ML20127G705

Application for Amend to License DPR-22,augmenting Insp of Piping Susceptible to Stress Corrosion Cracking

ML20127G705
Person / Time
Site:	Monticello
Issue date:	01/18/1978
From:	Wachter L NORTHERN STATES POWER CO.
To:
Shared Package
ML20127G692	List: ML20127G688 ML20127G705 ML20127G721
References
NUDOCS 9211170410
Download: ML20127G705 (11)

Text

--___ - _-

8 . ,

UNITED STATES NUCLEAR REGUIATORY COMMISSION .

NORnlERN STATES POWER COMPANY HONTICELLO NUCLPAR GENEPATING PIANT Docket No. 50 263 PIQUEST POR AMENItiDiT TO  :

OPERATING LICENSE NO, DPR- 22 t

+

(License Amendment Request Dated January 18, 1978) i Northern States Power Company, a Minnesota corporation, requests authorization for changes to the Technical Specifications as shown on the attachments labeled Exhibit A and Exhibit B. Exhibit A describes the proposed changes along with reasons for the change. Dr.hibit B is a set of Technical Specification pages incorporating the proposed changes.

This request contains no restricted or other defense infomation.

NORTHERN STATES POWER COMPANY t 3

By C'7 M ;/

FL J Wachter ' ~ ,

Vice Pt esident, Power Production 6 System Operation on this 18th day of January , 1978 , before tre a notary public in and for said County, personally appeared L J Wachter, Vice President, Power Production 6 System Operation, and first being duly-sworn acknowledged that he is authorized to execute this document in behalf of Northern States Power Company, that he knows the contents thereof and that to the best of his knowledge, infomation and belief, the statemento made in it are'true and that it is not interposed for delay, n -

i

}t_,%W,~] V d h / A, w d '

\

.l

[ , % DENISE E. HALVORSON

NOTARY PUOuO

• WINNESOT A ,

L.

HENNEPIN COUNTY Wy Commission Espues Oct 10,16t1 [

^

%:: . : .... :.::e+o::. ..:: :::oo+++++t

. 9211170410 780118 PDR ADOCK 05000263 p PDR-i

_ . . _ , . , , , _ _ , , __ . _ . , , - _ _ , . ~ . . _ - _ .. , . - - _ . - . . , . . . . . _ _ . . , . - _ _ ,_ ,_, _ . _ --,..- _.-.-. O

i t

EX111BZT A i l

110NTICELLO HUCLEAR GENERATING PIANT l DOCKET No. 50-263 LICENSE NO. DPR-22 LICENSE AMENDMENT REQUEST DATED JANUARY 18, 1978 PROPOSED CHANGES TO TECHNICAL SPECITICATIONS Pursuant to 10CFR50.59, the holders of Provisional Operating License DPR-22 hereby propose the following changes to the Appendix A Technical Specifications:

PROPOSED CilANCES A. Section 1. Definitions Add the following new definitions to Section It AB. Pressure Boundary Leakage - Pressure boundary leakage shall be leakage through a non-isolable fault in the reactor coolant system pressure boundary.

AC. Identified Leakage - Identified leakage shall be:

1) Reactor coolant leakage into drywell collection systems, such as pump seal or valve packing leaks, that is captured and conducted to a sump or collecting tank, or

2) Reactor coolant leakage into the drywell atmo-sphere from sources which are specifically located and known not to be Pressure Boundary Leakage and Which do not significantly impair the methods used to detect reactor coolant Icakage.

AD. Unidentified Leakage - Unidentified Icakage shall be all reactor coolant leakage which is not Identified Leakage.

AE. Non-Conforming Linen - Pipe and fitting material, including veld metal, which has not been shown to be highly resistant to oxygen-assisted stress corrosion in the as-installed con-dition. Type 304 stainless steel is non-conforming unless:

1) All piping and welds aFi~in the solution annealed condition, or

2) The component is protected from exposure to reactor coolant by cast or veld overlay austenitic stain-less steel with 5% minimum ferrite or other materials having high resistance to oxygen-assisted stress corrosion.

AF. Service Sensitive Lines - defined as those that have ex-perienced stress corrosion cracking in boiling water reactor service or are particularly susceptible to such cracking because of high stress or because they contain relatively stagnant, intermittent, or low flow coolant.

I .' ,'

i EX111 BIT A i

2-

B. Specification 3.6.D/4.6.D. Coolant Leakage l

! 1. Revise Specification 3.6.D to read:

4 1

D. Coolant Leakage )

1. Any time irradiated fuel is in the reactor I vessel and coolant temperature is above 2120F, reactor coolant system leakage, based on sump monitoring, shall be limited tot

a. 5 spm Unidentifica Leakage

b. 2 gpm increase in Unidentified Leakage

, within any 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> period

c. 20 gpm Identified Leakage

2. With reactor coolant system leakage greater than 3.6.D.1.a or 3.6.D.1.c above, reduce the leakage rate to within acceptable limits within four hours or initiate an orderly shutdown of the reactor and reduce reactor water temperature to less than 212 F within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

3. With an increase in Unidentified Leakage in excess of the rate specified in 3.6.D.I.b, identify the source of increased leakage within four hours or initiate an orderly shutdown of the reactor and reduce reactor water temperature to less than 2120F within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

4 If any Pressure Boundary Leakage is detected when the corrective actions outlined in

3.6.D.2 and 3.6.D.3 above are taken, initiate an orderly shutdown of the reactor and re-duce reactor water temperature to leco - than 2120 F within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

,y.p e -.

t c EXHIBIT A i

2. Revise Specification 4.6.D to read:

D. Coolant Leakage

1. Unidentified and Identified Leakage rates shall be computed at least once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> using primary containment floor and i equipment drain sump monitoring equipment.

2. Primary Containment atmospheric particulate radioactivity shall be monitored at icast once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

3. Drywell pressure and temperature shall be monitored at least once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

3. Revise the 3.6/4.6 Bases to include the above changes.

l Refer to Exhibit B Page 134.

C. Revisions to Inservice Inspection Program

1. Revise pages 189S and 189T of Dxhibit B of Monticello License Amendment Request dated August 30, 1977 to include new specif1-

! cations 4.13.A.2 and 4.13.A 3 as follows:

, 2. For Non-Conforming Lines which are not Service Sensi-tive, 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 examinations reveal no incidence of stress corrosion cracking, the examination schedule may revert to that specified in 4.13.A.I.

! 3. For Non Conforming Lines which are service Sensitive l a. The velds and adjoining areas of bypass piping ,

~

of the discharge valves -in the main recircula-l tion loops, and of the austenitic stainless

! steel reactor core spray piping up to and in-cluding 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 find the piping l

free of unacceptable indications, the examina-tion.may be extended to each 36 month interval, plus or minus 12 months, and may be limited to one bypass pipe run and one' reactor core spray.

pipe run. -

l

!' < t EX111 BIT A i

4, 1

b. If Service sensitive Lines other than those i 4

listed in 4.13.B.3.a above are identified. l l the welds and adjoining areas of this piping l

shalt be subjected to examination at each I reactor refueling outage or at other scheduled or unscheduled plant cold shutdowns on a sampling basis. Successive examinations need

, not be closer than sLx months apart. If un-l

! acceptable flaw indications are detected in any branch run, the remaining branch runs

, with similar functions and configurations

, shall be examined. In the event three suc-l cessive examinations find the piping free of 1 unacceptable indications, the examination

'i schedule may revert to that specified in 4.13.A.1 with the exception that all examina.

tions normally completed over a ten year

] interval shall be completed each 80-month period.

2. Add new pages 189U and 189V to Exhibit B of Monticello License Amendment Request dated August 30, 1977 to revise the Bases

. to include the above changes. Refer to pages 1890 and 189V in l Exhibit B, attached.

! REASON FOR CHANCES These changes are being proposed at the request of the NRC Staff. They are consistent with the recommendations contained in NUREG-0313, July,1977,

" Technical Report on the Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping."

) All of the reactor coolant system pressure boundary piping and fitting material, including weld material, has been reviewed and compared to the material' selection and processing guidelines contained in NUREG-0313.

Materials that do not meet these guidelines are identified in Table 1 (Non-Confoming Lines which are not Service Sensitive) and Tabic 2 (Non-Confom-ing Lines which are Service Sensitive). Tables 1 and 2 contain the following

, infoma tion:

t

a. System identification

] b. Weld identification i c. History of completed examinations l d. Examination sucnary and future examination schedule All Non-Confoming Lines at Monticello consist of pip;ng constructed of

- Type 304 stainless steel. Field welds in these lines were not solution annealed. Service Sensitive Lines at Monticello consist of

a. Core spray lines

b. Recirculation bypass lines-(welds between 304 and 304L stainless steel) l To date, no other lines can be statistically demonstrated to be l Service Sensitive.

--, -. - . . . - a .- - - - . - =--- . .. - - - a

I .

I j .

EX111 BIT A l

l l The proposed changes conform to the requirements of NUREG-0313 with j one exception. Type 316 stainless steel is not considered to be ,

1 Non-Conforming in the mill annealed plus field welded condition.

] General Electric Company records show that no incidents of stress corrosion cracking have occurred in any Type 316 stainless etce.

1 pressure boundary line in which the piping system is in the i metallurgical condition of ' bill anneated + welded." 1145 veld J joints of Type 316 stainless steel have performed satisfactorily in 1 BWR service with no cracking incidents. 305 of these Type 316

, stainless steel veld joints have seen duty in Service Sensitive '

Lines. Laboratory tests by General Electric have confirmed these i obse rva tions .

l Type 316 stainless steel is used at Monticello in the Residual IIcat

, Removal System piping. These lines are listed in Table 3 which

also contains a sunmary of examinations performed on them. No 4 augmented inspection of these lines is planned.

SAFETY EVALUATION a

j '

1 The proposed changes to the technical specifications consist entirely of additions to the limiting conditions for operation and surveillance require- ,

i ments now in effect. The proposed measures to detect the occurrence of

! stress corrosion cracking are positive actions consistent with existing

) technology.

j 1mplementation of these proposed changes will significantly reduce the possibility that stress corrosion cracking will go undetected.

i 1

l a

1 o u, "

n.99 vm,m.--,---q,, y -gy pry 7p,.,,,,-,.n w. , ,,eryt, w y v. y-w, .,-9ap

.,.y. 9,a9.,,7 y*,.g wy s y- -w-m% , 9--- --af-9.--,i-e-.9 -

e----+-4---- --- g---+

l '

j -

c j

• D311 BIT A Table 1 Page 1 of 3 1

NON-CONFORMING LINES WilICll ARE NOT SERVICE SENSITIVE I

i REPORT NO. OF

! SYSTDI WELD COMPLETED EXAMI!MTIONS j IDENIIFICATION IDENTIFICATION 1973 1974 1975 1975 1977 EXAMINATION REQUIRDiENTS SPRING FALL RECIRCU1ATION Shop welds on Recirculation

( SYSTD1 System were solution anneal-

! cd and, therefore, are not '

! eonsidered NON-CONFORMING i WELDS. t I

RECIRCULATION

! "A" A total of three (3) velds

,. (REW13A-28") RCAF-2 SEE TOOT: 0TE (1 ) vill be examined by the end e RCAJ-3 79,73 of the next outage on Recir-I RCAJ-5 culation A and B. If these RCAJ-7 examinations reveal no inci-

! RCAJ-15 dence of stress corrosion l RCAJ-17 cracking, the examination t RCAJ-21 schedule will revert to

! RCAJ-23 normal requirements.

! RCAJ-24 RCAJ-28 f

i RECIRCUIATION "B" (See Above)

(REW13B-28") RCDF-2 SEE FOOTNOTE (1) j RCBJ-3 68,72 i RCBJ-5 RCBJ-6.

! RCBJ-7 i RCBJ-13 RCBJ-15 l RCBJ-19 l RCBJ-21 RCBJ-22

! RCBJ-26 i RCBJ-26 f

RECIRCUIATION One (1) weld will be

. MANIFOLD "A" examined by the end of the (REW32-22") RMAJ-8 45 next outage on either Recir-I RMAJ-15 culation Manifold A or B...

RMAJ-16 If this examination reveals no incidence of stress cor-rosion cracking, the exami-nation schedule vill revert to normal requirements.:

RECIRCUIATION liANIFOLD "B" (PIW32-22") RMBJ-8 PJDJ-15

f. PJBJ-16

l,.

! LTdIBIT A Table 1 I l l' age 2 of 3 )

• 7=

NON-C0hTORMING LINES WilICil ARE NOT SERVICE SENSITIVE REPORT NO. OF l SYSTEll WF.LD C0f1PIETf D MVAffitMTIO"5 IDENTIFICATION IDENTIFICATION 197J 1974 1975 1975 1977 EXAMINATION RtX/UIREliENTS SPRING FALL l RECIRCUIATION Requirements are based on i RISERS all Recirculation Risers.

i

{ "A" (REW32-12") RFAF-2 116 Examinations perfortned 1 RRAJ-3 117 during the first 80 months

RRAJ-5 54 on Exam Category D J. did I j RRAJ-7 55 not reveal incidence of

stress corrosion cracking. i j "B" (REn'22-12") RFliF-2 Therefore, the examination

.' , RRBJ-3 schedule vill revert to )

RRBJ-5 nonnal requirements for this i l

l RRbJ-7 Exmn Category.

I j "C" (REW21-12") RRCF-2 106 Wolds RRDF-2, RRIT-2, RRllF-2 i RRCJ-3 107 and RRRF-2 (Exam Category RRCJ-5 43 B-T) will be examined by the j RRCJ-7 40 cnd of the next outage. If 3 these examinations reveal -

"D" (REW20-12") RRDF-2 77 no incidence of stress cor-I RRDJ-3 84 rosion cracking, the exami-l RRDJ-5 57 nation schedule vill revert j REDJ-7 56 to normal requirements.

j i "E" (REW19-12") RREF-2 204

! RREJ-3 -

203 i

RREJ-5 201 i RREJ-7 162 i

! "F" (REW14-12") RRFF-2 RPJJ -3 j RRFJ-5 j' RFJJ-7

"G" (REW15-12") RRGF-2 112 i RROJ-3 .

108

, RRGJ-5 44 i RRGJ-7 45 "11" (REW16-12") RFdlF-2 RRilJ-3 i RRilJ-5

IJJ1J 2 "J" (REW17-12") RRJF-2 82 RRJJ-3 83 RRJJ-5 51 RRJJ-7 52 "K" (REW16-12") RRRF-2 s

RFJJ-3 -

RRFJ-5 RRKJ-7 114 (1) - Component is protected l from exposure to reactor coolant by austenitic stainicss steci -

vold r: ictal overlay, with 57. minimum ferritt.

- - - . ----....--.-..-._,-,------.--..-_....-~..-:.-

EXIIIBIT A Table 1 i Page 3 of 3 J NON-CohTORMING LINES WilICil ARE NOT SERVICE SENSITIVE j REPORT NO. OF

! SYSTEM WELD COMPLETFD EXAMINATIONS l IDENTIFICATION IDENTIFICATION 1973 1974 1975 1975 1977 EXAMINATION REQUIREMENTS SPRING TALL

{

i j HICH PRESSURE Examinations performed i

COOLANT during the first 80 months INJECTION did not reveal incidence of (PS18-8") PSAT-215 15 stress corrosion cracking.

7 PSAT-2C 16 Therefore, the examination schedule will revert to

}

normal requirements.

REACTOR WATER Examinations performed CLEANUP during the first 80 months l

j (REW3-4") CWAJ-1 58 59 did not reveal incidence of a CWAJ-2A 59 56 strese corrosion cracking.

CWAT-2 57 36, 61 Therefore, the examination i schedule will revert to i j normal requirements. I l CONTROL ROD Control Rod Drive Return l DRIVE RETURN line capped during 1977 l (N0ZZLE N9) CAP TO SAFE-END- 224 outage with 316L Stainless

! Steel.

i i

5 i

i i

i t

i i

r i

l

. . . _ . . , _ . . . . , , ~ . . _ _ . . . . - _ . _ , _ , , _ . , . . _ _ _ . _ _ , . - , . , , , _ . . _ , . . , _ _ _ . _ , _ . _ . _ . _ . _ , . , _ _ . . . , . _ _ . . . , _ . _ _ . , . . , _ , ..,_. ,_.. - _

EXHIBIT A Tabic 2 Page 1 of 1

,9, NON-CONFORMING LIhTS k'llICll ARE SERVICE SENSITIVE REPORT NO. OF SYSTEM kTLD COMPI.ETED EXAMINATIONS IDENTIFICATION IDENTITICATION 1973 1974 1975 1975 1977 EXAMINATION REQUIREMENTS SPRING FALL CORE SPRAY "A" Three or more successive (TW7-8"ET) CSAJ-2A 73 11 129 examinations of these velds CSAJ-3 72 19 138 and adjoining pipe metal CSAJ-4 69 20 124 were performed and were CSAJ-5 64 21 164 found to be free of unac-CSAJ-8 66 22 123 ceptabic indications. Sub-CSAJ-9 75 23 125 sequently, further examina-CSAF-10 95 70 24 131 tions will be extended to CSAF-14 93 74 25 132 cach 36 month interval, CSAJ-16 98 68 26 165 plus or minus 12 months, CSAJ-17 99 71 27 135 and vill be limited to one CSAF-18 76 28 134 core spray line, A or B.

CORE SPRAY "B" (See Above)

(TW11-8"ET) CSBJ-2A 67 1 140 CSBJ-3 '

65 3 141 CSBJ-4 77 4 121 CSBJ-5 83 17 166 CSBJ-6 84 5 122 CSBF-9 81 6 144 CSBF-12 80 7 145 CSBJ-13 82 8 146 CSBJ-14 78 9 147 CSBF-16 79 10 167 RECIRCULATION The Recirculation By-Pass BY-PASS "A" Lines (A and B) were (REW24-4") RSAJ-2 2 replaced with 304L Stain-RBAJ-16 15 less Steci (<.035% Carbon) during the Fall 1975 RECIRCULATION refueling outage. The two BY-PASS "B" welds identified (between (REW25-4") RBBJ-2 18 304L and the original 304)

RBBJ-19 32 on each line, vill be examined during each of the next two outages. If these examinations reveal no incidence of stress corrosion cracking, fur-ther examinations vill be extended to each 36 month interval, plus or minus.

12 months, and vill be limited to one By-Pass line, A or B.

l -

Table 3 I EX111 BIT A i

! 'Page 1 of 1

t I

TYFE 316 STAINLESS STEEL LINES J .

R El'4 No. OF SYSTEll UELD COMPl.E'li o O'M11fMTIONS IDENTIFICATION IDENTIFICATIO!i 1973 1974 1975 1975 1977 EXAM 11;ATION REQUIREMEliTS ,

SPRING FAtl.

i l

RESIDUAL 11 EAT No incidence of stress

! REMOVAL corrosion cracking. No l (PIW10-18"ED) RllAJ-1 168 augmented inspection i RilAJ-2 169 program ; ~ snned ,

i RilAJ-3 170 RilAF-4 63 171 i

j RESIDUAL llEAT No incidence of stress j REMOVAL corrosion cracking. No j (TW20-16"ED) RilBJ-1 173 augmented inspection j lu!BJ-3 174 program planned.

4 RitBF-4 187 i RllBF-20 4 RHBJ-21 58 PJ1BJ-22 59 l RIIBF-24 1

I "

RESIDUAL !! EAT No incidence of stress l RD10 VAL corrosion cracking. No j (W30-16"CD) RilCJ-1 87 182 i

augmented inspection RitCJ-3 86 183 program planned.

, RliCF-4 188 j PJICF-20 87 96 l" Rl!CJ-21 112 RilCJ-22 113 RilCF-23 88 97 i

i h

d

,. , _ , _ . - . . , _ . , - - _ . . _ . _ . _ - , . , . - . _ , _ . . _ , _ . - _ _ _ - . _ . - . ~ , - . . _ . . _ _ , . . , _ . _ _ , . _ - -