ML20127G721

Proposed Tech Specs Re Augmented Insp of Piping Susceptible to Stress Corrosion Cracking

ML20127G721
Person / Time
Site:	Monticello
Issue date:	01/18/1978
From:	NORTHERN STATES POWER CO.
To:
Shared Package
ML20127G692	List: ML20127G688 ML20127G705 ML20127G721
References
NUDOCS 9211170414
Download: ML20127G721 (11)

Text

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i G11IBIT B LICENSE AMENDMEhT REQUEST DATED JAla!ARY 18, 1978 This exhibit consists of the following pages revised to incorporate the proposed Technical Specification changes 5

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Y. Chutdown - The reactor is in a shutdown condition when the renetor mode evitch is in the shutdown mode position and no core alterations are bein6 performd. In this condition, a reactor sc-am is initiated and a nyl block is inserted diractly from the mode switch. Tha scram can be reset I

after a short time delay.

1. -Hot Shutdown means conditions as above with raector coolant te=perature gmater than 2120F.

2. Cold Shutdown menns conditions as above with renetor coolont temperatum equal to or less ,

than 2120F.

Z. Simulnted Automatic Actuntion - Cimilated automatic actuation mans applying a simulated signal to  !

the sensor to actuate the circuit in question.

AA. Transition Boiling - Transition boiling means the boiling regim between nucleate and film boiling, also referred to as partini nucleate boiling. Transition boiling is the regime in which both nuclente and film boiling occur intermittently with neither type baing completely stable.

AB. Pressure Boundnry Inakage - Pressure boundary lenkey shall be leakage through a non-isolable fault in the reactor coolant system pressure boundary.

AC. Identified Ienkage - Identified leakage shall be:

1) Beactor. coolant leakage into dryvell collection systems, such as v2mp seal or valva packing

' leaks, that is captured and conducted to a sump or collecting tank, or

2) Reactor coolant leakage into the dr7vell atmosphere from sources which are specifically located and known not to be Pressure Boundary Icakage and which do not significantly

impair the methods used to detect reactor coolant leakage.

AD. Unidentifim! Ienkage - Unidentified leaka6e shall be all reactor coolant leakve which is not Identified leakage.

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AE. Non-Confoming Lines - Pipe and fitting mterial, including veld etal, which has not been shown to be highly resistnnt to oxygen-nssisted stress cormsloa in the as-installed

, condition. Type 30ft ctainless steel is non-conforming unless:

1) All piping and velds are in the solution annenled condition, or

2) The component is protected from exposure to the reactor coolnnt by cast or veld overlay ~

austenitic stainless steel with 5% minimum ferrite or other mterials having high resistance to oxygen-assisted stress corrosion.

AF. Servlee Sensitive Lines - defined as those that have experienced stress corrosion cracking in boiling vnter reactor service or are particularly susceptible to such crackir;; because of high stress or beesure they contain relatively stagnant, intemittent, or low flow coolant.

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d l Replace the following pages in Ddtibit B of the Monticello License Amend-l ment Request Dated August 30, 1977 with the attached revised pages:

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[ Remove License Amendment Insert Attached B l Request dated August 30, 1977 Page Humber:

Page Numbert

169S 189S i

189T 189T 189U

189V l

3 For convenience in reviewing this material, page 189R from Exhibit B of License Amendment Request dated August 30, 1977 is also attached. No changes to page 189R are proposed.

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30 LD4ITIrr, CONDITIo:G FOR OPERATION h.O SURVEILIARCE FIQUIFIMErd'S j INSERVICE IIGPECTION AND TESTIIU k.13 INSERVICE IIGPECTION AND T1T,TIrr, Applicability: Applicability:

Applies to components which am part of Appli-s to the periodic inspection and the reactor coolant pressure boundary and testing of components which am part of their supports and other safety-related the reactor coolant pressure boundary

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pressure vecsels, piping, pu:::ps, and and their supports and other safety-valves, related pressure vessels, piping, pu=ps, and valves.

Objective: Objective:

To assure the integrity of the reactor To verify the integrity of the reactor coolant pressure boundary and the coolant pressure boundary and the operational readiness of safety-related operational readinecs of safety-pressure vessels, piping, pumps, and related pressure vessels, pipin6, pumps, valves. and valves.

Specification: Specification:

A. Inservice Inspection A. Inservice Incuection

1. To be considered operable, Quality

1. Inservice inspection of Quality Group A, B, and C components shall .

Group A, B, and C components chall "

be perfo d in accordance with catisfy the requirements contained the requiraments for ASME Code Class in Section XI of the ASMU Poller 1, 2, and 3 mn<mts, mspectively, and Pressure Vessel Code and nppli-contained 1: ' action XI of the ACME enble Addenda for continued service Baller anl F msure Vessel Code aal of ASME Code C1nss 1, 2. and 3 'compo.

applicable Addenda as requimd by nents, respectively, except where 10 CFR 50, Section 50 55a(g), except relief has been requested from the where relief has been requested from Commission pursuant to 10 CFR 50, the Commission pursuant to 10 CFR 50, Seetion' 50 55a(g)(6)(1).

Section 50 55a(g)(6)(1),

313/h.13 189R REV

. i h.O SURVEILIANCE REQUIE34EIlrS ,

30 LD:ITING CONDITIO!G FOR OPERATION

2. For Non-Confoming Lines which are not Servica Sensitive, inspections required by 4.13.A.1 during the first 10-year inspection interval shall be completed by the end of the 1978 refueling outage. If these exemirntions reveal _

no incidence of stress corrosion cracking, the examination schedule may revert to that specifiad in 4.13.A.1.

3. For Non-Confoming Lines which am Service Sensitive:

a. The welds and adjoining areas of bypss piping of the discharge valves in the main recirculation loops, and of the austenitic stainless steel reactor core sprsy piping up to and includin6 the second isolation valve, shall be examined at each reactor refueling outage or at other scheduled or unscheduled plant cold shutdowns. Successive examinations need not be closer than six months apart. In the event three successive _

examinations firxl the piping free of unacceptable indications, the examination rny be extended to each 36 nonth interval, plus or minus 12 months, winciding with a refueling outage. and my be limited to one bypass pipe run arxi one ranctor core sprny pipe run.

189S 3 13/h.13 REV

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h.0 SUR7EILIANCE RNJIRDENIS .

30 LIMITIm CONDITICIG FOR OPEEATION

b. If Ser itee Sensitive Lines other than those listel in h.13.B.3.a above are identified, the vr lds and adjoining areas of this pipine, shall be subjected to examination at each reactor refueling outage or at other schadulmi or unscheduled plant cold shutdoi-ms on a sampling basis. Successive examinations naed not ba closer than six months apart. If l unacceptable flaw indications ara detectM in

.any branch run, the remaining branch runs with similar functions and conficumtions shall be examined. In the event three suc-cassive examinatione find the p* ping free of unacceptable indications, the examination schedule may revert to that specified in h.13.A.1 vith the exception that all examina-tions nomally completed over a ten-year interval shall be completed each 80-month period.

B. Inservice Testing of Pumps and Valves B. Inservice Testing of Pumps and Valves

1. To be. considered operable, Quality 1. Inservice testir4 of Quality Group Gmup A, B, and C pumps and valves A, B, and C pu=ps and valves shall be shall satisfy the requiremants con- perfomed in accordanca with the tained in Section XI of the ASME Ibiler requirements for ASME Cole Class 1, and Pressure Vessel Code and appil- 2, and 3 pu=ps and vn3.es, respectively, cable Addenda for operability of contained in Section VI of the ASME ACME Code Class 1, 2, and 3 pumps Boiler and Pmssure Vessel Cole and and valves, respectively, except applicable Addenda as required by where relief has been requested from 10 CFR 50, Section 50.55a(g), except the Commission pursuant to 10 JR 50, whem relief has been raquested fmm Section '50 55a(g)(6)(1). the Co==1ssion pursuant to 10 CFR 50, Section 50 55a(g)(6)(1).

189T 3 13/4.13 RE7

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Bnses 3 13 and h.13: .

'Ibe inservice inspection and testing program conforms to the requiremnts of 10 CFh 50, Section 50 55a(g).

j Where practical, the inspection and testing of components classified into URC Quality Groups A, E, and C will conform to the requirerents for ASE Code Class 1, 2, and 3 corponents contained in Section XI of the ASME Boiler and Pressure Vessel Code.

Using Regulatory Guide 1.26, Bevision 3, " Quality Group Classifications and Standards for Water, Steam, 4

and Padioactive-Waste-Containing Components of Nuclear Power Plants," as a guide, all Itnticello components have been classified into Quality Groups. 'Ihis classification serves as the basis for detemining which ASME Code Class inspection and testing requiremnts are applicable to -

a given component. 10 CFR 50, Section 50 55a(g) requires components which are part of the reactor coolant pressure boundary and their supports to meet the inservice inspection and testing requiramats applicable to ec=ponents classified as ASE Code Class 1. Other safety-related components must meet the inservlee inspection and testing requirements applicable to components classified as ASE Code Class 2 or 3

'Ibe inservice inspection program must be updated at kO month intervals. The progran for testing pumps and valves for operational readiness must be updated every 20 months. A description of the updated programs should be submitted to the N3C for review at least 90 days before the start of each period. A suggested format for this description is contained in Appandix A to reference (1).

The inservice inspection and testing program must, to the extent practical, comply with the requirements in editions and addenda to the ASE Code that are "in effect" no more than six months before the start of the period covered by the updated progmm. The term "in effect" means both having been published by the ASME, and having been referenced in paragrara (b) of 10 CFR 50, Section SO.55a. If a code required inspection or test is impractical, requests for deviations am submitted to the Commission in accordance with 10 CFR 50. Section 50 55a(g)(6)(1).

1 'ibe infomation specified in Appendix B to reference (1) should be submitted for each deviation -

requested. Deviation requests should, if possible, be submitted to the NRC for revi<r at least 90 days terore th- start of each period. Deviations identified durirs; an inspection inriod may be grouped and requested at the end of each calendar quarter. It is expected thM. c sen11 number of devintions will be identified during the inspection pariod, partiedlarily s a first period when new inspection and testire, techniques will be utilized. A requested devintion 3

request may be considered acceptable to the Co= mission until a romal disapproval has been received.

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4 3 13/4.13 BASES 189U EEV

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~ i fees '313 and h.13 ( continued):  !

Small, hairline cracks in austenitic stainless steel piping in LMR facilities has been observed on seveml occasions. Data indicates that. Type 304 austenitic stainless steel piping in -

j the ' reactor. coolant pressum boundary of the boiling water mactor is suscaptible to stress cormsion

• cracking. Such cracking is caused by a. combination of significant amounts of oxygen in the coolant, 7 h_igh stresses, and some sensitization of metal ad,]acent to velds. Cracks have occurred in the heat affected zones ad.jecent to velds, but are not expected to occur outside these areas, provided the pipe mterial is pmperly annealed. Pipe mns containing stagnant or lov velocity fluids have been observed to be more susceptible to stress corrosion cracking than pipes containing a continuously flowing: fluid during plant opemtion. Historically. these emcks have baen identified either by .;

volumatric examination, by leak detection systems, or by visual incpection. Because of the inherent j high material tou6h ness of austenitic stainless steel piping, stress corrosion cracking is unlikely .- -

to cause a rapidly pmpagating failure resulting in a loss of coolant accident.

Although 'the probability that stress cormslon cracks vill propegate far enough to create a si6 nificant  !

safety hazard is slight, the presence of such cracks .is undesirable. The follwing steps have been taken to minimize.this problem:

1. Where prnetical, pipe runs constructed of mterial susceptable to stress corrosion cracking and vhich contained stagnant or lov velocity fluid have been replaced with materials not susceptable to cracking or they have been elimirated.

2. 'Ihe reactor coolant leakage detection technical specifications have been amended to enhance the ability to detect unidentified I leakage that may include thmugh-vall cracks.

3. The pmgram of inservice inspection has been augmented to  ;

a incrence the probability of crack detection in lines susceptable to stress corrosion cracking.

This ' program confoms to the Commission?s guidelines for plants with operating licenses (; rrerance 2).

References:

. 1. Letter fmm D.~ L. Ziemann, Chief, Operating Beactors Branch !2, USNRC, to L. O. Mayer, IGP, dated November 2h,1976.

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2. NURE-0313, " Technical Feport on Ihterial . Selection and Processing Guidelines for EWR Coolant Pmssure Boundary Piping," July,197(.

313/h.13 BASES 189v RE7

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