ML17202U799

Forwards Safety Evaluations & Technical Evaluation Repts Accepting Licensee 880729,1005,1221 & 890301 & 0721 Responses to Generic Ltr 88-01 Re IGSCC in BWR Austenitic Stainless Steel Piping W/Listed Exceptions

ML17202U799
Person / Time
Site:	Dresden
Issue date:	08/23/1990
From:	Siegel B Office of Nuclear Reactor Regulation
To:	Kovach T COMMONWEALTH EDISON CO.
Shared Package
ML17202U800	List: ML17202U799 ML17202U801 ML17202U803 ML17252A595 ML20058K072
References
GL-88-01, GL-88-1, TAC-69132, TAC-69133, NUDOCS 9008290177
Download: ML17202U799 (8)

Text

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Docket Nos. 50-237 and 50-249

• August 23, 1990

• D ISTR!BUTION:

. (DQ~kef.. J'Jles---JNRC & Local PDRs

-PDIII-2 r/f DCrutchfield JZwolinski RBarrett BSiegel CMoore OGC-WFl EJordan Mr. Thomas J. Kovach Nuclea~Licensing Manager Commonwealth Edison Company-Suite OPUS West III 300 ACRS(lO)

PDIII-2 Gray Files TBoyce 1400 OPUS Place Downers Grove, Illinois 60515

Dear Mr~ Kovach:

. SUBJEtT:

REVIEW OF RESPONSE TO GENERIC LETTER 88-01, "NRC POSITION ON IGSCC. IN BWR AUSTENITIC STAINLESS STEEL PIPING," DRESDEN NUCLEAR POWER STATlON, UNITS 2 AND 3 (TAC NOS~ 69132 AND 69133)

The NRC staff has completed its review of your July 29, 1988, October 5, 1988, December 21, 1988, March 1, 1989 and July 21, 1989, submittals 1n respo*nse to Generic Letter (GL) 88-01 for Dresden Units 2 an 3. *A copy of the staff's Safety Evaluations (SE) a~d.a copy of the Technical Evaluation Repor._ts prepa_red by the staff's contractor, Viking Systems International, for each unit, are enclosed.

Based on the Safety Evaluations, the staff finds* your responses to GL 88-01 are acceptable with some exceptions.

You are requested to review these exceptions and n'qtify the staff of actions taken to address 'them.

When the,Committee to Review Generic Requirements (.CRGR) met on this topic, they approved Technical Specifications (TS) to ~e incorporated by licensees.

Thus, you are requested to,propose amendments to your TS for each unit, except as noted, to:~

(l,)

(2)

• ( 3)

(4)

Include a.statement in the surveillance or administrative controls section that includes th9° following:

"The inservice inspection program for piping identified in NRC Generic Letter 88-01 shall be performed in accordance with the NRC staff positions on schedule, methods, personnel, and sample expansion included in Generic Letter 88-01 or in accordance wrth alternate measures approved by the NRC stafL" Include an additional Limiting Condition for Operation (LCO) that specifies reactor coolant system leakage shall be limited to a two gpm

• increase in unidentified leakage within any 24-hour period.

Include a surveillance requirement that primary containment sump flow rate will be monitored at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. This is now considered acceptable.

It should be noted that the position in GL 88-01 on leak - *

. rate monit.oring was modified to permit leakage measurements base~J QQ sump flow instruments t9 be taken every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> instead of every *11.* "~;_qurs.

Include, for Dresden Unit 2, an* LCO on operability of sump monitoring instruments.

Since your plant has Intergranular Stress Corrosion

~

crac;~i-n g-:-:-:-~:-.:-:-c-'-'~;;; ;~r:....,:.,-,~-:~E _w~~~, 1 ~men ts, Gener1 c Letter 88-0l f O I.

PDR ADOCK 050002::::7*

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Mr. Thomas August 23,"!]990 hydrogen injection rate might not be large enough to effectively mitigate the IGSCC.

However, this explanation requires further testing and confirmation.

To ensure adequate inspection of IGSCC susceptible piping welds, the staff has determined that, any future request for. the reduction in the scope or frequency of IGSCC inspection will not be granted until our concern of the effectiveness of HWC in mitigating the IGSCC is completely resolved.

As a result of these findings the staff's position, as stated in ~he July 31, 1989 SE, is still valid and no credit to reduce the number of inspections based on HWC will be granted.

Since you previously submitted integrated IGSCC inspection plans, our safety evaluation precludes the necessity for you to submit IGSCC inspection* plans for.each future outage.

However, if flaws are found that do.not meet the criteria.of Section XI of the ASME Code for continued operation without evaluation, NRC a~froval of flaw evaluations or repairs in accordance with IWB 3640 and-n:lB 30 is required before resumption of operation.

Note that weld overlay repair is. considered as a non-C~de repa1r, which requires NRC approva 1.

Enclosures:

Sincerely, Original Signed By!

Byron L. Siegel, Project Manager Project Directorate III-2 Division of Reactor Projects - III,

' IV, V and Special Projects Office of Nuclear Reactor Regulation

1.

Examples of Acceptable Technical Specifications

2. Safety Evaluation for Dresden Unit 2 3~ Technical Evaluation Report for Dresden Unit 2 4 *. Safety Evaluation for Dresden Unit 3

5. Technical Evaluation Repoit for

• Dresden Unit 3 Office:

Surname:

Date:

PM/PDIII-~ PM/POI~

BSiegel/tg"'"/ TBoyce7'/f 0

~I 'VV/90 8' /1-2../90

. UNITED STATES NU.CLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555.

August 23, 1990 Docket Nos. 50-237 and 50-249*

Mr. Thomas J. Kovach Nuclear Licensing Manager

.commonwealth Edison Company-Suite 300 OPUS West III 1400 OPUS Place

.. Downers Grove, Illinois 60515

Dear Mr. Kovach:

SUB.JECT:. REVIEW OF RESPONSE TO.GENERIC LETTER 88-01,. "NRC POSITION ON IGSCC 'IN BWR AUSTENITIC STAINLESS STEEL PIPING," DRESDEN NUCLEAR POWER STATION? UNITS 2 AND 3 (TAC NOS. 69132 AND 69133)

The ~RC stiff has completed its review o~ your July 29~ 1988~ October 5, 1988,.December 21, 1988, March 1, 1989 a~d July 21, 1989, submittals in

• • response to Generic Letter (GU.88-01 for Dresden Units 2 an 3. A copy of the st~ff's Safety Evaluations.(SE) and a copy of the Technical Evaluatfon

• • Reports prepared by *the staff's contractor, Viking. Systems International,

• f.or each unit, are enclosed.

Based on the Safety Evaluations, the staff finds your responses to GL 88-0l

~re acceptable with some exceptions~ *You are requested to review these exceptions and notify the staff of actions ta.keri to address them.

When the Committee to Review Generic Requirements (CRGR) met on this topic, they approved Technical Specifications (.TS) to be incorporatec:t by licensees.*

Thus,.y9u are requested to propose amendments to your TS for each unit,.

• ~kcept as noted, to~

• l

~

(1)

Include a statement in the surveillance or administrative contra ls*

• section that' includes the following:

"The inservi ce inspection.

progr~m for piping identified in NRC Generic Letter 88-01 shall be performed.in accordance with the NRC staff positions on *schedule, methods~ pers~nnel, and sample expansion included in Generic Letter 88-01.or in accordance with. alternate measures.approved by the NRC staff!."

• (2)

• Ir:itlude *an addi.tiona*l Limiting Condition for Op.eration (LCO) that

. specifies reactor* coolant system.leakage shall be limited to a two gpm *

(3)

(4) increase in unidentified leakage "!.ithin any 24-tiour period.

Include a surveillance requirement that pr.imary containment sump flow rate.will be monitor~d at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. This is now considered

.acceptable. It should be noted that the position in GL 88-01 on leak

• rate monitoring was modified to *permit leakage me*asurements based on

• sump flow *instruments to be taken every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> instead of every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

Include, for Dresden Unit 2, an LCO on operability of sump monitoring instruments. Since your plant* has Intergranular Stress Corrosion Cracking (IGSCC) weld Category E weldments, Generic. Letter 88-01

Mr. Thomas August 23, 1990 provides an allowed outage time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for repairing the drywell floor drain sump monitoring system, or an orderly shutdown should be i_nitiat~d. As an* alternative, the staff recommends that when the drywell sump monitoring system 'is inoperable, the operator.

should use a demonstrated manual method for determining leak rate, such as measuring the time to manually pump the sump at a fixed interval (4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />).

The staff considers manual measurement a viable sump. monitoring method without hardship to the operator; therefore, this method could be added to the appropriate LCO section.

With the manual method operable, the outage time for the

.drywe 11 sump monitoring system could be extended to 30 days.

However, if the sump pump and drywell sump monitoring systems ar.e inoperable concurrently, then either system has to b*e repaired wi thi_n 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> o.r an orderly shutdown* should be. initiated.

Pertinent sections of the model BWR Technical Specifications are enclosed as

• example~ of acceptable positions for Items 2 and. 3 above.

Accordingly, please submit the requisite TS change requests or notify the staff of actions. taken within 60 days from receipt of this* letter

• Commonwealth Edison Company a 1 so needs *to address the Reactor Water Clean-up

. (RWCU) piping outboard of the isolation valves in the [nservice Inspection*

{ISI) Program.

If the piping is within the scope of GL. 88-01, you wi 11 need to modify your ISI program to.include the identity of the welds as well as

• plans for mitigation *and t-nspections. A minimum of 10% of the RWCU system piping outboard of the isolation valves *should be inspected at each refueling outage. If cracks are found, the licensee should discuss sample expansion and mjtigation methods with the ~RC staff~

In your July 29,*. 1988 submittal in respon.se to GL 88-01 you*requested a* 50%

reduction of the number of inspections of certain welds based on the use of hydrogen water chemistry (HWC) for Dresden 2. Viking Systems foternational, our contractor, statedthat the NRC staff has to make the determination of whether or not fully effective HWC is b.eing maintained at Dresden 2. The.

staff, in its Safety Evaluation (SE)* of IGSCC 'inspection and repairs for ~he tycle 11, Dresden 2 ~~fueling outage, transmitted a. letter to CECo dated July 31, 1989, stating the following:

As a result of Dresden Unit 2's successful i~plementation of the HWC for three consecutive fuel cyc.les, the staff ha~ approved CECo's request for a factor of two reduction in the inspection of Categories C, D and E weldments in the refueling outage.

However, in view of the extensive IGSCC found* in this outage, the staff has generic concerns regarding the effectiveness bf HWC in mitigating the.IGSCC.

The staff notes that the.HWC implemented ii1 this. Unit is neither monitored by electro-chemical potentia 1 (EPC) measurements nor confirmed by on-1 ine crack arrest verification* ( CAV) te-sting. Therefore, one *possible explanation for the reported inspection results *is that ttie

.e Mr. Thomas August 23, 1990 hydrogen injection rate might not be large enough to effectively mitigate the IGSCC.

However, this explanation requires further testing and confirmation.

To ensure adequate inspection of IGSCC susceptible piping welds, the staff has determined that, any future request for the reduction in the scope or frequency of IGSCC inspection will not be granted until our concern of the effectiveness of HWC in mitigating the IGSCC is completely resolved.

As a* result of these findings the staff's position, as stated in the July 31, 1989 SE,. is stil 1 va 1 id and no -.credit to reduce the_ number of inspections

. based on Hwc*will be granted.'

Since you previously submitted integrated IGSCG. inspection *plans, our safety evaluation precludes the necessity for you to submit IGSCC inspection plans for each future outage. However, if flaws are found that do not meet the.

criteria of*Section XI of the ASME Code for continued operation without evaluati.on~ NRC aaproval of f~aw evaluations or ~epairs *in ac~.ordance with IWB 3640 and IWB 130 is required before resumpt10n of operation. Note that.

weld overlay repair is considered as a non-Code repaTr, which requires NRC approva*l.

Enclosures:

SincereJy,

~-.. PUoi Manager oject Direct ate III-2 Division of Reactor Projects - ir1, IV, V and Special Projects Office of Nuclear Reac~or Regulation

1.

• Examples of Acceptable Technical Specifications *

2. Safety Evaluation for Dresden Unit 2

• 3. Technical Evaluation Report for Dresden Unit 2 4.. Safety' Evaluation for Dresden Unit 3 5~ - Technical Evaluation Report for Dresden Unit 3 cc.w/e.nclosures:

See next.page

Mr. Thomas J. Kovach Commonwealth Edison Company cc:

Michael I. Miller, Esq.

Sidley and Austin One First National Plaza Chicago, Illinois 60690 Mr. J. Eenigenburg Plant Superintendent Dresden ~uclear Power Station Rural Route #1 Morris, Illinois 60450

..

• U. S. Nuclear.Regulatory Commission.

Resident Inspectors Office

• Dresden Station Rural Route #1 Morris, *Illinois 60450 Chairman Board of Supervisors of

• Grundy County

• Grundy County Courthouse Dresden Nuclear Power Station Unit Nos. 2 and 3 MorriS; Illinoi-s 60450 Regional ~dministrato~

Nuclear Regulatory Commission, Region III 799 Roosevelt ~oad, Bldg. #4 Glen.Ellyn, Illinois 60137 Illi.nois Department of Nuclear Safety Office of Nuclear Facility Safety 1035 Outer Park Drive Springfield, Illinois 62704 Robert Neumann*

Office of Public Counsel State of illinois Center 100 W *. Rando'lph

• suite 11-300 Chicago, Illinois. 60601 0

ENCLOSURE 1 MODEL BWR TECHNICAL SPECIFICATIONS REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE

• LIMITING CONDITION FOR OPERATION 3.4.3.2 Reactor coolant sy.stem leakage shall be limited to:

• a.

No PRESSURE BOUNDARY* LEAKAGE.

b.

5 gpm UNIDENTIFIED LEAKAGE.

c.

25 gpni tota 1 *leakage averaged.over any 24-hour period.

d.

1 gpm leakage from any reactor coolant system pressure isolation valve specified in Table 3.4.3.2-1.

{e. 2 gpm increase.in UNIDENTIFIED;LEAKAGE within any 24-hour. period.)

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2 and* 3.

• .ACTION:

a.

With any PRESSURE. BOUNDARY LEAKAGE, be in at least HOT SHUTDOWN

~ithin 12 ~ours and in COLD SHUTDOWN within_~he next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b.

. With any reactor coolant system leakage greater than the Hrrtits i!l b and/or c, above, reduce *the leakage rate td within the limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> ri~ be in at.least HOT SHUTDOWN WITHIN THE NtXT 12 hgtirs and in COLD',SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

c.

With any reactor. coolant system pressure isolation..valve leakage greater th.an the above limit, i-solate the high. pressure portion of the affected system from the low pressure portion within 4 **

hours by use of at least. two.closed manual *or deactivated automatic valves, or be in at )east 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 /> and 1n COL~ SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> *.

d; Wit~ any reactor coolant systein leakage greater than. the llmit in.

. e above,.. identify the source of leakage within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or 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 /> and in COLD SHUTDOWN

,. within.the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.)

SURVEILLANCE REQUIREMENTS 4.4.3.2.l The reactor coolant system leakage shall be demonstrated to be within_ eath of the above limits by:.

a.

b.

Monitoring the primary containment atmo.spheric particulate

{and/or gaseous} radioactivity at least once per (4 or 12 as

• applicable to plant) hours, Monitoring the primary containment sump flow rate at least :Once per eight (8) hours, 3f 4 4-7

\\

. J REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)

c.. Monitoring the prima.ry containment air coolers condensate flow rate at least once per (4 or 12 as applicable to plant) hou~s, and d.. Monitoring the reactor vessel head flange leak detection system at least once P.er ~4 hours.

4.4.3.2.l Each reactor coolant system_pressure isolation valve specified in Table 3.4~3.2-1 shall be demonstrated OPERABLE pursuant to Specification 4.0.5, except that in lieu of any leakage testing required by Specification 4.0.5, each valve shall be demonstrated OPERABLE by verifying leakage to be within *its limit:*..

• a.

At least once.per 18 *mo'iiths.

b.

Prior to entering HOT SHUTDOWN whenever th~ plant has been in COLD SHUTDOWN for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or mor.e ~nd if *leakage testing has not been perfo~med in the ~revious 9 months.

c.

Prior to retu~ning the*valv~*to ~ervice foliowing m~intena~ie, repair or replacement work on the valve.

d~ *. Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following valve actuation due to a1,Jtomatic or.

manual. actio~ or flow thidugh the.~alve.

~

'~

.*