ML20087N874

Forwards Comments on Draft Order for Intergranular Stress Corrosion Cracking Insps

ML20087N874
Person / Time
Site:	Dresden, Browns Ferry, Pilgrim, Brunswick, Quad Cities, 05000000
Issue date:	08/15/1983
From:	Weeks J BROOKHAVEN NATIONAL LABORATORY
To:	Liaw B NRC
Shared Package
ML20083L924	List: ML20085H770 ML20085L392 ML20086F743 ML20087N832 ML20087N874
References
FOIA-84-15 NUDOCS 8404050119
Download: ML20087N874 (10)

Text

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BROOKHAVEN NATIONAL LABORATORY ASSOCIATED UNIVERSITIES, INC.

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j Upton Long Island. New York 11973 2617

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(516) 282s Depc'iment ct Hue.< tar Energy FTS 666' August 15, 1983

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Dr. B. D. Liaw 3'

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U.S. Nuclear Regulatcry Comission Mail Stop P-302 Washington. DC 20555 Subiect: Conssents on Dreft Order for IGSCC Insoections i

Dear Dr. Liaws

.The draft order you sent me this morning appears to rely, heavily on the ability of the utility to detect leaks from ICSCC in BWR pipes.

In view of the inconsistencies from the EPRI Round ;Eobin test of ultrasonic detection of these cracks, I believe the safety assessment has to rely heavily on the " leak before break" hypothesis, which to date has always been reliable.

In my opinion, the plan'cs could probably rur safely until their., next _ scheduled refuelling outage, given the increased requirements onesionitoring for unidentified leakage,* and intermittant visual inspect! ions for leakage, as spelled out in this draf t order. As you know, we will be lookir.g' carefully at the laak before break model in our reviews for your branch over the next few mont'hA.

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While one might argue that, a speedup in inspection schedule adds little to safety since the NDE techniques (in the hands of some _ inspectors) are imprecise in determining the depth of a given crack;, these inspections do appear reliable in detecting at least the pres'ence of.cracksc Therefore, the proposed ' inspections will provide an additional' margin of safety.

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JRW:ob John R. Weeks, Head cc:

H. J. C. Kouts-Materials Technology Division W. Y. Kato P. Grande M. Reich

• (I presume'dhe ~2 gal, increase in a 4 hr. period meant 2 gym increase in a i

.24 hr. period.)

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1 7590-01 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Matter of

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Docket No.

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LICENSEE NAME

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(PLANT NAME)

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ORDER DIRECTING LICENSEE'S TO CONDUCT IGSCC INSPECTIONS I.

(LICENSEE'S NAME

,the licensee), is the holder of Facility Operating License No. DPR-, which authorize the licensee to operate PLANT NAME (the facility), at power levels not in excess of megawatts thermal-(ratedpower).

The facility is a boiling water reactor located at the licensee' site in LOCATION II.

As a result of inspections conducted at 18 operating Boiling Water Reactors (BWRs) in conformance to recent IE Bulletins (IE Bulletin No. 82-03, Revision 1, " Stress Corrosion Cracking in Thick-Wall, Large-Diameter, Stainless Steel, Recirculation Syste Piping at BWR Plants," and IE Bulletin No. 83-02,

" Stress Corrosion Cracking in large-Diameter Stainless Steel Recirculation.

System Piping at BWR Plants"), extensive intergranular stress corrosion cracking n

l-(IGSCC) in primary system piping has been discovered.

These bulletins request'ed selected licensees to perform'a number of actions regarding inspection and l'

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testing of pipe welds.

Inspections conducted highlighted the IGSCC problem at several operating reactors.

Results of these and other inspections pursuant to'IE Bulletins 82-03 and 83-02 have revealed extensive cracking in large-diameter recirculation and residual heat removal system piping.

For these plants, repairs, analysis, and addi-tional surveillance conditions were required.

It has been concluded that other uninspected BWR facilties are likely to have similar IGSCC, which may be unacceptable for continued safe operation without additional surveillance requirements, repair and/or replacement of the affected pipes.

Therefore, the :chedule for the IGSCC inspection for the remaining uninspected plants is being modified.

The following licensees and plants are affected:

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Commonwealth Edison Company Quad Cities Unit 25a@ 65 Dresden Un;t 3 M9 Browns Ferry)Nnit 3 J-9Ce Tennessee Valley Authority BrunswickUnit2 3RT Carolina Power and Light Company Boston Edison Company Pilgrim Unit 1 19S By letter dated July 21, 1983, the staff, pursuant to 10 CFR 50.54(f),

requested the licensee to provide a justification for continued operation of the facility prior to conipleting the inspections of IE Bulletin'83-02.

The licensee responded by letter dated 1983.

The licensee'also attended a public meeting held in Bethesda, Maryland on August

, 1983.

In-the cor--

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3 respondence amd meetings the following issues have been discussed with the licensee:

(1) augmented leakage monitoring program; (2) a shutdown at least once every six weeks for a visual leak examination; (3) inform the reactor operators of the concerns about pipe cracks and the greater' potential need to implement emergency LOCA procedures; and (4) have the Plant Operating Review Committee review the results of any visual or UT inspections, or leakage records on a weekly basis.

In view of the previously observed cracking, balanced by the compensatory measures proposed by the licensee until the inspection program is completed, I have determined that the public health and safety and interest requires the licensee to be ordered to implement selected measures until the inspections are completed by an immediately effective Order.

III.

Accordingly, pursuant to sections 103, 1611, 1610, 182 and 186 of the Atomic Energy Act of 1954, as amended, and the Commission's regulations in 10.CFR Parts 2 and 50, IT IS HEREBY ORDERED THAT:

A.

Effective immediately, the licensee shall initiate a shutdown of the-facility to conduct UT examinations of reactor coolant system piping by no later than (Quad Cities-2, 9/4/83; Dresden 3,' 9/30/83;-

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, Brunswick 2, 11/1/83; Browns Ferry 3, 11/11/,83; Pilgrim 1,12/1/83)-

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4 B.

The facility shall remain shutdown until the Director, Office of Nuclear Reactor Regulation, finds in writing that the licensee has satisfactorily completed the following actions or has provided adequate justification for not requiring a given action.

1.

The licensee shall conduct an ultrasonic examination of the recirculation and residual heat removal system welds:

a.

100% of the welds in recirculation system piping including a

the large diameter piping.

b.

100% of the welds of the jet pump inlet riser piping and associated safe-ends.

c.

All sweepolet-to-header (manifold) welds of jet pump risers.

nearest the end caps, if applicable to the design.

d.

All welds in the ASME Code Class.1 portion of residual heat removal piping system.

2.

The ultrasonic inspection and evaluation of welds shall be in accordance with that specified in IE Bulletin 83-02.

3.

Inspection reporting requirements shall.be.in accordance with

[j those specified in IE Bulletin 83-02.

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4.

The teams conducting these inspections, if qualified, must be requalified under IEB 83-02 irrespective of the note of page 4 of IEB 83-02.

5.

The depth sizing capability of inspection teams should be demon-strated factoring into consideration the recent EPRI round robin test experience, including the three advanced systems.

t 6.

In lieu of the scope of inspections rehired by item III.B.1 above, the licensee may conduct inspections and evaluations as requested in the IEB 83-02, and in addi ion, conduct an ultra-sonic examination of six or 100*4 of the welds (if there are fewer than 6) in the ASME Code Class 1 portion of'the RHR system using an advanced flaw sizing technique, demonstrated, to size within 50% of actual flaw size.

7.

' The licensee shall provide a report of the results'of the Item III.B.1 inspection and any corrective actions (in the event cracking is identified). This report should also include the-susceptability matrix for welds examined (e.g., stress rule index, carbon content, high weld examined for the RHR system);

The written report shall be submitted to the Director, Office of Nuclear Reactor Regulation, Washington, D. C.

20555, under oath or affirmation, under provisions of Section 182a, Atomic Energy Act of 1954, as amended.

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C.

Effective immediately, during che interim period prior to the conduct of inspections discussed in III.B above, the following systems shall be operable (as defined in the generic letter dated April 10,1980):

ReactorCoreIsolat{onCooling(RCIC)-1 system 1.

2.

High Pressure Coolant Injection (HPCI) - 1 system 3.

Low Pressure Coolant Injection (LPCI) - 2 subsystems 4.

Low Pressure Core Spray (LPCS) - 2 subsystems 5.

Residual Heat Removal.(RHR) - 2 subsystems caoable o being utilized in at least both the core shutdown cooling mode and in the injection mode.

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Component Cooling Water System - 2 systems 7.

Service Water System - 2 systems 8.

All unit diesel generators With one system / subsystem of any of the above system (or with one diesel generator) inoperable, restore the inoperable system / subsystem to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or place the unit in at le~ast hot shutdown (as defined in the GE-STS) 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 (as defined in the GE-STS) within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

D.

Effective immediately, during the interim' period prior to the conduct.

of inspections discussed in III.B above, the current reactor coolant system leakage limits, the reactor coolant system leakage shall be limited to a 2 gpm increase in unidentified leakage within any four

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7 hour8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> period. With this leakage limit being exteeded, the unit shall be placed 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 />.

E.

Effectiveimmediately,duringtheinterimperiodpriortotheconduct of inspections discussed in III.B above, the following reactor coolant system leakage detection system shall be operable:

1.

At least one primary containment atmosphere gaseous radioactivity

monitor, 2.

At least one primary containment atmosphere particulate radio-i activitiy monitor, and 3.

At least one primary containment sump collection and flow monitoring system.

With the primary containment sump collection and flow monitoring system inoperable, restore the inoperable system to operable-status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-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 />. With one primary containment gaseous or particulate radioactivity monitor inoperable, (grab samples of the containment atmosphere shall be obtained and analyzed at least once per eight hours), and restore the inoperable monitor to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 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 shutdwon.

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

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8 F.

Effective immediately, during the interim period prior to the conduct of inspections discussed in III.B above, a visual examination of the reactor coolant piping shall be performed during each plant outage anticipated to be 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or more, but no less frequencly than once every six weeks. The examination shall be performed in-accordance with IWA-5241 and IWA 5242 of the 1980 Edition of Section XI of the ASME Boiler and Pressure Vessel Code.

G.

Effective upon receipt of this Order the licensee shall train the reactor operators so that they are aware of the consequence of the pipe crack problem and to drill them in use of the appropriate LOCA emergency procedures.

H.

The Plant Operating Review Committee shall review the results of.all visual and UT inspections performed in accordance with this Order and review leakage to containment records on a weekly basis.

I.

The 0irector, Office of Nuclear Reactor Regulation, may relax.or rescind any of the above conditions for good cause shown by the licensee, such as undue hardship due to the unavailability of.

qualified inspection personnel.

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IV.

The licensee may request a hearing on this Order within 20~ days of the date of g

~ publication of:this Order in the Federal Register. A request for a hearing i:

7590-01

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9 shall be addressed to the Director, Office of Nuclear Reactor Regulation, U. S. Nuclear Regulatory Commission, Washington, D. C.

20555. A copy shall also be sent to the Executive Legal Director at the same address. A REQUEST FOR HEARING SHALL NOT STAY THE IMMEDIATE EFFECTIVENESS OF THIS ORDER.

If a hearing is requested by the licensee, the Commission will issue an Order designating the time and place of any such hearing.

If a hearing is held concerning this Order, the issue to be considered at the hearing shall be whether the licensee should comply with the requirements set forth in Section III of this Order. This Order is effective upon issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Harold R. Denton, Director Office of Nuclear Reactor Regulation Dated at Bethesda, Maryland this day of 4

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