH OF PENNSYLVANIA :

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COUNTY OF CHESTER i-L D. R. Helwig, being first duly sworn, deposes and says:

That he is Vice President of-Philadelphia Electric Company; the Applicant herein; that he has read the foregoing Application for Amendment of facility Operating Licenses to modify the control rod scram accumalator Techneial Specifications, and knows-the contents thereof; and that the statements and matters' set forth therein are true and correct to the best of his knowledge, information and belief.

g Vice Pres den-Subscrioed anc sworn to beforemethis/7c/

L ay of $ M 1989.

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LIMERICK GENERATING STAT!oN i

Docket Nos.

50-352 50-353 License Nos. NPT-39 NPT-85 TECHh'ICAL SPECITICATIONS CRANGE REQUEST

" Control Red Scram Accumulator Technical Specific tions Changes" t

a A$. 74-/2-Supporting Information fe; Changes - 8 pages

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s Docket No. 50-352 50-353 Philadelphia Electric Company, Licensec under Facility Operating Licenses NPT-39 and RPT-85 for Limerick Denerating Station (LGS) Unit 1 and Unit 2 re6pectively, hereby requests that the Technical Specifications operating Licenses be amen (TS) contained in Appendir A of the ded as proposed herein to modify T8 Section 3/4.1.3, Limiting Condition for Operation (LCO) 3.1.3.5 and Surveillance haquirement (SR) 4.1.3.5.

The proposed changes are 1-10 for both LCS Unit 1 and Unit 2 TS, and are contained ini.

Philadelphia Electric Company (PEUo proposed herein to be effective upon issuance o)f the Amendments. requests the cha TS changes, a safety assessment of the proposed changes, infor supporting a finding of No Significant Bazards Consideration, and information supporting an Environmental Assessment.

Discussion of Changes:

On June 9, License, with Unit 1 at 50% power, we discovered that seventeen Unit 4.1.3.5.b.2 during a leak test performed on May 9, 1989.1 control satisfy the criteria of SR 4el.3.5.b.2, which requires a leak test Failure to of each accumulator check valve, renders the associated control rods inoperable and thus requires a plant shutdown.

I was in a condition where the accumulators-were determined to notBowever, since U be necessary to comply with design bases scram requirements, the WRC issued a temporar and subsequently,y waiver of compliance allowing continued operation applicability of the affected SR.an emergency TS change was issued revising the the emergency TS change dated June As committed in our request for 10, 1989, following TS changes as an appropriate final resolution of thiswe are proposing the izsue.

AI)

Remove SR 4.1.3.5.b.2 (and the associated footnote) which requires CRD scram accumulator check valve testing once p.e(.18 months and specifies test acceptance criteria.

B)

Modify LCO 3.1.3.5.a.2.a to allow the reactor operator twenty (20) minutes to restart a tripped CRD pump provided that reactor pressure is greater than or equal to 900 psig.

the operator will immediately place the reactor modeIf reactor pr switch in the Shutdown position.

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Docket No.56-352

$0-353 In addition, since we are proposing changes to the seras accumulator TS, we are also using this Change Request to propose a change to the scram accumulator low nitrogen pressure alara setpoint.

A number of TS violations have occurred at operating nuclear power plants due to setpoint drift of the nitrogen 1

accumulator pressure sensors.

As a result, General Electric Service Information Letter (SIL) 429, Revision 1, "SCU Accumulator Pressure Switches," issued January 18, 1988,- recommends lowering the low nitrogen pressure alarm setpoint of the scraa accumulators to equal to or greater than 940 psig.

This recommendation is intended to maintain the validit ofaTSviolationwhy-ofthealarasetpointwhilereducingtherisk

.ch could occur-due to setpint drift.

We have determined that GE SIL 429, Rev. 1 is applicable to LCS, although we have not experienced a TS violation due to setpoint drift of the pressure sensor for the low nitrogen pressure alara.

To avoid such a TS violation, we are proposing a TS change consistent with the intent of the GE SIL.-

Bowever, the change we are conservative than the change recommended by GE in proposing is more that the-proposed alarm setpoint is equal-to or greater than 955 psig, t

(C)

Change the 18 month scram accumulator pressure sensor channel calibration (setpoint), SR 4.1.3.5.b.1.b. from "970 plus or minus 15 psig" to " equal to or greater than 955 psig".

Safety Assessment The changes proposed in (B) above ensure the scram capability of all control rods.

Control' rod scram accumulators and accumulator check valves are required to support the scram function at reactor pressure less than 600 psig.

At reactor pressures above-600 psig, *< actor pressure alone is sufficient to scram the control rods.

The proposed TS changes require an immediate shutdown-if reactor pressure is less than 900 psig.- Therefore, the scram capability of all control rods is ensured.

When reactor pressure is greater than 900 psig, the allowance of 20 minutes to restart a CRD parsp provides plant staff a reasonable time to restore pump operation.

The additional tco Action i.e., shutdown if reactor pressure $s less than 900 prig) propose (d provides for prompt operator action to prevent reactor operation in a condition where the accumurafErs are required to support the scram function.

Removal of SR 4.1.3.5.b.2,-proposed change (A) above,

-j does not compromise proper testing and maintenance of the scram-

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accumulator check valves since operability is ensured by TS SR 4.0.5 which requires inservice testing (IST) in accordance with section XI of the ASKE Boiler and Pressure Vessel Code.

The IST program requires a reverse flow surveillance test of the scram accumulator l l

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Docket No. 50-352 50-353 check valves once Mr calendar quarter if the plant is in Cold Shutdown.

This will ensure check valve testing at least during every Refueling Outage (i.e., 18 months).

To verify that the check valves close, the charging water header is depressurized and the accumulator pressure and low pressure alarms are monitored.

tested, maintained as needed, installed properly and functiontesting This correctly following maintenance.

Therefore, proper testing and operation of the scram accumulator check valves will continue to be ensured by TS required IST program.

In addition, the current TS SR for the 18 month be required in the TS as the scram accumulators are not required safely shut down the plant during normal reactor operation.

failed accumulator check valve would allow the accumulator pressure A

to bleed down on that one drive only if the CRD pump was tripped.

An operable CRD pump maintains charging water pressure at a higher pressure than the accumulator nitrogen pressure for all CRDs.

a loss of the CRD pump and failure of a scram accumulator checkGiven valve, the accumulator pressure would bleed down and not be able to assist that particular drive in a scram condition.

Bowever, reactor pressure in excess of 600 psig is sufficient to fully insert a contro; rod with a failed check valve.

At 600 psig reactor

pressure, the scram insertion time of an individual control rod with zero accumulator pressure would be within TS and design basis requirements.

continue meet design requirements.Also, the average scram time for all drives would Therefore, failure of an accumulator or accumulator check valve is not significant with respect to the abilit operating conditions.y to shut down the plant during normal The recommended TS change described in part (C) above would change the low nitrogen pressure alarm setpoint of the scram accumulator to " equal to or greater than 955 psig."

justifiable since the proposed setpoint maintains the same minimumThis change is alarm pressure currently allowed by TS.

An upper limit is not necessary since, setting the low alarm setpoint psig provides a more conservative setting (i.e., greater than 955 will result in earlier detection of decreasing pressure).

is in accordance with the GE recommendation (SIL 429 Rev. 1) toThis proposed TS chang provide for'ad'e'quate instrument drift allowance ta violations of the TS while maintaining sufficien.

avoid possible for required scram performance.

itrogen pressure The proposed TS changes for the control rod accumulators were reviewed against the design basis of the LGS Pinal Safety Analysis Report (PSAR).

This review showed that the proposed changes would have no significant effects on system operation,

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j Docket No. 50-352 50-353 average scram times, Minimum Critical Power Ratio (MCPR), or the

' Anticipated Transient Without Scram" (ATWS) design bases and assumptions.

1 FSAR Section 15.1.3 ' Pressure Regulator Failure - Open",

l discusses a loss of reactor pressure event.

The analysis does not require the scram accumulators to shut down the plant.

t FSAR Section 4.6.2

• Evaluations of the CRD Systea,"

f discusses control rod operability assuming CRD equi p ent failure such as hydraulic line breaks.

However, the analysla does not i

require the scram accumulators to shut down the plant.

FSAR Section 15.8 " Anticipated Transient Without Scram,"

which discusses mitigation of an ATWS event does not require the scram accumulators to mitigate-the event.

In conclusion, the proposed changes to the TS do not represent a change in the plant or the design bases as described in t

the LGS FSAR.

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In summary, the changes proposed are accepta'ble'because operability of the scram accumulators will continue to be ensured by l

compliance to TS requirements.

The reliability of the control rods to scram is enhanced by the proposed Action statement, and the proposed changes do-not change the design bases as described in the-i FSAR.

Information Supporting a Finding of No Signifleant Bazards consideration We have concluded that the proposed changes to the LGS scram accumulator TS,-do not constitute a Significant Bazards Consideration.

In support of this determination, an evaluation of each of the three standards set forth in 10 CFR 50.92 is provided l

below.

}

l 1.

The proposed changes do not involve a significant-increase in the probability or consoquences of an-accident previously evaluated.

Three changes have been proposed.

3 A)

Remove SR 4.1.3.5.b.2 which requires CRD scram accumulator check valve testing once per 18' months and. specifies test acceptance criteria.

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Docket No. 50-352 50-353 5)

Modify LCO 3.1.3.5.a.2.A to allow the reactor operator twenty (20) ainutes to restart a tripped CRD pump provided that reactor pressure is greater than or equal to 900 psig.

If reactor pressure is less than 900 psig the operator will immediatel position. y place the mode switch in the Shutdown C)

Change the 18 month scram accumulator pressure sensor channel calibration (setpoint), SR 4.1.3.5.b.1.b froa '970 plus or minus 15 psig* to

" equal-to or greater than 955 psig."

The safety function of the scram accumulator is to assist in control rod insertion when reactor pressure alone is insufficient.

The proposed changes do not change the capability of the control rod to perform its safety function and provide proper reactivity insertion wathin the required time.

Removal of the 18 month leak test specified by SR 4.1.3.5.b.2 does not affect the reliability of the check valves since operability of the scram accumulator check valves is assured by TS Section 4.0.5 which requires that-inservice testing of the check-valves comply with the ASKE Code,Section XI.

The proposed additions to the TS LCO action statement 3.1.5.a.2.a described in (8) above impose additional requirements on operations personnsi to prevent plant operation in a condition when the accumulators are required to support-the scram function.

Finally, GE SIL 429 Rev. 1 provides a' recommendation to change the appilcable TS to allow for scram accumulator pressure instrument setpoint drift and thus avoid-an unnecessary TS violation.

The setpoint we have proposed in accordance with this CE SIL is within the currently allowed range but.does not provide an-upper limit.

AnL upper limit is unnecessary since.any pressure alara activation above the minimum setpoint-value is more denservative-than alara actuation at the1miniana setpoint value.

In summary, the proposed will' not affect nor change any plant hardware, plant design or plant systea operation from that already described'in the FSAR.

Therefore the

-proposed changes do not nodify or add any initiating parameters that would significantly increase the,

Docket No. 50-352 50-353 probability or consequences of any accident previously analyzed.

2.

The proposed changes do not create the possibility of r new or different kind of accident from any accident previously evaluated.

As discussed in (1) above, the design bases of the LGS will remain the same.

Therefore, the current PSAR will retain accurate with respect to its discussion of the licensing basis events and its analysis of plant response and consequences.

The proposed changes do not affect any equipment nor do they involve any potential initiating events that would create any new or different kind of accident.

As such, the plant initial conditions utilised for the design basis accident analyses are still valid.

3.

The proposed changes do not involve a significant reduction in a margin of safety.

As discussed in item (1) above, the safety function of the scram accumulator is to assist in control rod insertion when reactor pressure alone is insufficient.

The proposed changes do not change capability of the control red to perform its safety function and provide proper reactivity insertion within the required time.

TS Section 3/4.1.3 requires that any control rod with an inoperable scram accumulator be restored to operable ctatus or be declared as being inoperable and inserted.

This requirement is unchanged by the proposed TS changes.

At normal reactor pressure (i.e., greater than 900 psig) reactor pressure alone is sufficient to scram the control rods.

The proposed TS allow the plant operator 20 minutes to restore a tripped CRD pump if there is more than one inoperable scram accumulator and reactor pressure is equal to or greater than 900 psig.

Control rod scram accumulators and accumulator check valves are required to support the scram function only at reactor

,pr.tssure less than 600 psig.

To prevent approaching the 600 psig 1 ?lt, the proposed TS require plant operators to immedia.

ty scram the reactor if there is tore than one inoperable scram accumulator and there is not a CRD pump operating when reactor pressure is less than 900 psig. - - _ _ - _

ATTACID(Drf 2 4

e LIKERICK GENERATING STATION Docket Nos.

50-352 50-353 License Hos. NPF-39 NPF-85 PROPOSED TECHNICAL SPECIFICATIONS CEANGES-List of Attached Change Pages Unit 1 3/4 1-9 3/4 1-10 Unit 2-3/4 1-9 3/4 1-10 l

I l-

C0t: TROL R0D SCRAM ACCUwyLATOR$

I LIMITING CONDITION FOR OPERATION i

- -i 3.1.3.5 All control rod scram accumulators shall be OPERABLE.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 5*.

ACTION:

a.

In OPERATIONAL CONDITION 1 or 2:

1.

With one control rud scram accumulator inoperable, within 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />s:

1 b)

Restore the inoperable accumulator to OPERABLE status, or a

)

Declare the control rod associated with the inoperable accumulator

'i inoperable.

i Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

2.

With more than one control red scram accumulator inoperable, declare tne associated control rods inoperable and:

a)

If the control rod associated with any inoperable scram accumulator is withdrawn, irrediately verify that at least one j

i control roc drive pump is operating by inserting at least one withdrawn control rod at least one notch.

If no control rod drive i

pump is coerating and.

+

1)

If reactor pressure is y control red drive pump w,900 psig, then restart at least one i

ithin 20 minutes or place the reactor moce switch in the shutdown position, or 2)

If reactor pressure is <900 psig, then place the reactor mode switch in the Shutdown position.

b)

Insert the inoperable control rods and disarm the associated c:mtrol valves either:

~

1)

Electrically, or s

i 2)

Hydraulically by closing the drive water and exhaust water isolation valves.

Otnerwise, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

b.

In OPERAi10NAL CONDITION 5*:

eU With one withdrawn control rod with its associated scram accumulator inoperable, insert the affected control rod and disarm the associated RL j

direct 4cnal control valves within one hour, either:

i a)

Electrically, or k

b)

Hycraulically by closing the drive-water and exhaust water l

c isolation valves.

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2.

i With more than one withdrawn control rod with the associated scram a

accu

• slater inoperable or no control red drive pump operating.

imeciately place the reactor moce switch in the Shutdown position.

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'At least tne accumslater associated with each withdrawn control rod. Not acchcaole to control rocs removeo per Specification 3.9.10.1 or 3.9.10.2.

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REACTIV!TY CON 7AOL SYSTEMS SURVElLLANCE REQUIREMENTS 4.1.3.5 Each control rod scram accumulator shall be determined OPEP.ADLE:

At least once per 7 days by verifying that the indicated pressure is a.

greater than or equal to 955 psig uniess the control rod is inserted and disarmed or scramped.

b.

At least once per 18 months by:

1.

Performance of at a)

CHANNEL FUNCTIONAL TEST of the leak detectors, and b)

CHANNEL CAtlBRATION of the pressure detectors, and verifying an alarm setpoint of equal to or greater than 955 psig on l

decreasing pressure.

LIMERICK - UHlT-1 3/4 1-10

'3,0NTPOL900SCRAWACCUvutaT005 LIMITING CONDITION FOR OPERATION 3.1.3.5 All control rod scram accumulators shall be OPERA 8LE.

APPL}CABILITY: OPERATIONAL COND]TIONS 1, 2, and $8 ACTION:

a.

In OPERATIONAL CONDITION 1 or 2:

1.

With one control rod scram accumulator inoperable, within 8 houtst a)

Restore the inoperable accumulator to OPERABLE status, or b)

Declare the control rod associated with the inoptrable accumulator inoperable.

Otherwise, be in at least HDT SHUTOOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

2.

With more than one control red scram accumulator inoperable, declare the associated Control rods inoperable and:

a)-

If the control red associated with any inoperable scram accumulator is withdrawn, imediately verify that at least one control red drive pump is operating by inserting at least one withdra.n control red at least one notch.

If no control rod drive pump is operating and:

1)

If reactor pressure is 1900 psig, then restart at least one control red drive pump within 20 minutes or place the reactor moce switch in the shutdown position, or 2)

If reactor pressure is <900 psig, then place the reactor mode switch in the Shutdown position.

i b)

Insert the inoperable control rods and disarm the associated control valves either:

2))

Electrically, or 1

Hydraulically by closing the drive water and exhaust water isolation valves.

Otherwise, be in at least HOT SHUT 00WN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, b.

In OPERATIONAL CONDITION $*:

1.

With one withdrawn control rod with its associated scram accumulator inoperable, insert the affected control rod and disam the associated diractjcnal control valves within one hour, either:

a)

Electrically, or b)

Hydraulically by closing the drive water and exhaust water isolation valves.

1 2.

With more than one withdrawn control red with the associated scram accumulator inoperable or no control rod drive pump operating, imediately place the reactor mode switch in the Shutdown position.

'At least the accumulator associated with each withdrawn control red.

Not applicaole to control rods removed per Specification 3.9.10.1 or 3.9.10.2.

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4.1.3.5 Each control rod scram accumulator shall be determined OPERABLE:

At least once per 7 days by verifying that the indicated pressure is a.

greater than or equal to 955 psig unless the control rod is inserted and -

disarned or scrammed.

b.

At least once per 18 months by:

1.

Performance of at a)

CHANNEL FUNCTIONAL TEST of the leak detectors, and b)

CHANNEL CALIBRATION of the pressure detectors, and verifying an alarm setpoint of equal to or greater than 955 psig on decreasing pressure.

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Docket No. 50-352

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50-353 1

The proposed change to the. scram accumulator pressure instrument calibration setpoint is within the currently allowed range and is more conservative than the setpoint recommended by GE.

1 For the reasons stated above the proposed changes do not involve a significant reduction in a margin of safety.

Information Supporting an Environmental Assessment An environmental assessment is not required for the changes proposed by this Change Request because the requested changes conform to the criteria for ' actions eligible for categorical exclusion" as specified in 10 CTR 51.22(c)(9).

The requested changes will have no impact on the environment.

This Change Request does not involve a significant hasards consideration as discussed in the preceding section.

This Change Request does not involve a significant change in the t pes or significant increase in the amounts of any effluents that may be released offsite.

In addition, this Change Request does not involve a significant increase in individual or cumulative occupational radiation exposure.

Conclusion The Plant Operations Review Committee and th's Nuclear Review Board have reviewed these proposed changes to the TS and have concluded that they do not involve an unreviewed safety question,

do not involve a significant hazards consideration, and will not endanger the health and safety of the public.

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NUCLEAR AEGULATORY COMMi&510N j

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i SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULAT.ON SUPPORTING PENDMENT N05. 39 f.ND 6 TO FACILITY OPERATING LICENSE N05. NPF 39 AND NPF-85 i

l PHILADELPH1 A ELECTRIC COMPANY j

LIMERICK GENERAT1HG STATION. UNITS 1 AND 2 D0CKET N05. 50-352 AND 50-353 j

1.0 INTRODUCTION

By letter dated November 17,1989, philadelphia Electric Company (PECo or 4

the licensee) recuested an amendrent to Facility Operating License Nos.

NPF 39 and NPF 85 for the Limerick Generating Station, Units 1 and 2.

t These proposed acendments would change the Technical Specifications (TSS) for Limerick 1 and 2 tot a) remove surveillance requirement (5R) 4.1.3.5.b,2 (and the associated footnote) which-requires Control Rod Drive (CRO) scram i

accumulator check valve testing once per 18 months and specifies test i

acceptance criteria, b) modify Limiting Condition for O 3.1.3.5.a.2.a to allow the reactor operator twenty (20)peration-(LCO) minutes to restart a tripped CRD pump provided that reactor pressure is greater than or equal to 9C0 psig or if reactor pressure is less than 900 psig, the operator will irrediately place the reactor mode switch in the Shutdown position i

and c) change the 18 month scram accumulator pressure sensor channel

-i calibration (setpoint),.5R 4.1.3.5.b.1.b. from '970 plus or minus 15 plig'

{

to " equal to or greater than 955 psig.'

2.0 BACKGROUND

Limerick, Unit I was shutdown for the second refueling outage from January 11 j

19B9 to May 15, 1989. On May 9, 1989, prior to startup, the licensee perferred surveillance tests on the control rod drives as required by the i

i T5s.

One of the TS surveillances-(4.1.3.5.b.2) specifies that at least i

once per 18 months u ch control rod scram accumulator shall be determined i

operablebymeasurIngandrecordingthetimefor-upto10minutesthat j

each individual accumulator check valve F.sintains the associated accumuister pressure above the alsrm set point with no control rod' drive pump operating. During the surveillance tests on May 9.-1989 of the 185'.

CRD accumulator check valves - 17 of the check valves did not maintain -

bydraulic control unit (HCU), accumulator pressure abovt the low pressure alarm setpoint of 970 psig for the test time interval of 10 minutes (LER.-

l 1-89 042). The data acoutred from performing this and previous surveillance tests of the check valves was not used to make any operability judgements,.

i but was used for trending purposet to schedule-preventive maintenance.

During a review of the surveillance test data results on June.8,1989, the i

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2-NRC resident inspector cuestioned the station's interpretation of the T$

recuirement. The inspector's interpretation was that the results ef the surveillance tests should be used to determine the operability of the accumulator for the associated HCU rather 'han trending for maintenance.

(Inspection Reports 50 352/89 10. Section 4.1 and EO 352/8912 with notice ofviolationandPECoresponsecfAugust?9,1989). At the time this was identified. Lircrick At syster pressures a,bove 600 psig, reactor pressure provides adequat energy to insert the control rods without the assistance of the accumu-1stort, so there was no safety issue with respect to the 17 malfunctioning check valves. To resolve the irrnedtate Questien of operabilit a teeporary waiver of ccepliance on June 9, 19E9. On June 10, y, we issued 1989, the licensee requested a change to the TS surveillance requirerent on the accurulator check valves te note that the requirement was only applicable when reacter vessel pressure is at or below 600 psig. This change was approved by Arendnent No. 31 to License No. NPF-39 on July 10,1989. In the application cf June 10, 19E9, the licertee agreed to review all of the TS serveillance test requirerents on the CR0 teram accumulators. The TS changes prcpesed in this subject application of November 17,1989 are the result of the licensee's reassessmert.

3.0 EVALUATION The licensee is proposing three changes to the T$s as described in the introductory para sbmarized below. graph above. Our svaluation of the changes is The Cp0 system is cescribed in Section 4.6.1 of the W erick Final Safety Analysis Report (FSAR). For each control rod, there is a hydraulic control unit (HCU). The NCU package includes two vertical cylinders, a scran water accumlator ard a scram accumulator nitrogen cylinder. The latter is pressuriced from a nitrogen charging header.

As stated in the F$AR, the scram accumlator stores sufficient energy to fully insert _ a control rod at lower reactor pressures. At higher vessel pressures, the accumulator pressure is supplanted by reacter vessel pressure. The accumulater is a hydraulic cylinder with a free-floating pistnn. The pisten separates the water on the top from the nitrogen below.

A check valve in the accumulator charging line is intended to prevent loss of water presture in the accumulator if supply pressure is lost.

The check valve is located above the tun eccumisters as shown in the attached Figure 4.6 8 from the FSAR. The performance of these check valves is the focus of this safety evaluation.

During normal plant operation, one CRD supply pump is operating at all tires and the other pump is maintained in standby. The operating pump rr.intains the required pressure in all 185 control rod scram accu m lators such that the accumulators contain sufficiert stored energy to ensure the complete insertion of all control rods in the required time at any reactor pressure.

However, when reactor pressure is close to, or at full operating pressure, reactor pressure alone will insert the control rods in the required time. The stored energy in the accumulators may assist in

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3 I

f accelerating the control rods initially I

necessary to ensure a successful scram,, but this assistance is not in fact, reactor pressure alone is suf ficient to fully insert all the control rods at a reactor pressure l

as low as 600 psig.

1 At a reactor pressure of less than 600 psig, reactor pressure alone ray l

not be sufficient to fully insert all the control rods in the required tire.

Therefore, the scram accumulators must contain sufficient stored erergy.to ensure a complete scram under these conditions. With a supply pump operating, accumulator pressure is maintained and a successful scram i

is assured. However, assuming that the charging pressure from the supply pump is lost, the 6ccumulators alone rvst retain sufficient energy to f

complete a scram upon demand. The ball check valves in the accumulator j

charging lines will prevent a rapid loss of accumulator pressure when the supply pump is lost, if the balls properly seat.

l

$R 4.1.3.$.b.2, which the licenses is proposing to delete, presently l

recuires measuring and recording the time for up to 10 minetes that each indivicual Cc0 scram accumulator ball. check valve maintains the associated 1

accumulator pressure above the alarm set point with no control rod drive ouep operating.

I As part of our evaluation, the staff has reviewed the requirements on the l

CR0 hydraulic system in the TS for all domestic BWRt. including surveillance requirerents on the pumps, valves end instrumentation. There are 38 t

cperating EWRs in this country,19 of which have ' custom

• T5s and 19 with i

sone version of the 'BWR Standard TSs.'

rene of the older BWRs have a similar requirement in the T5s for testing the check valves. This includes Big Rock Doint Brunswick 1 and 2 Cooper, Dresden 2 and 3. Cuane Arnold, Fitzpatrick, Natch 1 and 2. LaSalle I and 2. Milestone 1. Monticello, Nine Mile Point 1, Oyster Creek Peach i

tettom 2 and 3. pilgrim. Quad Cities 1 and 2, Vermont Yankee, and, Browns i

Ferry 1, 2 and 3.

l In the 16te 1970's, and early 1980s, the staff was giving increased attention to CRD hydraulic systems in NTOL reviews due to cracking i

detected in some CRD return line nozzles (NUREG.0619 'tWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking," November 1980),

the failure of half of the control rods to scram at Browns Ferry,) Unit

("tWR4 cram Discharge System Safety Evaluation.' December 1.1980,

assessment of whether there would be adequate flow if the CR0 hydraulic system was the only available emergency high pressure water source to the core as was the case during part of the Browns Ferry. Unit 1 fire and

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several incidents at operating IWRs during which both the CRD pumps became temporarily disabled. Revision 3 to NUREG-0123 ' Standard Technical-SpecificationsforGeneralElectrictoilingWaterReactors(tWR/5),

issued Tall 1980.-included as Surveillance Requirement 4.1.3.5.b.2:

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' Verifying that the accumulator pressure (and level) remains above the i

i alarmsetpoint(s)forgreaterthanorequalto20minuteswithno control rod drive pump operating.'

I LaSalle. Unit 1 was the first BWk to be licensed after the accident at TM1 2.

As discussed in the staff's SER on LaSalle (NUREG 0519) there were several results from the preeperational testing that focused the staff's attention on the CRD hydraulic system.

For LaSalle. General Electric calculated a flow rate of 180 gpm would be required to keep the t

core covered assuming loss of all other makeup systems to the vessel 40 i

minutes after shutdown. The applicant perfonned a preoperational CR0 l

flow test and the test results indicated that the actual makeup capability was only 128 gpm, which was insufficient to meet the 180 gpm i

minimum recommendation of NUREA 0619.

The staff recommended that the ta5611e 15s include the above_ survet11ance requirement to verify that the accumulator pressure and level would remain above the alarm set points for at least 20 minutes with no CRD pump operation. Preoperational tests i

by the applicant also determined that because of check valve leakage, j

accumulator depressuritation below the alarm set point could occur within i

three minutes. As discussed in Supplement 2 to the staff's SER (55ER 2 to HVsEG 0519) the applicant proposed an alternative to the surveillance i

j requirerent on the accumulator check valves. The applicant proposed i

installation, prior to startup after the first refueling outage, of an f

automatic reactor trip that would scram the control rods in the event of l

low control red drive pump discharge pressure. The trip would be activated during startup and refueling modas only. The staff concluded i

that this proposal was acceptable, since the accumulators are only needed at lower reactor-pressures. However the staff's position was that the surveillance requirement on the accum,ulator check valves should remain in the T5s until the modifications were completed. Thus. LaSalle Unit 1 was the first BWR to include a surveillance requirement on the accumulator check valves. The requirement to test the valves for up to 20 minutes i

was, however, deleted. The same surveillance requirement for testing the I

accumulator check valve was also incorporated in the LaSalle. Unit 2 T5s when it was licensed.

The reactor scram on low CR0 pump discharge pressure modifications were subsequently completed for both LaSalle Units I and 2.

The survetilance requirement on the accumulator check valves i

was deleted from the Unit 2 T5s by Amendment No. 6 to License No. NPF.18 i

on December 17,1984 and was deleted from the Unit 1 TSS by Amendment No.

33 to License No. N7F.11 on February 4,1986.

ThesuiseNancerequirementontheaccumulatorcheckvalveswasincluded in T5s for those IWRs licensed after LaSalle. Unit 1 in 1982 1983. 1984 and 1985. TheplantsincludedSusquehannaUnitIwhichwasIssuedafull power license on November 12, 1982 through Limerick Unit 1 and River Bend, which were-issued full power licenses on August 8, 1985 and November 1985, respectively. For-all of these plants, the T5s required I

holding the pressure above the alarm set point for up to 10 minutes.- The i

10 minutes was the estimated time it would take to startup the standby CRD pump if the operating pump failed.

I l

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As a result of the NTOL review for LaSalle, Unit 1. the staff initiated i

Generic Issue No. 98 fRD Accumulator Check Valve Leakage. The issue was not actively pursued. By remorandum dated August 13,1984 from the Chief, Auxiliary Systems Branch to the Chief, Safety Program Evaluation i

freech, the latter was requested to prioritize the issue. On February 19, l

1985, the Director NRR approved " Dropping' this generic issue. The' i

menorandumandevaluationsupportingthisactionareenclosedtothis safety evaluation. According to the staff who were involved in this t

assessrent, the resolution of generic issue no. 98 was to have been the basis for removing the surveillance requirement on the accumulator check i

valves from the T$s. On this basis, Hope Creek, Perry alid Clinton, which were issued full power licenses on July 25, 1986, November 13, 1986, and April 17, 1987,lves in their TSs.respectively, do not have the surveillance requireme i

1 on the check va The recuirement is in the Limerick.

Unit 2 TSs since the criteria was to have identical T$s for Units 1 and 2 i

insof ar as possible.

Although the enclosed evaluation provides justification for removing the i

surveillance recuirement on the accumulator check valves-from the T$s, the staff has perforced a supplemental evaluation. As a result of these j

evaluations, the staff is proposing that the surveillance requirement be recoved from the other 8 operating BWRs that have this requirement as part of the TS leprovement Program.

Removal of SR 4.1.3.5.b.2 does not eliminate testing and maintenance of the scram accumulator check valves. Surveillance recuirement 4.0.5 in the TSs requires that inservice testing (1$T) of ASME Code Class 1, 2 and 3 puros and valves shall be performed in accordance with Section XI t

of the ASME Boiler and Pressure Code. The IST program requires a reverse i

flow surveillance test of the scram accumulator check valves once per calendar quarter if the plant is in Cold Shutdown. As a minimum this requires testing of the check valves at least during every refueling outage (i.e., 18 months). To verify that the ball check valves will i

properly seat, the supply pump is secured, the charging water header is depressurized and the accumulator pressure and low pressure alarms-are monitored to verify that the valves have closed (are on their seats) on loss of pump flow.. The IST test requires th3t the check valves maintain the pressure above the low pressure alarm setpoint for.about 30 seconds whereas the surveillance requirement being deleted recuired maintaining the pressure for 10 minutes. The IST program demonstrates that'the scram accumuistor check valves are operable and are functioning. The 15T i

prograin is 'the basic requirement that ensures the all valves in safety-related systems (including the scram accumulator check valves) are l

i periodically tested, that identifies tne need for maintenance, that demonstrates that the valves are installed properly and function as l

intended and that requires retesting following mainter.ance. The present surveillance requirement on the. check valves which is being. deleted is in i

addition to the tests performed on the same check valves by the 15T program. Therefore, testing and operation of the scram accumulator check valves will continue to be demonstrated by the TS required IST program.-

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l 6-The only difference between the two tests is that the $R assesses the h

leak. tightness of the accumulator check valves for 10 minutes whereas the IST test interval is about 30 seconds. The bases for the 30 seconds could not be determined.

It was beyond the scope of the evaluation to determine what percentage of check valves in tWRs pass the 15T testing l

but not the 10 minute leak test. While it is suspected that dirt may be one of the reasons sore ball-check valves do not tightly seat, the staff t

did not review maintenance records to assess whether corrosion ard pitting may also be factors, i

4 The CRD hydraulic system is described in Section 4.6.1.2.4 of the FSAR.

A very simplified drawing of the overall system rigure 2-14 from_the i

General Electric EWR Technology Manual, is attached. The CRD hydraulic l

system supplies and controls the pressure and flow to and from the drives through the HCUs. There is a HCU for each of the 185 control rods. Two drive water supply purips pressuri:e the system with water from the condensate i

treatment system and/or condensate storage tank. Normally, only one supply i

pump is operatirg with the other on standby. The supply pump maintains a nominal 1400 to 1500 psig in the charging water header.

(The pressure is monitored in the control room and can range from about 1250 psig to 1510 psig.1 As long as a supply pump is operating, the accumulators are not i

needed even at low reactor pressure, since the pump maintains the pressure l

upstream of the check valves discussed above. Thus, the leak tightness of the C20 scram accumulator check valves is not a safety concern as long as one drive water supply pump is operating.

i As discussed in the FSAR. at system pressures above 600 psig, reactor pressure provides adequate energy to insert the control rods without the assistance of the accumulators. Thus, during normal operation, the leak-tightress of the sciam accumulator check valses is not a concern, since t

the scram accumulators are not necessary to safely shutdown the plant.

Reactor. pressure in excess of 600 psig is-sufficient to fully insert a l

control rod with a failed check valve. At 600 psig reactor pressure, the i

scram insertion time of an individual control rod with zero accumulator L

pressure would be within T$ and design basis requirements. Also, the average scram time for all drives would continue to raet design require-l ments. Therefore, failure of an accumulator or accumulator check valve i

is not significant with respect to the ability to shut down the plant during normal operating conditions, if there were a loss of-reactor pressure, the isolation actuation-instrumentation would initiate a scram-before.the pressure dropped to 600 psig due to M51V closure (756 psig trip 3

setpoint). Below 600 psig, the nitrogen accumulator would provide adequate pressure to scram a' control red even if the charging water pressure was reduced and the check valve did not retain water pressure in

}

the accumulator.

t One of the other postulated' scenarios evaluated by the staff was loss of I

the operating CRD charging pump during startup when the reactor pressure is below 600 psig. As discussed previously, the nitrogen pressure in the accunulator is adequate to scram a control rod even if the check valve is

o l not holding and pressure in the header bleeds down. Below 600 psig, the reactor is critical but all heat is being used to build up pressure. The M51Vs cannot be opened to start warming up the steam lines until pressure i

is above t5e 756 psig trip setpoint.

If a CPD pump were to fail durirg i

startup, the plant would be shutdown to repair it. The check valves wculd not have to retain pressure in the accumulators for any tignificant length of time to make the reactor subtritical.

I With the assistance of the NGC resident inspectors, the Limerick regulatory engineers and the Limerick CAD system engineer, the staff evaluated the t

overall operation, maintenance and testing of the CR0 hydraulic systems at Limerick. One operable CRD water supply pump maintains sufficient charging i

water pressure to scram the control rods under all conditions, irrespective of whether the accumulator check valve functions as intended. Limerick is one of the minority of plants that has TS requirements on the CRD pumps.

The 155 for most BWR$ have no operability or surveillance reouirements on the CR0 pumps. This is the case for Big Rock Point Prowns Ferry 1, 2, and 3, Brunswick 1 and 2. Cooper Dresden 2 and 3, Duane Arnold, Fitzpatrick, Fatch 1 and 2. Millstone 1. MontIcello, Nine Mile Point 1. Oyster Creek.

Peach Bottom 2 and 3. Pilgrim. Quad Cities 1 and 2, and Vermont Yankee.

i The pressure in the charging water header is shown in the control room from pressure indicating switch 46 1N600. The operattrs are alerted if there is a trip of the CR0 pump or charging water low pressure. The r

operators are also alerted if the pressure in any of the accumulators were to drop below the alarm setpoint. This cannot occur as long as the CRD pump is operating and cressure is maintained in the charging water header.

There is a pressure switch in the charging water line to each accumulator downstream of the check valve, (located ad;acent to the instrumentation block at the base of the HCV.

See attached Figure 1 from the 1&C surveillance test procedure.) The alarm setpoint is listed in Section 4.1.3.5.b. of the T5s. One of the proposed changes is to decrease the alarm setpoint from 970 to 955 psig. The functional and calibration requirements on l

these pressure switches are also specified in the same section of the T5t.

l These TS requirements are implemented by Surveillance Test Procedures ST 2-047 400, Rev. 5.

" Control Rod Drive Scram Accumulator Level and Pressure Detector Calibration / Functional Test." If the operating CRD charging pump was lost (which has not occurred at Limerick), there could l

l be a reduction in pressure in the charging water header during the time it takes the operators to manually start the standby pump.- if an accumulator i

check valyA was not fully seated, pressure in the accumulator could bleed down.'As soon as pressure in the accumulator reached 955 psig (in the proposed T5s), this would alarm in the control room. When a HCU accumulator alarm condition occurs (either a low N bottle or a high N water level), the Main Control Room (MCR) reactor o,perators receive a,

flashing accumulator trouble alarm indication on the Full Core Display panelintheMCRforthespecificHCU(panel *0C600). The reactor operator must examine the Full Core Display to identify the specific HCU accumulator that is in alarm. Acy alarmed accumulator trouble alarm on t

l

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the Full Core Display will flash until the reactor operator acknowledges the alarm on a specific accumulator trouble alarm acknowledge button on the reactor console (panel '00603). This alarm condition is accompanied by an audible and flashing annunciator alarm. " Accumulator Trouble.' in the MCR. The Reactor Operator nust acknowledge the alarm on a general annunciator acknowledge button to silence the alann noise and stop the flashing alarm window.

If a second HCU accumulator alarm is received i

after the first alarm is acknowledged, the annunciator re. alarms and the operator must again acknowldge the alarm to silence the alarm noise and stop the flashing alarm window. The second HCV accumulator trouble alarm also flashes on the Full Core Display. This sequence is the same for multiple HCV accumulatir alarms.

Therefore, adequate PCR indication exists for the operator to be alerted to r'ultiple HCV accumulator trouble alarms.

The proposed T5s require that if more than one accumulator is inoperable and if no CRD pump is operating and if reactor pressure is less than 900 plig. the reactor mode switch is to be placed in the shutdown l

position.

L fased on our evaluation. the staff concludes that 1) the IST program will dennstrate that the CR0 accumulator check valves are operable and that they are not leaHng at an excessive rate and 2) the present surveillance recuirement for leak-testing the check valves for up to 10 minutes is not i

t necessary and can be deleted.

The second of the three T5 changes requested by the licensee was to change the IB month scram accumulator pressure seasor channel calibration (setpoint).

So 4.1.3.5.b.1.b. from '970 plus or minus 15 psig' to ' equal to or greater than 955 psig." A number of T5 violations have occurred at operating nuclear power plants due to setpoint drift of the nitrogen secumulator pressure sensors.

As a result. General Electric Service Information Letter (Sill 429. Revision 1 "HCV Accumulator pressure Switches ' issued January 18. 1988 recomends lowering the low nitrogen pressure alarm setpoint of the scram accumulators to equal to or greater than 940 ps

. This recomendation is intended to maintain the validity of the alar:.,etpoint while reducing tha risk of a i

T5 violation which could occur due to setpoint drif t.

General Electric performed a safety assessment to support the $1L and concluded that the slightly lower setpoint still provided adequate notification to the MCR operators of less of pressure in the accumulators. FEco has determined that GE Sit 429. Rev.1 is applicable to timerick, although they have not experienced a TS violation due to setpnint drift of the pressure sensor for the low nitrogen pressure alarm. The Itcensee is proposing a T5 change tensistent with the intent of the GE Sil. However, the change they.

are proposing is more conservative than the change recomended by GE in that the proposed alarm setpoint is equal to or greater than 955 psig.

[

4

=9 I

All BWRs have pressure and level alarms for each accumulator but most EWRs do not list the pressure setpoint limit in the T5s. Of the 14 EWRs other than Limerick that do list the limit in the T$s, 10 have the 940 psig limit recomended by GE. The 955 psig alarn setpoint proposed by PECo is more conservat'ive than that recomended by GE since it will result in earlier detection of decreasing pressure. The pro 7

pressure sensor channel calibration is acceptable. posed change to the i

The third change to the T$$ requested by the licensee was to modify LC0 3.1.3.5.8.2.a to allow the reactor operator twenty (20) minutes to restart a tripped CRD pump provided that reactor pressure is greater than i

or Muni to 900 psig. If reactor pressure is less than 900 psig the operator will imediately place the reactor ecde switch in the Shutdown position. As discussed previously, there are two CR0 pumps, one of which i

is operating and the other of which is on standby. The CR0 pumps are located on the 201' elevation of the turbine building to have a positive i

suctien head from the condensate storage tanks located at ground level, i

The switchover from one pump to another is not automatic. An operator has to reposition various valves and manually start the standby unit. Because of the high reliability of the CR0 pumps in BWPs, an automatic transfer arrangenent has not been considered to be warranted. timerick is one of the minority of BWRs that has an operability requirement for the CRD puGi in the TSt. Twenty-three PWRs have no requirement to have an operable LRD pump, even if a number of accumulators have been determined to be inoperable.

The TS change proposed by the licensee is the same as that approved for Fermi 2. Hope Creek and Perry. As discussed previously, at reactor pressures above 600 psig, system pressure alone is sufficient to scram the control rods. Having an operable CR0 pump to maintain pressure in the charging water header is not a significant safety concern while the plant is operating. The operators in the control room are imediately alerted if the operating CR0 pump trips and/or if there is low pressure in the charging water system. With reactor pressure greater than 900 psig, the allewance of 20 minutes to restart a CR0 pump is not unreasonabic and is the time approved for three other RWRs. The proposal to trip the reactor if pressure is less than 900 psig and if nu CR0 pump is in operation and there is more than one inoperable accumulator is a conservative reaction. The staff finds the proposed changes to the TSs acceptable.

4.0 ENVIRONMENTAL CON 510 ERAT 10N These amendments involve a change to a requirement with respect to the insta34atten or use of a facility component located within the restricted area as defined in 10 CFR part 20 and changes to the surveillance require-The staff has determined that these amendments involve no significant-ments.

increase in the amounts, and no significant change in the types, of any effluents that may be released offsite and that there is no significant increase in individual or cumulative occupational radiation exposure.

The Comission has previously issued a proposed finding that these amendments iuvolve no significant hazards consideration and there has been no public l-

l j,

l coment on such finding. Accordingly, these amendments meet the eli f

criteria for categorical exclusion set forth in 10 CFR $1.22(c)(9). gibility Pursuant to 10 CFR 51.22(b), no environmental impact statement nor environ.

}

mental assessment need be prepared in connection with the issuance of these amendrients.

5.0 CONett!$10N 6

The Comission made a proposed. determination that these amendments involve no significant hazards consideration which was published in the Federal:

Register (54 FR 51258) on December 13, 1989 and consulted with the i

comonwealth of Pennsylvania. No public coments were received and the Comorwealth of Pennsylvania did not have any coments.-

[

t The staff has concluded, based on the considerations discussed above, that:

(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner and

-(2)suchactivitieswillbeconductedincompliancewiththeComIssion's regulations and the issuance of these amendments will:not be inimical to

}

the comon defense and the security nor to he health and safety of the

public, i

i Dated: May 22, 1990-I

-Principal Contributors t

t l

Richard Clark, PM, NRR

-Herb Williams, Reactor Engineer, R1 i

l tarry Scholl, Resident inspector l

Ron Eerit, Generic Issues, RES-j i

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ttEMORANDUM FOR: Rebert M. Bernero. Director Division of Systems Integration FRC:4 Harold R. Denton, Director Office of Nuclear Reactor Regulation 50! JECT:

SCHEDULE FOR RESOLVINC AND COMPLEi!NG GENERIC ISSUE NO. 98 CRD ACCUMULATOR VALVE LEAKAGE This remorandum approves of a priority ranking of ' DROP" for Generte issue 98. "CRD Accumulator Valve Leakage." The evaluation of the subject issue is provided in the enclosure.

In accordance with NRR office Letter No. 40, " Management of Proposed Generic

!ssues." there is no resolution. to this issue to be mnitored by the Generic 1ssue Panagement Control System evaluation will be incorporated (GIMCS). However, the attached prioritization into-NUREG 0933. "Prioritization of Generic Safety Issues." and is being sent to other NRC offices, the ACR$. and PDR for coments on the technical accuracy and completeness of the prioritization evaluation.

Any changes as a result of coments will be coordinated with you.

Should you have any questions pertaining to the contents of this memorandum.

pleasecontactLouisRiani(24563).

/

i Harold R. Denton, Director Office of Nuclear Reactor Regulation

Enclosure:

l Prioritization Evaluation ec: S u next page i

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ENCLOSURE 1 PRIORITIZATION EVALUATION GENERIC !$$UE NO. 98 "CRD ACCUPULATOR VALVE LEAKAGE

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ISSUE 98:

CPD ACCUMULATOR VALVE LEAKAGE DETCRIPTION Historical Backcround:

During 11. review of LaSalle the ASB identified a potential problem which ccult be generic to all BWRs.a. b The problem relates to ability of the

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control rod drive accumulators to retain pressure for a sufficient period of 14me aftet the failure of a control rod drive (CRD) hydraulic pump.

The CRDs are safety-related, as are the accumulators and their associated check valves.

For r3pid reat. tor shutdown, the stored hydraulic pressure in the necumulator, in conjunction with the reactor system pressure, rapidly inserts all the control rods. At reactor pressures below 500 psig the accumulators provide all the motive force to insert the control rods.

Each control rod 4 provided with its own accumulator. With the reactor pressure above 500 p'sig the accumulators provide the initial acceleration force for the control rocs with the majority of the work provided by the reactor pressure.

The technical specifications for RWRs have a CR0 accumulator check valve s

E ags curveiiience M nemenT. wnich is ambiguous and does not have_an action statement for failure to pass the surveillance requirement.'

Safety S':nificance t

The concern of this issue is the potential for the loss of control rod drive hydraulic system pump at a low reactor essel pressure with leakage of multiple chec.k. valves followed by an accident situation that would require a reactor shutdown.

During such an event it-is possible that there would be a faiture to scram the reactor and the Standby Liquid Control System (SLCS) would be required to achieve cold shutdown.

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Possible Solutions Two possible solutions have been identified as follows:

First;[the CRD pumps, asseciated valves, and instrumentation could be made safety-related with the redundant pump automatically starting upon failure of the running The second possible solution would require that with the reactor pump.

pressure less than 500 psig and more than one control rod withdrawn that both CR0 pueps be running.

For those plants requiring manual action to open a step check valve for the redur. dant pump to perform its function, an j

operator must be stationed by t valve, monitor the header pressure, and to operate the valve when the header pressure drops to a predetermined value.

_ PRIORITY DETERMINATION Assumptiens I

For this issue it is assumed thtt operatica below 500 psig will occur only during ascent to power and controlled descent from power operation.

Further it will be assumed that to achieve 500 psig operation during controlled descent from power operation the CRDs will be inserted by the time reactor pressure will'have been reduced to 500 psig. During ascent to power, the time int (rval between going critical and reactor pressure reaching 500 psig is estimated to be usually one hour.

It will also be assumed that a power ascent will occur, on an average, monthly for purposes of this calculation.

The assumption of monthly power ascents will result in conservative calculations since the average number of plant trips is about eight per year, not all of which result in. reactor pressure falling below 500 psig.

For this prioritization it will be assumed that check valve le,akage will reduce the accumulator pressure below the pressure required to insert the control rod in ten minutes.

3 It is assumed that the accident requiring a scram is one that results in the loss of primary system pressure. With system pressure at or below 500 psig the negative reactivity feedback wi11, with decreasing temperature as a result of decreasing pressure, increase reactivity wi,thout control rod insertion.

Thus, only those accident situations in which system pressure is lost and prin.ary coolant temperature decreases requires the insertion of the centrol rods to limit the core reactism, which would be the LOCA events.

Major PRA studies have assumed that a minihium of three adjacent control rods in a BWR must remain withdrawn for the reactor to remain critical.

For this analysis the same assumption will be considered valid.

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FREQUENCY / CONSEQUENCE ESTIKATE The undesired event (U), that of being unable to shutdown the reactor with the reactor protection system in an accident situation due to t)e loss'of I

control rod drive accumulator pressure can be defined as the product of the following probabilities.

They are:

A, the probability that an accident event requiring reactor trip occurs during any one year (1.4E-03). This quantity is based upon the total LOCA initiating event frequency as given in WASH-1400.ss B, the probability' that the reactor vessel pressure is less.than 500 psig with the reactor critical (1.7E-03). This probability is based upon the assumption that 12 ascents to power occur annually; that one l

hour elapses from attaining criticality until the reactor vessel pressure is greater than 500 psig; and that the average operating time per year is 7000 hours0.081 days <br />1.944 hours <br />0.0116 weeks <br />0.00266 months <br />.

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4 C, is the probability the operators fail to scram the reactor within 10 minutes following the failure of the CF.D Pydraulic pump (0.1). This valueisbasedupontheHumanReliabilityHandbook38'nomjnalmodelfor operator error.

D, is the f requency that' the on-line CRD hydraulic pump fails during a one year interval, (0.73 The WASH 1400:s failure rate for pumps was between 3E-06/hr and 3E-04/hr with a median of 3E-05/hr.

Since the CRD hydraulic pump is not a safety related classified component, but is believed to have a quality level above standard off-the-shelf hardware a failure rate value of IE-04 per hour was assigned. As previously stated, an annual operating time of 7,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> was assumed.

E, is the probability that the operators will fail to start the standby CRD hydraulic pump within 10 minutes after the failure of the on-line pump (0.1).

This value is also based on the Human Reliability Handbook 8' nominal model for operator error.

Pump failure to start 3

is negligibly small in comparison.

F, the probability that three adjacent accumulator check valves leak, (0.1). This probability value was chosen with the belief that it conservatively covers common failure causes as well as the multitude of 3 adjacent control rod combinations involving independent failures.

Even with an ambiguous action statement, it is unlikely that a large number of check valves will leak.

G, the probability that the operator failed to follow procedures by pulling a control red adjacent to two other rods which are already.

pulled,(0,1).

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H, the probability that the reactor protection system failed to detect the pulling of the out-of-sequence rod and then failed to initiate a scram signal (0.01).

2, the probability that the loss of CRD hydraulic pressur'e occurs i

before the accident event (0.5).

Hence U=A B

C-D E

F G

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= (1.4E-03)(1.7E-03)(0.1)(0.7)(0.1)(0.1)(0.1)(0.01)(0.5)

= 8.4E-13 per reactor year.

Suberiticality following a LOCA can not usually be maintained by the SLCS, but may be maintained for a tine in some LOCAs.

The ECCS could control some LOCAs even if some of the control rods are not inserted. As a conservative j

assumption no c.redit will be taken for the SLCS and it will be assumed that the accident initiating event and the failure of the reactor protection systes I

will result in a core melt accident..

l As defined in the Grand Gulf RSSMAP study.s4 accident sequences involving l

LOCAs and the reactor protection system were dominated by the category 2

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releases. The whole body man-rem dose obtained by Using the CRAC code #4 assuming an average population density of 340 persons per square alle (which is the mean for U.S. domestic sites) from an exclusion area of a one half mile radius about the r'eactor out to a 50 mile radius about the reactor. A typical midwest meteorology is also assumed.. Based upon these assumptions-the public dose resulting from a BWR category 2 release is 7.1E+06 man-rea.-

i Based upon an average life of 25 years for each BWR the public risk per reactor is 1.5E-04 man-rem.

For the class of all BWRs, 44 reactors, the risk is E.6E-03 iian r'em.

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Cost Estimate The least expensive resolution to this issue involves turning on the standby CRD hydraulic pump and assigning a dedicated operator at the slop check valve control to monitor pressure and to transfer to the standby system if

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the hydraulic pressure drops.. While it is not exactly known 'the nutbar of plants having this configuration, for purposes of the calculation it will be assumed that 25% are so configured.

For each reactor requiring the dedicated operator, assuming 12 power ascents and descents per year at one hour per change, will utilize 24 operator hours per year.

Based upon 1964 dollars and assuming a cost of $52 per operator-hour for 11 reactors the lifetime cost for all BWRs wi11 be $0.3M.

The cost of upgrading the CRD hydraulic system to a safety related quality level system will be much more expensive.

If 0.5 person years of technical experience were required for evaluation of the existing system and no harMre changes were required the cost would be $50,000 per reactor or i

$2.2M for the 44 reactors involved.

Value/Irneact Assessment

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Based upon a reduction in risk of 6.6E-03 man-rem and a cost of $0.3M the (S) score is calculated to be (6.6E-03 man-rem)/(50.3M) 5 2.2E-02 man-rem per $M.

CONCUJSION In general, accident frequencies on the order of 10.s /yr.,.even for a very sptcif te sequence, must be used with caution.

Errors of incompleteness, and overlooked dependencies, as well as other modeling errors, will generally be very large compared to such frequency estimates.

In'this case', a conscientious effort has been made to identify other sequences and eependencies.

Even with a large error, this issue pos~s a very small risk.

e Therefore the issue should be placed in the DROP category.

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References:

Memorandum from O. Part to W. Minners, "Prioritization of Proposed 4

Generic Issue on CR0 Accumulator Check Valve Leakage," dated August 1984*

b Memorandum from C. Thomas to 0. Parr, "CR0 Accumulators-Proposed Improved Technical Specification," dated August 1984.

C NUREG-0123, " Standard Te:hnical Specifications for General Electric Boiling Water Reactors (BWR-5) Rev. 3," U.S. Nuclear Regulatory Commission, Fall ISJ'.

16 WASH-1400 (NUREG-75/014), " Reactor Safety Study, An Assessment of Accident Risks in U.S. Commercial Nuclear Power Plants," U.S. Nuclear e

Regulatory Commission, October 1975.

54 HUREG/CR-1659, " Reactor Safety Study Methodology Applications Program,"

U.S. Nuclear Regulatory Commission, 1981.

64 NUREG/CR-2800, " Guidelines for Nuclear Power Plant Safety Issue Prioritization Information Development," U.S. Nuclear Regulatory Commission, February 1983.

NUREd'[CR-1278,"HandbookforHumanReliabilityAnalysiswithEmphasis 339 on Nuclear Power Plant Applications," U.S. Nuclear Regulatory Commission, February 1983.

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