ML20058K125

Summary of CRGR Meeting 172 on 891025 Re Listed Items, Including Proposed Rev 3 to Reg Guide 1.35 & Std Format for Backfit Discussions to Be Included in Future Generic Ltrs & Bulletins.W/O Stated Encls

ML20058K125
Person / Time
Site:	Hatch
Issue date:	11/24/1989
From:	Jordan E Committee To Review Generic Requirements
To:	Taylor J NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO)
Shared Package
ML20058A334	List: ML20058A331 ML20058A336 ML20058J967 ML20058K007 ML20058K097 ML20058K115 ML20058K125
References
RTR-NUREG-CR-2719, RTR-REGGD-01.035, RTR-REGGD-01.035.01 NUDOCS 8912140177
Download: ML20058K125 (41)

Text

- _ _

s

. flovember 24, 1989 MEMORANDUM FOR:'

James M. Taylor Acting, Executive Director for Operations i

FROM:

Edward L. Jordan,' Chairman-g Committee to Review Generic Requiremente

SUBJECT:

MINUTES OF CRGR MEETING NUMBER 1H

[

i The Committee to Review Generic Requirements (CRGR) met on Wednesday, October 25, 1989 from 1:00-5:30 p.m.

A list of attendees for this meeting is enclosed (Enclosure 1).

The following items were addressed at the meeting:

1.

The Committee reviewed proposed Revision 3 to Reg.: Guide 1.35, '! Inservice Inspection of Ungrouted Tendons in Prestressed Concrete Containments,"

and proposed Reg. Guide 1.35.1, " Determining Prestressing Forces for.

Inspection of Prestressed Concrete Containments." The Committee did not recommend. in favor of issuing the proposed guidance in its current form.

This matter is discussed in Enclosure 2.

2.

The Committee discussed, but did not complete ~their review of, the NUBARG-appeal to the EDO of the staff position that prohibits use of nuclear heat,

{

and requires pressurization using solid water for hydrotesting of BWRs.

The Committee will give further consideration to this item.at the next CRGR 1

meeting.

This matter-is discussed in Enclosure 3.

j i

3.

The Committee discussed a proposed standard format-for backfit discus-sions to be included in future generic letters and bulletins.

Several changes were suggested to further clarify.the intent of the proposed format. This matter is discussed in Enclosure 4.

In accordance witt ie ED0's July 18, 1983 directive concerning " Feedback andr i

Closure of CRGR Reviews," a written resonnse is required from the cognizant '

office to report agreement or disagr m with the CRGR recommendations in these minutes.

Th response, which

• required.within five working days after receipt of these minutes, is to be #,rwarded to the CRGR Chairman and if there is disagreement with CRGR recommendations, to the EDO for decisionmaking.

Questions concerning these meeting minutes should be referred to Jim Conran (492-9855).

Of.cina! Shned by:

j E. L Jman Edward L. Jordan, Chairman Committee to Review Generic Requirements

Enclosures:

cc:

See next page As stated 3

Distribution: (w/o enclosures)

D. Allison (w/ enc.)

Central File M. Lesar CRGR CF-(w/ enc.)-

N. Randall (w/ enc.)

POR (NRC/CRG P. Kadambi (w/ enc.)

CRGR SF (w/ enc.)~

E. Jordan-(w/ enc )

S.;Treby f

M. Taylor (w/ enc.)

T. Marsh (w/ enc.)

.J.' Heltemes (w/ enc.

W. Little R. Bosnak (w/ enc.)

K. Wichman (w/ enc.)

J. Conran (w/ enc.)

g A

l OFC : A

CRGR :.AE0 D

C/ G 00 :

n:

g.___

4, NAME :J ra t.'cg

CJ temes :

dan DATE $ 11//J/89 11AA/R1 $ lg /89 OFFICIAL RECORD COPY

2-cc w/ enclosures:

Commission (5)

SECY.

r J. Lieberman P. Norry M. Malsch Regional Administrators CRGR Members 8

E b

M p

RdNWIS AND TRANSMRTAL SUP f[2$[fd toi genio ges.e.=ner, nom.

oote

. t.

BCS PDR S

1

- a, l

4.

l a,

hsmen Pee Nees ens netwn hansevel Fw cesseense per conversemen hs Asmusstes For Cervesten Presero RepW kneusses tw vo w inmemeten see ese hennent invesusste sinnetwo 4

'^

2^

Jue:Hr meannes This previous Central File material can now be made publicly available.

MJfff/4t file 7ED 70 CMP Mffr/MG Mo. / 72.

i CC (//$7 04/l V) Jf$4/ $AffH, i

PDA

1. S T M i r

)

00 NOT use this form os a RECORD of approvels, concurrences, disposals, eteerences, and simiter estens Fit 004:(Name, org. symbof, Agency / Post)

Room No.-Bids, I

DEMA//S Al/ / Soy Puone No.

2t//Vf

• l-le2 OPTIONAL FORM 41 (Rev. 7-76) m c

31.M6 Qt s.cpo.ut..s.ni.ur me i

0.?'"1 i

9 m y

,.,y.,,

,--s_-,,9--..-

,,,,,..,,----,..--,.-,,,_.,s-,

..--.,._,,--+-.e.

m...

.m-m -

a --------

I L g to r on m 8 l.2 Y 96 M4MAMt fftertD 70 cACA Mitr/MG NO.

l72-To Af MJef ///5t/Cl7 AVJ/44144

/

MEMO fpA

), IA740/l ff9H E. J04/A// M7/D Nf.2YfSi Syd/fC7/ M/4/WilS #F CRCA Mffr/Mc NUMJfA / 72.

/UCdVJ/MC /MAfot49WAIC Amesents IV#KHtvMr 469r MfV/0VslY Af4fASOi 1

CC 11l u ceoS =

e RSl A SWMMMP 0lC D/SCVSS/e#S of A /AtgMste MfV/S/o# 3 7e #lCUL4FMP' rg/et /.33' i

23k.

44D Mo/*SU MSWL4?Ml' ff/M /*ffa/

NNN 0

b, s#ctosynt A $UMM/AY #f ' J/ Scuff /*HC ff M A'444RC A//dAl 79 r/t kJo pHr/f MSL #f,Wre414 A

1. 2Ar FeA #F'#ROMGr/MS AA/A 'S.

C.

BAKleSWni N

A Sq'MMJ4f' #F //JCWSS/0#,5 97 A flAO)VS/D 374WeJ4) AntM4/ f*A A4CMHr b/Sekssteds /q l'AtA Arp/t4/ /ACKActJ

?.

Mmo FoA L.JoAJJ# f,nosf f,JJckJeAb J47tD 7/21l96 FCA&JA//// Ae*V/fk' sws//AML.5 su A PAC /VS4D ALC//LAre4 Y GW/)c /.JS, MAV3....

AMD 44dHl47MY /W/PL /rJC/

l l

i 4

l t

1

i I

ATTENDANCE LIST FOR CRGR MEETING NO. 172 October 25, 1989 i

CRGR MEMBERS i

C. J. Heltem s (Acting Chairman)

J. Sniezek O. Ross J. Moore (for J. Goldberg)

G. Arlotto L. Reyes CRGR STAFF J. Conran D. Allison NRC STAFF R. Bosnak l

H. Graves G. Arndt L. P11sco D. Jeng C. P. Tan K. Wichman R. Hermann B. Elliot i

C. Y Chr l

L. Mars l

M. Tay19.-

G. Mizune l

P. Kadambi B. Hayes N. Randall

+

l

(.'

i to the Minutes of CRGR Meeting No. 172 l

Proposeo Revision 3 to Regulatory Guide 1.35 and Froposed Regulatory Guide 1.35.1 October 25, 1989 TOPIC Bob Bosnak and Herman Graves presented the subject proposed guides for CRGR review. The proposed Revision 3 t, Regulatory Guide 1.35 would implement a number of changes in inservice inspection orograms for ungrouted tendons in pre-stressed concrete containments as currently specified in Revision 2.

It was, with some exceptions, parallel to the recently issued subsection IWL of ASME Section XI. The proposed Regulatory Guide 1.35.1 would provide essentially new guidance on predicting and evaluating pre-stressing forces. A copy of the-slides used in the presentation is attached to this enclosure.

b BACKGROUND i

The package submitted by the staff for CRGR review of this matter was transmitted by a memorandum dated July 28, 1989 frnm E. S. Beckjord to E. L. Jordan. The package included:

1.

The proposed regulatory guides 2.

Public comment letters on the proposed guides 3.

Resolution of public comments on the proposed guides 4.

NUREG/CR-4712 " Regulatory Analysis of Regulatory Guide 1.35, Rev. 3, Draft 2...."

5.

Incident of Tendon Anchor Head Failures at Farley 2 6.

NUREG/CR 2719. " Evaluation of Inservice Inspections of Greased Prestressing Tendons," and 7.

Regulatory Guide cover letter In addition, the CRGR staff requested the following information from RES and i

distributed it to the CRGR:

1.

The current regulatory guide 2.

Subsection IWL of ASME Section XI, 1988 addenda, and 9

l 3.

Sample technical specification (copy attached to this enclosure) 1 i

At the meeting, RES provided a revised regulatory guide cover letter (copy-l.

attached to this enclosure).

l

I g.

i i

CONCLUSIONS / REC 0t#iENDATIONS The Committee did not complete its review of this' matter at this meeting, but did identify a number of questions to be addressed at a future meeting The major points discussed are summarized below:

t 1.

The staff proposed and a coneensus of the CRGR agreed that the new recommendations should be forward fit (strictly voluntary for existing plants). There did not appear to be the substantial safety improvement required to pursue cost-justified backfitting (nor did the matter appear to qualify as a compliance backfit or an adequate protection backfit).

It was noted that the new recommendations could not be strictly, voluntary and yet prohibit licensees from adopting part instead of all of the'new 2

recommendations. Nor could they be voluntary and imply that they will be used to judge future proposed technical specification changes, 2.

The staff agreed to consider whether, instead of issuing the new recommendations as written, it could endorse Subsection IWL with at most a few exceptions.

If so, this approach would be preferable.

In any i

event, the reasons for exceptions or differences with Subsection IWL should be explained.

3.

In endorsing Subsection IWL, priority should be placed on using a rule rather than a regulatory guide (ASME Codes are normally endorsed by rule).

It was noted that such a rule st rid be forward fit just as the regulatory guides would be.

4.

The effect of 10 CFR 50.55a(g) was discussed. This regulation provides for application of the latest version of,ASME Section XI that is endorsed in 10 CFR 50.55a(b) and this process was considered exempt from backfitting considarations.

The RES staff expressed the opinion that such automatic bypassing of backfit considerations would not apply to Subsection IWL because it addressed a new area not addressed in previous versions of Section XI.

However, the RES staff agreeb to check again j

with OGC on this point.

5.

The staff agreed to estimate the time required in proceeding with'the main

~

options discu: sed above:

(1) publication of a rule endorsing Subsection IWL with few exceptions; (2) publication of regulatory guides endorsing j

Subsection IWL with few exceptions; and (3) publication of regulatory guides in essentially their current form.

6.

The staff agreed to discuss the options with the Subsection IWL Chairman and to obtain NRR as well as RES concurrence on any proposals to be made at the next meeting.

7.

If the regulatory guides were to be published in essentially their current 1

form, the CRGR would provide detailed technical comments at the next meeting.

-,.r.

3 y

.,+

u y,m, g

g h

k 1

I 4

F i

1 i

o, CRGR REVIEW OF -

i I

I i

{

REG. GUIDES 1.35 & 1.35.1

.l j

b i

e f

ll i

i i

I i

s i

.s I

6 i

k l<

n o

R I

e 3

i

(

bd 1

I D

a a

t OCTOBER 11,1989 i

C

"E O

l l

r>

t fi k

r i

1 1

l l

L PURPOSE OF REGULATORY GUIDES t

t i

R.G.1.35 PROVIDES GUIDELINES FOR DEVELOPING 4

AND IMPLEMENTATING ISI OF PRESTRESSED CONCRETE CONTAINMENTS WITH UNGROUTED TENDONS j

4

)

R.G.1.35.1 PROVIDES GUIDELINES FOR DETERMINING l

PRESTRESSING FORCES TO BE USED FOR ISI BY CLARIFYING HOW.TO CONSTRUCT TOLERANCE BANDS (UPPER & LOWER BOUNDS OF PRESTRESSING -

FORCES AS A FUNCTION OFTIME)

t.....

---

J

I i

i i

l PRESTRESSED CONCRETE CONTAINMENTS 3

l 4

I

\\

i.

4 ONE TWO THREE TOTAL l

REACTOR REACTOR REACTOR REACTORS i

UNIT UNITS UNITS b

i i

{

i l

11 13 2

43

]

L 4

i i

I

.i 6

i

I n=

1 i

1 l

J l

Y 1 TYPICAL DOME TENDON h

2 -TYPICAL VERTICAL TENDON m

3 -TYPICAL HOOP TENDON Q/

4 - TYPICAL U-SHAPE TENDON Y

[-

t..g t

\\

~ 01 4,

[-

t

t I

i i

WHY PUBLISH GUIDES NOW 1

l l

GUIDES ARE READY TO PUBUSH

-i CLARIFICAT!ON OF ISSUES NOT FULLY ADDRFN IN REV.2 i

WILL PROVIDE CONSISTENCY IN REVIEWS AND A L411 FORM j

STANDARD FOR ASSESSING THE ISI CONDITION OF UNGROUTED TENDONS, AND COULD FORM THE BASIS FOR LATER NRC REGULATION I

t i

SUBSECTION IWL," CONURETE COMPONENTS " OF ASME CODE WAS PUBLISHED DECEMBER 1988. IT WILL TAKE A.2 BOUT 2 YEARS FOR NRC TO ENDORSE IWL IN A REGULATION l

GUIDES ARE NON-MANDATORY, WHEREAS REGULATION ARE l

MANDATORY. THIS WILL PROVIDE A 2 YEAR " TRIAL USE" PERIOD BEFORE THE REGiK.ATION IS STARTED i

t i

i l

i A

~.. - - ~

-. ~ -.

4 I

i 1

L t

f t

i i

PROVISIONS IN REV. 3, R.G.1.35 1

t t

3 1

i l

i.

y SAMPUNG REQUIREMENTS CHANGED FROM EITHER A FDGED l

NUMBER OR A PERCENTAGE FOR EACH TENDON GROUP b

TO ONLY A FIXED PERCENTAGE OF THE POPULAT'Oii OF EACH GROUP

+

l DETENSIONING OF ALL TENDONS 30 SAMPLE CHANGED TO DETENSIONING'OF ONE RANDOMLY SELECTED TENDON t

FROM EACH GROUP I

i i

I

?

l WHEN TWO ESSENTIALLY IDENTICAL CONTAINMENTS EXIST AT A SITE, REY. 3 RECOMMENDS PRESTRESS MONITORING OF SECOND CONTAINMENT WHEREAS REV. 2 RECOMMENDED ONLY A VISUAL INSPECTION s

t l-I t

t j

t:

m.

, - ~...

.. ~.

i a

i I

j PROVISIONS IN REV. 3, R.G.1.35 (CONT.)

i i

+

.l i '

4 i

VISUAL INSPECTION RECOMMENDATIONS IN REV. 3 HAVE BEEN INCREASED DUE TO THE INCIDENCES OF TENDON ANCHOR HEAD FAILURES AT FARLEY NPP j

t l

l iiEV.3 COMPLETELY DELINEATES ITEMS WHICH SHOULD BE

[

i CONSIDERED DURING INSPECTION OF THE SHEATHING FILLER GREASE 1

REPORTABLE CONDITIONS EXISTS WHEN WATER CONTENT CHLORIDES, NITRATES, SULFIDES, GREASE VOIDS, OR RESERVE ALKALINITY EXCEED PRESCRIBED LIMITS l

l l

l l

i

.f i

l

I I

I

- t a

t i

i 4

i e

l lMPLEMENTATION-i i

i i

i l

l i

I ALL NUCLEAR POWER PLANTS WITH PRESTRESSED CONCRETE i

4 j

CONTAINMENTS; 26 PLANTS OR 43 REACTORS 1

l t

NEW GUIDES' PROVISIONS SHOULD BE USED WHEN UTR.ITY PROPOSES OR AMENDS RELEVANTTECHNICAL SPECIFICATIONS 1

i I

I i

I i

l a

w

- - + -

w

+

r

-w

-N.-

-,r,

+

c w

e

.-w

-w.

w-

i i

l t

i I

1 CONCLUSION i

i l

)

i t

I-

"LTHOUGH THERE ARE CERTAIN AREAS WHERE ADDITIONAL COSTS MAY BE INCURRED BY INDUSTRY A SUBSTANTIAL NET i

COST SAVINGS WILL RESULT DUE TO CHANGES IN TENDON i

SAMPLING i

i i

CHANGES IN THE GUIDES HAVE A POSITIVE IMPACT ON SAFETY EVEN THOUGH CHANGE IN RISK IS UNQUANTIFIABLE i

REDUCTION IN NUMBER OF DETENSIONED TENDONS INCREASED VISUALINSPECTIONS IMPROVED RECOMMENDATIONS FOR GREASE PURITY LEVELS 4

a' MON!TORING OF PRESTRESS LEVEL IN SECOND CONTAINMENT i.

AT A SITE 1

4 i

t

~

I i

1 i

i I

e i

b 1

4 i

i CONCLUSION (CONT.)

I i

i i

PROVIDES ADDITIONAL GUIDANCE ON DETERMINING PRESTRESSING i

FORCES i

i i

13 OF 26 PLANTS ARE ESSENTIALLY USING REV.3 OF GUIDE I

i.

10/19 AFTER 1974 3/7 BEFORE 1974(PRE 4UIDE)

{

)

i i

i l

i I

i i

i

[

l I

,e.-

l 1

l 1

SAMPLE TECHHICAL SPECIFICATION t

CONTAINMENT SYSTEMS CONTA1HPENT STRUCTURAL INTEGRITY LIMITlHG CONDIT10H FOR OPERATION 3.6.1.6 The structural integrity of the containment shall be maintained at a level consistent with the acceptance criteria in Specification 4.6.1.6.

APPLICABILITY: Modes 1, 2, 3, and 4 ACTION:

With the abnormal degradation indicated by the conditions 10 a.

Specification 4.6.1.6.la.4, restore the containment to the required level of integrity or verify that containment integrity is maintained within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and perform an engineering evaluat?;...!

the containment and provide a Special Report to the Commission withir, 15 days in accordance with Specification 6.9.2 or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

b.

With the ir?.scated abnormal degradation of the structural integrity other than ACTION a. at a level below the acceptance criteria of Specification 4.6.1.6, restore the containment vessel to the required level of integrity or verify that containment integrity is maintained within 15 days; perform an engineering evaluation ot the containment and provide a Special Report to the Commission within 30 days in I

accordance with Specification 6.9.2 or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REOUIREMENTS 4.6.1.6.

CONTAINMENT pRESTRESSlHG SYSTEM i

The structural integrity of the prestressing tendons of the containment vessel shall be demonstrated at the end of 1, 3, and 5 years following the initial containment vessel structural integrity test and at 5-year intervals thereafter.

For combined inspections of two containment vessels in a plant,-

the inspection schedule shall be as shown in Attachment 1.

4.6.1.5.1 The adecuacy of prestressing forces in tendons shall be demonstrated by:

1 Attacnment 2 to Enclosure 2 l

4 i

2 Determining that a random but representative sample of at least a.

tendons (

hoop, vertical, dare inverted i

U).each have an observed lift-off force within the predicted limits established for each tendon. For each subsequent inspection, one i

tendon from each group shall be kept unchanged to develop a history and to correlate the observed data.

1 The procedure of inspection and i

the tendon acceptance criteria shall be as follows:

1.

If the measured prestressing force of the selected tendon in a group lies.above the prescribed lower limit, the lift-off test is considered to be a positive indication of-the sample tendon's acceptability.

If the measured prestressing force of the. selected tendon in a group 2.

lies between the prescribed lower limit and 90% of the prescribed lower limit, two tendons, one on each side of this tendon shall be checked for thetr prestressing forces, it the prestressing forces of these two tendons are above 95% of the prescribed lower limits for the tendons, all three tendons shall be restored to the required level of integrity, and the tendon group shall be considered as acceptable, it the measured prestressing force of any two tendons falls below 955 of the prescribed lower limits of the tendons, additional lift-off, testing shall be done to detect the cause and extent of such occurrence. The conditions shall be considered as an indication of abnormal degradation of the containment structures.

3.

It the measured prestressing force of any tendon lies below 90% of the prescribed lower limit, the defective tendon shall be completely detensioned and additional lift-off testing shall be done so as to determine the cause and extent of such occurrence.

The condition shall be considered as an indication of abnormal degradation of the containment structure.

4 If the average of all measured prestressing forces for each group (corrected for average condition) is found to be.less than the minimum required prestress level at anchorage location for that group, the conditten shall be considered as abnormal degradation of the containment structure.

5.

If from consecutive surveillances the measured prestressing forces for the same tendon or tendons in a group indicate a trend of prestress loss larger than expected and the resulting prestressing forces will be less than the minimum recuired for the group before the next scheduled surveillance, additional-lift-off testing shall be done so as to determine the cause and extent of such occurrence.

The condition shall be considered as an indication of abnormal degradation of the containment structure, k

.m p

.c

t 3

t 6.

Unless there is abnormel degradation of the containment vessel during the first three inspections, the sample population for subsequent inspections shall include at least tendons (

hoop, vertica l,

dome, invertid'"U7.

b.

Performing tendon detensioning, inspections, and material tests on a i

previously stressed tendon from each group. A randomly selected tendon f ru each group shall be completely detensioned in order to identify broken or damaged wires and determining that over the entire length of the removed wire sample (which should include the broken wire if so identified) that:

1.

The tendon wires are. free of corrosion, cracks, and damage, and 2.

A minimum tensile strength of psi (guaranteed ultimate strength of the tendon material) exists for at least three wire samples (one from each end and one at mid-length) cut from each removed wire.

+

Failure to meet the requirements of 4.f,1.6.1.b shall be considered as an indication of abnormal degradation of the containment structure.

Performing tendon retensioning of those tendons detensioned for.

c.

inspection to at least force level recorded prior to detensioning or the predicted value, whichever is greater, w'ith the tolerance within minus zero to plus six percent (61), but not to exceed 70% of the guaranteed ultimate tensile strength of the tendons. During retensioning of these tendons, the changes in load and elongation should be measured simultaneously at a minimum of three approximately equally spaced levels l

of force between zero and the seating torce.

If the elongation corresponding to a specific load differs by more than 101 from that L_

recorded during the installation, an investigation should be made to l

ensure that the difference is not related to wire failures or slip of j

wires in anchorages.

1 This condition shall be considered as an indication of abnormal degradation of the containment structure.

Verifying tne OPERABILITY of the sheathing filler grease by assuring:

d.

1.

There are not changes in the presence or physical appearance of the sheathing filler-grease including the presence of free water.

{

2.

Amount of grease replaced exceeds 5% of the net duct volume, when injected at 10% of the specified installation pressure.

3.

Minimum grease coverage exists f or the different parts of the anchorage system, 1

4 During general visual examination of the containment exterior surf ace, the grease leakage that could affect containment integrity is not present, and 4

i i

l 4

5.

The chemical properties of the tiller material are within the tolerance limits specified as follows:

Wat r Content 0-10% (by dry wt.)

Chlorides 0 - 10 ppm Nitrates 0 - 10 ppm J

Sulfides 0 - 10 ppm Reserved Alkalinity

> 50% nf the installed value; (Base Numbers) 0 - 5 (for older grease)

Failure to meet requirement of 4.6.1.6.1.d shall be considered as an indication i

of abnormal degradation of the containment structure.

4.6.1.6.2 End Anchorages and Adjacent Concrete Surfaces As an assurance of hardware (such as bearing plates, stressing washers, wedg of all tendons selected >for inspection shall be visually exa, mined.and buttonheads).

full inspection, only visual inspection need to be performed.

During Tendon anchorages selected for inspection shall be visually examined to the extent practical without dismantling the load bearing components of the anchorages.

Bottom grease caps of all vertical tendons shall be visually inspected to detect grease leakage or grease cap deformations.

chould also be checked visually for indicattori of any abnormal condition.The s Significant grease leakage, grease cap deformation 6* iMormal concrete condition shall be considered as an indication of WNwal degradation of containment structure.

4 6.1.6.3 Containment Vessel Surfaces The exterior surface of the cbntainment shouId be visually examined to detect areas of large scaling. D-crackino in an area of 25 sq deterioration or d'isintegration, or grea. ft. or more, other surface se leakage, each of which can be considered as evidence of abnormal degradation cf structural integrity of the containment.

This inspection shall be performed prior to the Type A containment leakage rate test, i

I L

SAJJPLE SIZE CRITERIA (SEE SECTION 4.6.1.6.1) 4%

2%

0]

5 1,0 2,0 3,0 TIME AFTER INITIAL STRUCTURAL INTEGRITY TESTING OF CONTAINMENT, YEARS (Liftoff Testing Schedule, Containment F!o. 1) 2 YEARS (f1AXIMUM) l h}

I5 2,5 3,5 TIME AFTER It!ITIAL STRUCTURAL INTEGRITY TESTING OF CONTAINf1ENT, YEARS (Liftoff Testing' Schedule, Containment No. 2)

Schedule to be used provided:

The containments are identical in a?1 aspects such as size, tendon a.

system, design, materials of construction, and method t,f'

+"

construction.

E b.

Their I$1Ts are performed within two years of each other.

I There is no unique situation that may subject either con-3 c.

tainment to a different potential for structural or tendon de teriora tion.

Fig.

Schedule of Lif toff Testing for Two Containments at a Site b __ _ _____ _

~'

~ ~

'~ '

. - - - - ~ - - -

i lo, UNITED STATES

[

NUCLEAR REGULATORY COMMISSION c

n i

j wAsHiwovow, o, c.noess r.

• %...../

REGULATORY GUIDE DISTRIBUTION LIST (Division 1)

SUBJECT:

REVISION 3 0F REGULATORY GUIDE 1.35. " INSERVICE INSPECTION OF UNGROUTED TENDONS IN PRESTRESSED CONCRETE CONTAINMENTS," AND i

REGULATORY GUIDE 1.35.1, " DETERMINING PRESTRESSING FORCES FOR INSERVICE INSPECTION OF PRESTRESSED CONCRETE CONTAINtiENTS" The purpose of this letter is to call your attention to the attached copies of Revision 3 to Regulatory Guide 1.35 and Regulatory Guide 1.35.1 which will become effective 1989. These guides have been revised to reflect, (1) public comments, (2) insights gained by the survey and study of actual i

inservice inspections (NUREG/CR-2719), and (3) consideration of the IWL on

" Inservice Inspection of Concrete Pressure Components" of the ASME B&FV Code,Section XI published January 1989.

The staff will use the provisions of the guides as indicated in Section D of the guides when the relevant technical specifications are proposed or amended.

Licensees are advised to adopt the revised guider, in their entirety in l

formulating their inservice inspection programs. If the technical specifications i

remain unchanged,.then the staff will continue use of existing guidelines.

The-staff-wev4d-eneevrage-6he-44sensees-of-opera 44mg p4 ants-to-pev4ew-the4r en4st4mg-tenden-4mserv4se-4mspest4en-prespass-and-ovaivate-them-from-%he e46mdpe4mt-of-opera 44ng-senveniense,-safety-4mprovements,-and-sest-pedus44en poten64a4.

Thestaffrecogni:esthatinsomeolderplants.(plantsoperatingbefore initiai issuence of Regulatory Guide 1.35-1974) complete adoption of all provisions of the revised guides may not be feasible without extensive retrofitting, in-these-eases,-the-44sensees-spe-advised-to-present-their rev4 sed-4nserv4se-4nspes%4en-pregpass-for-staff-pev4ew, In those cases, if the licensees should decide to improve their inservice inspection plan by adopting Reg Guide 1.35, they are advised to present their revised technical J

specification to the s3ff for review, noting those provisions they are unable i

to implement.

l A copy of sample technical specifications incorporating the provisions of the guides is available upon request.

Eric S. Beckjord Director Office of Nuclear Regulatory Research l' to Enclosure 2 i

_ to the Minutes of CRGR Meetina No. 1672 NUBARG Appeal to EDO on use of Nuclear Heat for Hyarotestina BWRs October-25, 1989 TOPIC.

The Committee discussed with the staff the generic backfit appeal by the Nuclear Backfitting and Reform Group (iFJBARG) of the NRC staff position prohibiting the use of nuclear heat and specifying solid water for pressuri-zation in the hydrotesting of BWRs.

L. B. Marsh (NRR), P. N. Randall (RES),

and R. Herwann (NRR) sponsored the disputed staff position for purposes of 4

discussion.ith CRGR at this meeting.

Briefing slides used by the staff to guide their presentations and discussion with the Committee at this meeting are enclosed (Attachment 1).

BACKGROUND 1.

The generic backfit repeal item sent by the EDO to CRGR for. review in this matter was trant,3itted by memorandum dated September 18 1989, i

J. M. Taylor to E. L. Jordan; that document is identified as follows:

a.

Letter dated March 16, 1989, from N. S. Reynolds and D. F. Stenger (Counsel to NUBARG) to V. Stello, Jr., setting forth the subject i

generic backfit appeal.

2.

The following additional documents (referenced in item 1.a above, or otherwise relevant to CRGR consideration of this matter) were provided for the information of CRGR members by CRGR staff (Attachment 2):

a.

Memorandum dated April 29, 1986, R. M. Bernero (NRR) to A. F. Gibson l

(Rll), providing staff position regarding the performance of BWR system pressure tests (Leakage and Hydrostatic) using nuclear heating and a definition of plant start-up with regard to pressure testing.

t b.

Memorandum dated May 2, 1986, G. C. Lainas (NRR) tu D. R. Muller (NRR), reiterating staff-position in 1.a. but also stating bases for granting GPC one-time exception to staff position.

l c.

Letter dated May 5,1986, D. R. Muller (NRC) to J. T. Beckham, Jr.

(GPC), granting one-time-only relief from staff position in 1.a for Hatch 1 startup.

d.

Letter dated April 10, 1987, J. H..Sniezek (NRC) to J. P. 08Reilly (GPC), denying GPC's plant-specific appeal of staff position in 1.a.

e.

Letter dated April 25, 1988,;N. S. Reynolds (NUBARG) to T. E. Murley (NRC), transmitting NUBARG generic backfit claim (arising from i

denial of GPC appeal in 1.d).

e v

g.

f.

Letter dated August 17, 1988, L. C. Shao (NRC) to N. S. Reynolds (NUBARG), denying NUBARG generic backfit claim.

g.

10 CFR Part 50, Appendix G. Fracture Toughness Requirements (annotated).

h.

Memorandum dated November 13, 1978, O. Parr (NRC) to G. Sherwood 1

(GE), transmitting NRC staff Topical Report Evaluation on HEDO-21778, " Transient Pressure Rises Affecting Fracture Toughness Requirements for BWRs."

1.

Excerpts from Topical Report NEDC-31140, dated January 1986, "BWROG l

Evaluation of Reg. Guide 1.99, Rev. 2 Impact on BWRs."

j j.

Excerpts from ASME Code (version dated July 1 1989);Section III, NCA-9000 - Glossary (Definition of " Hydrostatic Test"), and Section XI. Article IWB-5000 (System Pressure Tests).

i i

l CONCLUSIONS / RECOMMENDATIONS i

The Committee's discussions with the staff focused in the three general areas identified for CRGR consideration in the ED0's request to CRGR for review of this matter, i.e., (1) the technical correctness of the staff position on the nuclear heat issue, and the specific safety / risk considerations involved, (2) the procedural correctness of the staff's backfit denial process, and the specific backfit issues involved, and (c) the correctness of staff inter tions and positions being taken pursuant to applicable regulations (e.g.preta-

, 10 CFR part 50, Appendix G)', and possible ambiguities in those regulations that need correcting. Specific topics covered within these three. general areas are indicated in Attachment 3 to this Enclosure.

As a result of the discussions with the staff at'this meeting, the Committee I

requested additional information from the staff in several of the specific areas covered. The Committee will consider a draft response to the NUBARG dppeal in accordance with the E00 request at the next meeting in order to s

complete this review.

I e

?

l

}

k

---,e e

a m -.

- ~ - - -

v

t'; I, i

CRGR BRIEFING PACKAGE USE OF NUCLEAR HEAT FOR BWR PRESSURE TESTS --

RESPONSE TO NUBARG APPEAL l

l l

PRYOR N.-RANDALL

)

OCTOBER 25, 1989 l

Attachment I to Enclosure 3

1 t

i l

i I

l

[

ss-l l

r e--

N f/*Merl W p f r*

' "T R R ~r- =

, Y eni

/

5 w---

f h 4'

\\/

CORE O

D

(~~~~b i

~

===ee. MOTOR RECIRCULATION PUMP l

FEED o CONTROL HEATERS ROD DRIVES i

l l

4

//M

/ne JMMAG6/V CUkVf

%Ab 1.

Sf'd

,W eiz ses A

Tep,pera kWQ Y o

l

~..

~-

I i

4 ferehen f NCA 9000 - G1.055 ARY j

requirements. This term, when used in conjuction with by this Section. For Division 2, refers to metallic ma.

quah6 cation of personnel to perform quality related tenals, as well as to nonmetallic matenals, conforming activities, shall mean a wntten examination.

to the specifications permitted in this Section.

Fabricatoon. Those actions required to manufacture Materal Manufacturer (Metallic). An organtistson components, pans, and mpunenances. Dese actions which certines that the materialis in compliance with may include forming, machining, assembling, welding, the requirements of the basic matenal spec 6 cation in brating, heat treating, examination, testing, inspection, addition, the Matenal Manufacturer (Metallic) per.

and ceniAcation. Fabricanon does not include design, forms or supervaes and directly controls one or more Nold Point. ' A designated stopping plane during or of the operations which afect the material propenies followmg a specine activity at which inspection or es.

required by the matenal speci6 cation and vennes the amination is required before further work can be per.

satisfactory completion of all the requirements of the formed.

material specincation performed prior to that certin.

Hydrosterac Test. De pressuriaation of an item to a test pressure using water or other liquid as a testing M88"i81Specpcetion. A document which establishes medium with the required examinadon preactibed by the requirements for a material.

the Code.

Maseria/ Supplier. An organization which supplies ma.

Identvication and Verpcetion Propen The Identifi.

terial produced and ceni6ed by Material Manufactur.

canon and Verincation Program is a system for the ers, but does not perform any operadons which afect positive identiAcation of material during storage and the material except when agreed upon by the Certificate handling and veriAcation of the identity of material on Holder who uses the material in Code construction or the accompanying Certi6ed Material Test Report or when so authorised in a Quality System certi6cate Certi6cate of Compliance at the time of shipment both (Materials). He Material Supplier may perform,and i

into and out of the Material Supplier's facility, cenify the results of tests, eaminations, repairs, or Inspector, ne Authorized Nuclear Inspector as de.

mtments r@ed h ee material s#6ca6m w by Aned in NCA 5122. De term laspector, as used in this Section which were'not performed by the Material

"""I"*""'

(

this Subsection, is the same as the term Authorized Inspector as used in Division 2.

Medpcotion. A change to an item made necessa.y by, Instructions Detailed written directions provided to or resulting in, a change in design requirements.

persons or organizations to ensure proper completion Monitor, To watch, observe, or check to ensure com.

of a task.

pliance with this Section and the requirements of the Joiniat. De act of connecting two or more items to Owner's w Cenincate Holder's Quality Assurance one another, by welding, brazing, bolting, or other Program. h,s act:Wy is ngt neessarHy dmumented i

mechanical means.

or required to be on a continuous basis.

Jurisdictional Roundaries ne physical limits of a Nonco@rmance. A deAciency in a characteristic, doc.

l Code item which are identi6ed to determine the ap.

umentation, or procedure that renders an item or ac.

plicability of Code rules for that item.

tivity unacceptable or indeterminate.

Linear Support (Lineer Type Support). A structurag Nonmetallic Material Cmutituent Supplier. An orge.

(

element aG ag under essentially a single component of nization which manufactures, produces, and supplies direct stress. Such elements may also be subjected to de concrete constituents for plastic concrete or grout shear stresses. Examples of such structural elements in accordance with the Construction Speci6 cation.

are tension and compression struts, beams and columns Jonmetallie Material Manufacturer. An organimation subjected to bending, trusses, frames, rings, arches, and which receives, stores, conveys, and combines the con.

cables.

crete constituents to produce pir.stic concrete or grout load Capacity Dora Sheet. De design document used in accordance with the Construction Speci6 cation.

in lieu of a Design Report when a component repon Owner, ne organization which obtains a construction is designed by lead Rating to verify tha' the require.

permit from the Regulatory Authority for the con.

ments of NF.3000 have been met.

struction of a nuclear power pl.nt.

Material. For Division I, refers to metallic materials Port. An item which is attached to or becomes a portion l

which are manufactured to an SA SB, or SFA spec.

of a component or component support before comple.

incation or any other material specincation permitted tion and stamping of the component or component 11 2,

T i

l f

l ARTICLE NB-6000 TESTING L

NB 6100 GENERAL REQUIREMENTS ommente.d that special precautions for prote: tion of f,, [

g,,,

NB4110 PRESSVAE TESTaNG OF COMPONENTS, APPURITNANCEF, AND SYSTEMS NB4111 Seepe of Pressere Testing NB4113 Witness 6ng of Pressera Tests All pressure retaining components, appunenances, Pressure toting required by this Article shall be and completed systems shall be pressure tested. The pedormed in the presence of the inspector, except that

% preferred method shall be a hydrostatic test using water testing of each line valve and each pump having piping as the test medium. Bolts, studs, auta. washers, and connections of 4 in, nomin.al pipe size and less need l

gaskets are esempted from the pressure test.

not be witnessed by the laspector. For line valves and l

pumps 4 in. nominal pipe size and less, the inspector's NB4112 Pneumatie Testing reviaw and acceptance of the Certdcate Helder's test M will be authonaation to sign the Data Repon A pneumatic test in sagt m with NB4300 may Form and take precedence over NCA.$230.

be substituted for the hydrostatic test when permitted by NB4112.l(a).

NB4112.1 Poemmette Test IJaltations NB4114 Time of Pressere Testing (a) A pneumatic test may be used in lieu of a by' Adrostatic test anig when any of the following conditions NB4114.1 System Pressere Test. The installed sys-exists:

tem shall be pressure tested prior to initial operation.

(/) when componenta, appurs=== or systems (a) The pressure test may be performed progres-are so designed or supponed that they cannot safely a vely on erected portions of the system.

be filled tith liquia;'

(b) Systems which are open ended, such as spray (1) when components, appure===== or systems systems, may be pressure tested with the noaale et.

---== which are not readily dried are to be used in services tachment opemas plugged. The spray nozzles and their where traces of the testing medium cannot be tolerated, attachment weld joints or mechanical joints need not be pressure tested.

(b) A pneumatic test at a pressure not to exceed 25% uf the Design Pressure may be applied, prior to NB4114.2 r' n; mt and A;f r------- Pressere either a hydrostatic or a pneumatic test, as a means Test of locanng ~ leaks.

(a) Components and appunenances shall be pres-NB4112.2 Precautions to Be Employed im Pnee-sure tested prior to installation in a system, except as permitted in (b) below, matie Testing. Compressed gaseous Guid is hazardous (b) The system pressure test may be substituted for when used as a testing medium. Therefore, it is rec

• a component or appurtenance pressure test, provided:

(1) the component can be repaired by welding in i

accordance wi:h the rules of NB-4130 and NB4450, h teu any te sede mth tlw item W.3 tested pamaHy sDed if required, as a ruult of the system pressure test; muh hqwd. d desired (2) the component repair weld can be postweld 233 3

a 1.

4 9

p jr 4

TABLE IWB-2500-1 (CONTT l

EXAMMATION CATEGORIEE EXAMf00AT10N CATEGORY 8-P, ALL PRESStfRE RETAINfMG COMPONENTS (C00tT*D)

Estent and I'veguency of Enamination

- 1st

, Socceflive Deferral of item Parts Test Esandantion Acceptaece 8aspe tion Inspection laterwafs.

Inspection to No. -

. Examened

%....., :s' Method

• Standard fervat 2nd, 3ro. *th End of laterval Piping 915 50 Pressure Retairdng Soundary System ledage Visual. VT-2 IW8-3522 Each refb eens autage' Each refuesng outage'. Not pervWssible

. test ' (fw8-8 5221) 815.51 Pressurs Retaining Sc.nulary System hydro-Visual. VT-2 IW88-3522 One trat' One test per intervat*

Perwssitde statk test' tlW S-5222) i M

B C

PRIfnps h

815 60 Presu? Retain?ng Sountry Systern lydage

- Visual. VT-2 IW8-3522 Each refueang outage

• Each refuenne outage' Not permissilde h.

testt' (fWS-m 5221)

^

815 61

- Pressure Reta'ning Soundary System hydro-Vtyeaf. VT-2 fW 8-3522 One test

• One test per intervat*

Pernussstde k

JLie test 8 y

tlW8-5222)

M g o

s hI va.ve.

=,

815 70 P' essure Retaining Bownfary System ledage

. Visual. VT-2 IW8-3522 Each refuesing outage' Each refuesne outage' Not perrnissdrie test 8 ' (fW8-5221) g 815 71 Pressure Retaining Soundary System hydro-Visual, VT-2 ~

IW8-3522 One test

• One test per intervat*

PermisAple g

statk tests

' (fW8-5222) m

.'i f~

NOTES:

(1) The.vessure setaining boundary during the system leakage test shall currespond to the reactor coolant system boundary, with all valves in "

- the normal position, whkh is required for eormal reactor operation startup. The VT-2 esamination she. % wever, estend to and include the ss*ouf closed wahre at the boundary entremity.

M (2) The pressore retaining boundary during the system hydrostatte ecs' sf!381 include all Class 3 cosuponents wit 9dn the system boundary

[~

t3) System p.vssure tests of the reactor coolant system shall be conducted in accordance udth IWA-Seco. System pressure tests for repaired.

repliced. or altered components shan be governed by IWA-5214tc).

g Asu=8 enamination of IWA-5240. -

gj NAM S he system tedaae test ifW8-5221) st all be conducted prior to plant startup fossen.q each reactor refueNag outage.-

te.) The system hydrostatic test tIW8-5222) shau be ronducted at or near the end o8 each Inspection interwaf.

.f.

7.t?) A system hydrostatic test flW8-5222) and the accompanying VT-2 examination are acceptalde in lieu of the system ledage test (IW8-

$2211 and VT-2 examination.

E g

//CA(16 A,pfusis 6, fa,.ppropf 3.J. 3r 1

. 3. Wiien the core is critical (other than for.

the purpose of low-level physics-tests), the j

temperature of the reactor vessel must not be lower than 40*F (22*C)' aboveL the mini-

. 1 i

mum permissible temperature of paragraph 2; of this sectioninor lower than the mini-mum" permissible < temperature for"the in-service-systern hydrostatic premire test. An exception may be made for boiling water re-actor 'v'ssels:when water level is within:the t

normal = range for power: operation and the; i l

. pressure s ess than 20 percent of the pre '

- service system hydrostatic t'st pressure, In-e this case the minimum permissible tempera-ture is 60*F (33*C) above the reference tem-perature of the closure flange regions that s re highly stressed by the bolt p. reload.

S J

1 4

[-

---#p,-

mm y-v,..

y,

.-m.,

e 5~- a. - *U..

+. - + -. - + - - -.. * - - -, - - - +. - = -. - - ~

a

i

/0 - CF/C 10 Appeka's 6, fa,ap,a f y j. 3

5. If there is fuel in the reactor during i

system hydrostatic. pressure 1 tests or leak tests, the requirements of paragraphs 2 or 3 -

i of this section apply. depending on whether

'4 the core is critical during' the test.

m 6

5 %

I e

1 1

i i

1 ij'

):

'.E

~

~t'

' i -

p' ??, ', *

'.-}-

. ' ~ ~ ~.

'~5 '

~. -.,.. _

l NEDC-31840 l

Table 6-4 i

NUCl, EAR HEATING OFTION INFORMATION l

Not Yes No # Sure:

~

L Isnuc1Aarheatingpossible?

3.

'10 2-Time to reach 200*F (hours)?

4-16 Time to reach 220*? (haur::)?

4-24

)

l s

Operational. Problems 1

1. Use of nuclear heating would require that the hydrotest be--

scheduled late in the outage, on critical-path.

2. Contradicts some plants' interpretations of ASME Code Section II.

Table IWB-2500 requirement that' hydrotest be ~ performed before'

- 1 resuming operation.

3. Nuclear operation must be' done with normal' water, level; cannot-be done with solid system needed for.hydrotest at temperatures nsar 200*F.

This is not a probles if: the hydrotest is-at. operating temperature.

4. CRD cperability must be demonstrated before use; a hydrotest ' is I

requ' ired to demonstrate'CRD operability.

4 i.

t.

5. Hydrotest would be -in startup sequence, so -any problems l

identified during hydratest would' impact.startup schedule.

L

• Possibly_due to Technical Specification requirements.

i 6-9/6-10

[

+

.t p

- d 2

gwe e. -e,, n-tw,

-,, e w e yr.--~

+<.-.m-.-

..+._,m

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

.4 4

me e,.,

w.<

.)

i l.. '.

,c.?

)

SYNOPSIS OF ' NUCLEAR HYDROTEST ISSUE' April 17,1986 - Region !! Clarification of NATCH Position - Region 11 request for staff position J

' April 30,1986 - Georgia Power - NARCH - Request for one time relief -

obtain cose inquiry l.

May 2,1986 - Lanias to Mueller - One time exception - No procedures -

Hardship - Safety Significance (one time for 4 to 7 day start-up delay)

May 5,1986 - Letter to-May 1,1986 phone call pemit one time excepti

- enforce discretion per (Further letter writt

'ter discussion of Div, DirectorwithUP-GeorgiaPower.)'

July Pd,1986 - Request to Anderson - expedite @ position-on Georgia Power 2

September 30,-1986 - Letter from~stcff granting one. time relief for HATCH-t 2 - inquiry not limited - Regulatory Guide 1.99_ working.

February 10, 1987 - Staff letter reaffirmed staff position.

i

- February 24, 1987 - Georgia Power intention to appeal March 13,1987 - Georgia Faer Appeal March 24,1987 - Meeting Announcement April 6, 1987 - Appeal Meeting Held April 6, 1987 - Supplemented Information Georgia Power April 10,1987 - Sniezek to O'Riley - Generic Letter April 25,.1988 - Letter - Nicholas Reynolds to Tom Murley - NUBARG Appeal

- August 17, 1,988 - Shao to Reynolds -

March 16, fff f M V F M G 4 e A A l ' ? ff' W b A N I

I 1

l i

l

- i t

),

~..

s

APR 17 tgg; Robert M. Bernero 2

Region 11 requests NRR to provide an interpretation of. ASME Section XI relative to ~

't the definition cf." plant startup" and whether use of nuclear heat for system leakage and hydrostatic tests mets the requirements / intention of Section XI.

-:p aedition, regardless of Section,XI requirements, does NRR consider the practice In I acceptable.of using nuclear heat for obtaining system leakage and hydrostatic conditions

/

. Albert F.~ Gibson O

9 Y

I to 1

L a

=

w-m m

m[..

mm m

m

- m.

. ~, _.. _.

.~

i

+

. ig J

4

'c, UNITE D si ATEs NUCLE AR REGUL ATORY COMMISSION

. ' W,!

g n A5 mmc 10N D C. 20S56 l

1

\\

April 29,1986 MEM3RANDUM FOR: Albert F. Gibson, Director Division cf Reactor. Safety Region 11 L

FROM:

Robert M. Bernero, Director Division of.BWR Licensing.

SUBJECT:

INTERPRETATION OF SYSTEM PRES $URE TEST REQUIREMENTS This responds to your memorandum-dated April 17, 1986,- requesting NRR r provide a position regarding the performance of system pressure tests (Leakage and, Hydrostatic) using nuclear heating and a definition of-3 plant' start-up with regard to pressure testing.

The ASME Code,Section XI requires that the System Leakage Test be performed prior to startup following each refueling outage (Table

.I IWB-2500-1, Note 5). The examination method required is a. visual, Type VT-2.

Later additions of Section_ XI pernit system hydrostatic tests to

)

be performed in lieu of the system leakage test.

Further, the Code states that reactor e.oolant shall be used ss the pressurizing medium.

Also, the Code permits visual examination to be performed at 200'F with appropriate pressure corresponding to a pressure. consistent with fracture prevention for system compor,ents that require a' test temperature

-)

above 200'F.

As you are aware ASME Section~ XI. is incorporated by reference as-part of HRC regulations. Therefore, this is an. issue whether regulations are.

i being met,'not just a simple code compliance question.

.-y. The position of the staff is that System: Pressure Tests (Leakage and Hydrostatic) are to be performed before the reactor goes critical:

from a refueling outage. The Systes Leakage Test-is a test to determine ~

if any abnormal leakage is occurring in the reactor coolant pressure J

boundary af ter its $pening and closing. The Hydrostatic Test'is-a proof test of repairs on the reactor coolant pressure boundary or.other.-

Prudence dictates that both of these tests be'perfomed at component.

the lowest temperatures that are consi:, tent with the fracture prevention criteria for the reactor vessel or other component'so that stored energy I

L can be minimized during testing, conditions by-having the system water solid. The temperature correu1rn: terms are provided to account for changes in material properites when the vessel must be heated for fracture provention. -We ( tt ; believe the temperature corrections are L

an invitation to perfon e tetting at higher temperature to. minimize the test pressures. The pressc.re!ng medium'is to be reactor coolant l

rather than stetm. He neocrue that some flashing to' steam'of any-L potential leakage could own when temperatures in excess of boiling are necessary for the test.

u

Contact:

R. Hermann a

X-27798

/p fE

~

J w

r.-

t

-mt-a-.

4 m-m--

m--.

-I------e=*

MO, twn a _ C L i

t

l J

i l

\\

(

L:

ARTICLE IWB-5000 1

l SYSTEM PRESSURE TESTS q

IWB 5200 SYSTEM TEST REQUIREMENTS TABLE IWB 52221

'IWB 5210 TEST TEST PRESSURE

'(a) The pressvu retaining components-shall be Test Temswatere, *1' -

Test Presswet2 c

tested at the frequency stated and visually esamined N

• 8

by the method specised in Table IWB 25001 Ex-amination Category B P:

33/,

(/) system leakage test IWA 5211(a);

een 3.o4 pg soo w yester 1.02P.

(1) system hydrostatic test, IWA 5211(d),

3 nus,

~'~

(6) The system pressure tests and visual examina-tions shall be conducted in accordance with IWA 5000 m e. is uw nommat eowetane oreiswe ca..

-.4 *ta loos and this Article. Reactor coolant shall be used as the 4888"Q*"y,,

pressurizing medium.

. pm, gg 1

i j

IWB-5220 PRESSURE l

IWB 5221 System f amirage Test (c) The pressure measuring instrument used ' for measuring system hydrostatic or pneumatic test pres-(a) The system leakage test shall be conducted at a.

sure s meet ee @ rents ofIWA N test pressure not less than the nommal operating pres-sure assed.ths with 100% rated reactor power.

(of The system test pressure and temperature shall IWB 523C TEMPERATURE be attained at a rate in accordance with the heat up (a) The minimum test temperature for either the limitations specined for the system.

. system leakage or system hydrostatic test shall not be l

. lower than the minimum tarnperature for the associ.

IWB 5222 System Hydrostatic Test sted pressure speciaed in the plant Technical Speci6 cstmas; (a) The system hydrostatic test may be conducted -

-(

st any test pressure speci8ed in Table IWB-52221 (6) The system test temperature shall be modined

-l corresponding to the selected test tempe.ature, pro-as required by the results obtained from each set of vided the requirements of IWB 5230 are met for all.

material surveillance specunens withdrawn from the ferritic steel comynents within the boundary of the.

reactor vessel during the service lifetime.'

(c) For tests of systems or portions af systems mn-system (or portion of system) subject to the test pres-structed entirely of austenitic steel, tes: tempe sture sure (see IWA 5245).

(b) Whenever a system hydrostatic test is conducted

-limitations are not required to meet fracture p. *ention in which the reactor vessel contains nuclear fuel andcriteria. In cases where the components of the. system the vessel is within the system test bounds y, the test are constructed of ferritic and austenitic stee'.s that are pressure shall not exceed the limiting conditions spec-

'nonisclable from each other during a'syrem leakage -

ined in the plant Technical Spe:i6 cations.

or system hydrostatic test, the test temprature shalla

. be in accordance with (a) above.

137 L

a m ;,.

a

~ - -.. - - - - -- - - - - ~

^^ ~ ~ " ~ ~ ~ ~ ~ ~ ~

~

~

L I

2.

Mr. J. T. Beckham, Jr.

-y for Hetcr. Unit 1 on a one tin.e basis to rerfor.. '.... -*'" tired pressure tes t-

[

using nuclear hett es it has been doing for a :vmter of years. This was in;1emented by Region 11's exercising enforcer +nt discretion es dis:Usat 4 e cenference cell with Region 11._'NRP, and GEC personnel on May 1,19EC.

Sincerely, t'.

.M Dat.iel R. Muller, Director BWR Project Directorate #2 Divisor of BWR Licensing Enclesure:

As Stated j

cc: See next page e

'I l

1 l

i 1:

1 lr.

l '.

(.:

i 4

e l

z

v...

-2 SFP 3 01996 iW

'o 1

account' for shift charges from some residual elements' that differ from those censidered significant in Revisior.1 of.the Guide. Corsnents on Draf t Revision-

? to this Guide are currently being considered and are expected.to be resolved within the next four to six months.

(

a It is possible that the code interpretation, the resolution'of. Regulatory Guide i

1.99 comments, and the adjunctive issues of radiation dose to plant personnel-and reactor vessel beltline inspection may impact our position. In view of our l

desire to consider any new information that may have a bearing on this' issue, we.

t agree that if GPC chooses not to fulfill its commitment to use pump heat for pressure testing during the current outage, it will be accentabir to perfom the f

required pressure tests on U_011.2,= or a one-time basis only, using nuclear heat i

as has been the GPC practice for a number of years.

~

. Since rely..

wl y

w7.L bert M. Bernero, Director Division of:BWR Licensing Office of Nuclear. Reactor Regulation 1

cc: See next page l

e k

4

ll 1

' XI 1 a6 34, XI.1-86 35 Secuon XI - interpreic.ons No. 20

-+-

Interpretation: XI 186 34

.i

(

Subject:

Section XI, Division 1, IWA 5211(a), Table IWB 2500-1 Examination Category

~

B.P, and IWB 5200, Hydrostatic Test Temperatures - Fracture Prevention.

Criteria, Pressurizing Medium (1980 Edition With Addenda Through Winter

> 1961)

Date issued:

September 18,1986

)

File:

IN86 0178 Question (1): IW8 5222 and IWB-5230 require the modification of hydrostatic test tempestures 1

and pressures to meet fracture prevention criteria. DoesSection XI address the techneques by which the elevated temperatures are produced during the conduct of the hydrostatic test!

(

i Reply (1): No.

.)

Question (2): IWB 5210(b) states ? Reactor coolant shall be used as the pressurizing medium,"-

when performing system pressure tests. Can a mixture of steam, water, and noncondensable gases, in

[ a proportbn no greater inan that present dunne normal startup, be used as th 1

1 s

meet the requirements of IW8-5210(b) when performing system pressure tests of the primary reactor.

coolant systems for boiler water reactorst Reply (2): Yes.

Interpretation: XI 146 35

(

Subject:

Section XI, Division 1, IWA-4400(b)($) and IWA 5000, Component Connection p

NPS 1 and Smaller (1977 Edition Wah Addenda'Throqilh Summer 1978 and Later Editions and Addenda Through the 1986 Edition)-

y Date issued:

September 18, 1986 File:

IN86-019 Question: May a cover welded over a 1 in. diameter or smaller opening, in a valve latiger than NPS 1, be consedered a component connection NPS 1 and wnaller under IWA 4400(b)(5), and the weld and cover be exempted from the hydrostatic test of IWA 5214i

(

Seply: Yes.

1 o

e Nr 6

t20'

]f,

- - -. -.. ~... -

XI 186 33,11186-54 Secten XI. Interpretatkms Nol 21

% interpretation: XI 186-53

'l i

Subject:

Section XI, Division 1, IWA 5211, Noncritical. State - Pr*ssure Tests (1980 Edition With Addenda Through Winter 1981) i Date issued:. February 11, 1987 File:

IN86-017A I'

Question: DoesSection XI, Division 1, IWA-5211 require that a reactor be in a noncritical state

)

when pressure tests (hydrostatic and leak tests) are performedt -

r l

l Reply: Core criticality dug pressure testing is not addressed by Section XI, Division 1,

.i 1

~

87^ m'ation: XI 186-54 l-Subsect-Secten XI, Division 1, IWA-2400, Extensen of Inspection intervaF, (19741 Edition Wah Addenda Through Summer 1975)

Date issued:

February 11, 1987 File:

IN87 001 Question (1): Is it a requirement of Section XI, IWA*2400, IWS 2400, IW62400, or IWD-2400 ' '

that the extension of inspection intervals by as anuch as one year to be concurrent with plata outages (as allowed by IWA 2400) be apphed only dunns the last one-third of the intervait -

(

l Reply (1): No.

t I

l Questen (2): May the extensions in inspecten intervais in (WA 2400 for both out-of-service and plant outage conditions be applied serially?

Reply (2): Yes.

1 1

'l e

g 136 4

.s

.t

Mr. James P. O'Reilly 2

- ~.

__ ____ g g

-gp i

tests with the reactor critical at about 5 percent rower, at temperatures

~!

mil above the air.imum required system temperatures.and with the drywell i

uttoned up (i.e., as GPC described its' test condittors), such drywell 1

'envirorsnental conditions are not conducive to a' thoroegh and deliberate i

j visual inspection. We believe that a more taliberate visual inspection and one posing less potential danger to the. inspectors can better.be perforsed in the cooler and less hazardous environment associated with reactor coolant system temperatures near the minimum regt.4 red test temperature. ;This is,' in our view, more consistent wit' '.no intent of completing a satisfactory boundary pressure test before resuming operations.

4 We have therefore concluded that the Section.XI system hydrostatic and -

system leakage pressure tests for Hatch Units 1 and 2 are to be performed-with the reactor non-critical as stated in our May 5,1986 position.

We appreciate your effort in providing your views and in quickly responding.

to our request for information so that we coula resolve this issue.

Sincerely, C.iginal signed by James H. Snierek dames H. Sniezek, Deputy Director I

0.ffice of Nuclear Reactor Regulat,1on ec:.See next page

\\

e DISTRIBUTION

. Docket File

.JPartlow i

,NRC PDR.

NThompson-

.l.ocal POR Glainas.

PDf2 Plant File....-

RHennann '

$'GRivenbark 5Norris.

0GC-Bethesda.

HDenton

~

.JSniezek

e tt r.nACRS (10)

.i RBernero.

s EJordan

(,

BGrimes 4

3 w

I!

N n

\\

..0. BED!.2.....p.sk/wL.......9hl/.ADh

==>

......DB(,,

,.. p.B.t/)A)0f2....

,,,,,,....m._

c.R.1.v.8n.s.a r.k.u

!.. Bw.K.... n.....

..w

..88rnsc.c.....

h nck --

...W1 8

->.9.4.7!81.......?!.?!.s.?.......o..ue?...... ;.9.M.I.e.L....... 9.Rit.sL....oAdD.R.. -

wee romw sie no uoiwacu on.o OFFICIAL RECORD COPY-

.* u 5 *

• 8" *~'**

ypr radh/ Clah, civen the fact that u+

testing for a number c:

aff has permitted nuclear pressure testing with the core erat cal, and that the ASKE has im s, that the regula

- -?

the code not to preclude such testing, it is clear thatnterp_reted position constitutes a backfit within the meaning the Statf 50 109..

i! a backfit and should not be imposed until a b has been performed in accordance with Section 50 109ackfitting analysis as NUBARG appreciates the Statf's'sonsideration of this This issue is a significant one for lisensees because th claim.

pump heat to-conduct pressure tests may-add up to $3g89 e use of th,e,L,duratiers of refueling outages and any not be a viabl.4 ys,to for all plants.

4 s

significent plant modifications and would result-in sub eo s

-replacement power costs.

that *there is a minimal difference in the safety affOn the other ha between the past practies and the Staff's? mew position orded* -

4, annr.a.-

See note-

.An additional-problem faced by til reactor types is the inability to control coolant temperature during testin heat.

Pressure testing during a normal startup does not pr g with pump the same probles.

as those described in Generic Letter 84-11 (which cont esent i

inspections for intergranular stress sorrosion oracki stainless stoel piping in boiling water reactors), are performed ng in during reastor critical conditions.

Should you have-any questions be happy to discuss them with you.regarding this claim, we would Sinoare y, t

6-l

.stosah8.iernoids b'f:.

l-

,t

,n-CounseZW to

$ lea...

. 7 *, -

i ::.t..: and. Refeas aroup%Ang((.

r

'.u.,

Dh111%7 I lit

.t, :.

t -

,' 3 y~.

,....v.

7.o.,,

..r 3

n 7 y,.

.g.g,. :,,

..y.- -

' Y. '. t '. W G. tt.. '. '.

( u gj, ; y, J,Ao*

museative careeter for" operations, l:-'.$,j,Q

,y

w...

,c....

pg. :;. ; 9-p

• -O Q

/~7 l-s

. -. - - -. ~. -

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

t N

NUCLCAR UTILITY sACKFITTING'

]

w- /@

A;

.~o - - -

g o.....

.smes,on. e. c. aooos a so, March 16, 1989 t 68*** uo8' an s,co

_ Q ;& f U.S. Nuclear Regulatory commission Document Control' Desk 3

Washington, D.C.

20555 y.

Attn: ' Victor Stallo, Jr.

. Q J

\\

~

.Re:

Backfitiding Appsel Regarding System.

i MML and L*4kana Testine

Dear Mr. Stallo:

i Pursuant to Section 044'of Staff-Manual Chapter'0514, the Nuclear Utility Backfitting and Reform Group (NUBARG) appeals a Staff denial'of a claim of.backfit under 10 C.F.R.-5 50.109..The claim concerns a Staff. interpretation of system hydrostatic and leakage testing requirements-under ASME code Section XI.-

NURARG presented. its claim in a letter to the Director of Nuclear l

Reactor Regulation on April 25,.1988'.

The Staff denied the c1&in by letter dated August:17, 1988.from the. Director, Division of t

(

l Engineering and Systems Technology.

BACKGROUND A.

Factual Backgrount' This appeal.is concerned with a:new Staff position on the i > acceptability of.* nuclear" hydrostatic and leakage testing by BWRs,1J,,l the use.cf euclear power during norsp1'startup, as..

pressurize. heat. generated by, recirculation pumps, to heat.up and opposed to the reactor coolant system for, performance of.tre3.

tests.

As discussed,below;sthis testing.nethod is clearly s

permitted by Section'ZV.A.5.of-.20 C.F.R. Part 50, Appendix G, and the Staff has recognized ^thht there is minimal difference.in-safety between testing.with nuclear heat rather than pump heat.

In accorriance with Section:IV.A~.3.of Appendix G,-testing.with.

nuclear heat is conducted at low power and with the vessel water j

j level within the-normal rance for never'cssration.-

.-o i

j The f-lexibility to use this method is important 'because -

control of reactor coolant temperature is more difficult with the use of pump heat and because testing with'the reactor at low l

l l2 l

\\

j

-.~...

1:

.l l

1 i

OBJECTIVES OF PRESSURE-TESTING-i 1.

LEAK DETECTION o

LEAKS MAY BE PRECURSOPS OF A PIPE BREAK OR OTHER RUPTURE-0F THE REACTOR COOLANT PRESSURE BOUNDARY.

l o

LEAKING FLUID MAY DAMAGE ADJACENT CABLES, INSULATION, ETC.:

o

- LOSS OF REACTOR COOLANT.

2.

UNCOVER GROSS NEGLIGE.1CE IN MAINTENANCE o

. IMPROPER REPAIR WELDING.

l

(

o IMPROPER REASSEMBLY OF MECHANICAL JOINTS.

l i

s 4

r g

i I

USE OF NUCLEAR HEAT FOR PRESSURE TESTING 15 A SAFETY ISSUE i

l 1.

BEING CRITICAL WITH THE REACTOR AT 550*F HINDEPS FINDING LEAKS-BY VISUAL INSPECTION,'ESPECIALLY INSIDE THE DRY WELL.

'2.

BEING ESSENTIALLY A PNEUMATIC TEST WITH STEAM AND WATER AT 550'F,

-THE INCREASE IN STORED ENERGY GRE'TLY INCREASES THE SEVERITY OF" l

A THE CONSEQUENCES OF.A RUPTURE OF THE PRESSURE BOUNDARY.

t 3.

HAVING THE CORE CRITICAL ALSO EXACERBATES THE CONSEQUENCES.-

o INSPECTORS AND OTHER PLANT PERSONNEL ARE IMPACTED DIRECTLY.

o PUBLIC RISK'IS INCREASED, BECAUSE THE RUPTURE COULD BE A

, l PRECURSOR OF AN ACCIDENT SCENARIO.

1 i

'.f C

o

pr~aw,r m-vn-sw-~~++~ ~-

-e,

~~-, m ~t

,~ ~,-~~ ~ c e ~.- ~ ni~- - ~,e,

,m,n J

, ~j L

44 e

e

-l

?t e

i t,

i a

k 8

t f

rr t

t

&t c

[

7

.t t

t t

t; 9

4 4

?

.t-c o

e L

i

,1 i

i 4 to Enclosure.3 un

, _ :~ --

-..,<.a.,~.+___

4.

April 29,1986-MEMORANDUM FOR: ' Albert F. Gibson, Director-Division of Reactor Safety Region !!

FROM:

Robert M. Bernero. Of rector kI trivision of BWR Licensing

.k SUQCT:

INTERPRETATION 0F SY$ TEM PRES $URE TEST REQUIREMENTS g

This respones to your memorandum dated April 17, 1986, requesting NRR'to'-

provide a position regarding the performance of system pressure tests (Leakage and Hydrostatic) using nuclear heating and a definition of-olant start-up with regard to pressure testing.

The ASME Code,Section XI. requires that the System Leakage Test be performed prior, to startup following each refueling outage (Table IWB-2500-1. Note 5). The examination method required is a visual, Type-

-VT-2.

Late

• additions of Section XI permit system hydrostatic. tests to be performell in lieu of the system leakage test. Further,'the Code states that reactor coolant shall be used as the pressurizing medium.

Also, the Code permits visual examination to be performed at 200*F with appropriate pressure corresponding to a pressure consistent with fracture prevention for system components that require a. test temperature above 200'F.

As you are aware. ASME Section XI-is incorporated by reference as part of NRC regulations.. Therefore, this is an issue whether regulations are being met, not just a simple code compliance-question.

The position.of the staff is that System Pressure Tests (Leakage and Hydrostatic) are to be performed before the reactor goes critical' from a refueling outage. The. System Leakage Test is a test to detennine-if any abnormal leakage is occurring in the reactor coolant pressure boundary after its opening and closing. The Hydrostatic Test is a proof test of repairs on the reactor coolant pressure boundary or other component. Prudence dictates that both of these tests be performed at the lowest temperatures that~ are consistent with the' fracture prevention criteria for the reactor vessel or other component so that storeo energy can be minimized daring testing conditions by having the system water solid. The temperature correction terms are provided to account for changes in material properities when the vessel'aust be _ heated for fracture provention. We do not believe the temperature corrections are.

an invitation to perform the testi'ig at higher temperature to minimize the test pressures. The cru suririna medium ic +a h __reactar coolant rather than steam. We~ recognize that, some-flashing to' steam of anyc IIotent1al leakage could occur when temperatures in excess of boiling are

~

necessary for the test.

Contact:

R. Hermann ll 1 27798'

/

fu 0 g)y I.jh y

,, A 4 1

l 4

3 r

l

. ~.

.. - ~

t

M 89 M A.-Gibson.

We believe the Code section allowing reduction of temperatures below 200*F.

at corresponding pressures is prudent for the visua* examination in that risk to plant personnel is reduced and any leakage would be liquid and, therefore, more readily detectable. Nith reprd to C MffMtion of start up,-- start-up occurs when the mode switch ;& r waa ia "a'a**% hot s

standby and controt roc wn...i is se pn. The ASME Code intended j g both Systes Leakage and Hydrostatic Tests to be performed prior to reactor

( ;

criticality from a refueling. Tha later Code position (footnote 7 to d>'

Table.!WB-2500-1, Category B P) Mfeitting the Systes Hydrostatic Test to be used in lieu of the Leakage Test is a clear indication that the code:

intended the systes hydrostatic also be performed =at low temperatures consistent with fracture prevention considerations prior.to reactor start-up.:

/ 4.

Robert M.:Bernero,-Director

' Division of BWR Licensing i

i cc:

W. Houston-T. Speis J

E. Rossi

0. Crutchfield.

G. Lainas S. D. Ebneter,'RI C. J. Paperiello, R!!!

E. H. Johnson, RIV D. Kirsch, RV

~,

B. O. Liaw R. Ballard C. Y. Cheng R. Hermann OISTRIBUTION:

M**S$tkB99970er BWR - Reading Files BWR - Generic Files 4

'..v'.....u 1'

s i-J 'i oPf pgg(.

3 OFC :BWR,

9

,, j

IE

...........'.......... y.

5...........;.........

F(iBWD0

-a g BW g

NAME$$HERMANN:cdd:E.Dillawl :W.F. Anderson: G.

inas

R. Bernero :

...........................................................,/2'l DATE :4/'.u/86

4/ s /86

4/ 4H/86 : 4/ ] /86 : */

OFFICIAL RECORD COPY

'