ML19282B022
| ML19282B022 | |
| Person / Time | |
|---|---|
| Issue date: | 01/08/1979 |
| From: | NRC COMMISSION (OCM) |
| To: | |
| Shared Package | |
| ML19282A997 | List: |
| References | |
| REF-10CFR9.7 NUDOCS 7903080621 | |
| Download: ML19282B022 (6) | |
Text
.
VALUE/ IMPACT STATEMENT ON LIMITED REVISION CF APPENDICES G AND H, FRACTURE TOUGHNESS AND SURVEILLANCE PROGRAM REQUIREMENTS I.
The Proposed Action A.
Description Three detailed technical requirements in Appendices G and H are being revised to delete restrictions that now appear unnecessary.
1.
Material toughness requirements for bolting are being deleted from the regulation, and ASME Code requirements are referenced instead.
2.
The specific limits on lead factor (the ratio of neutron flux at the surveillance capsule to that at the vessel wall at the peak fluence location) are being deleted.
3.
The prohibition of attachment of surveillance capsule holders to the vessel wall is being deleted, but a requirement is being added which directs that such attach-ment be treated by ASME Code rules for construction and inspection of attachments.
790308 Qi 1
B.
Need for Proposed Action The three technical requirements that are being revised by these amendments were part of a larger package of revisions, which have been in preparation for some time.
The need for changes had been identified in the course of more than 5 years of use of Appendices G and H.
In the case of item 1, above, ASME Code requirements have now been written which parallel those in the regulation and which we will reference to avoid duplication.
In the case of item 2, there is no safety consideration that requires exact limits on lead factor, thus we plan to delete the require-ment, because it affects design considerations.
In the case of item 3, we have concluded that the restriction can safely be lifted to permit more flexibility in design, provided ASME Code rules are followed as to the other details of construction and to inspection procedures.
The reason for separating the three items from the complete package of revisions is to speed up the amendment process and thereby save staff time in review of OL license applications.
As the following value/ impact statement shows, there are a number of such reviews to be completed in the coming year.
2
e It is expected that exemptions may be requested in many cases, and it is to avoid the need to grant such exemptions to requirements that we are in the process of revising that has prompted the proposed action.
Value/ Impact Statement A.
Manpower Savings The limited revision of Appendices G and H can be accomplished (with a 45 day public comment period), about a year earlier than the complete revision.
To quantify the estimates of savings in staff time for OL reviews, schedules were prepared for the two options.
About 15 plants in the OL review stage will be affected.
The estimated total manpower saving is about 20 man weeks for DSS review and management personnel, about 12 man weeks for NRC project and legal staff, and about 75 man weeks for applicants' assembly and evaluation of the information necessary to address compliance with Appendices G and H for these items.
Should staff positions or staff appearance be required to address these items at hearings, an additional 4 man-weeks for preparation and participation will be required per plant for NRC staff.
We also estimate that implementation of the proposed limited revisions to Appendices G and H would result in the NRC having to grant from 2 to 4 fewer exemptions per plant.
Projecting OL work over the next 12 to 15 months indicates that 30 to 60 fewer 3
exemptions will be required if the limited revision of Appen-dices G and H is accomplished promptly.
There should be little impact on the applicants' schedules for either of the proposed options.
With regard to the exposure of plant personnel to radiation, the addition of a specific requirement for inspection of the capsule holder attachments may impact some licensees, depending on what their past practices have been.
However, the requirement put in these amendments does not go beyond present ASME Code requirements.
B.
Safety Considerations It is our opinion that the changes proposed for Appendix G, paragraphs IV.A.3 and 4 have no safety impact, because they incorporate existing ASME Code rules that are identical to those deleted from Appendix G for bolts over 4 inch diameter, as reactor head bolts are.
We believe that the deletion of the requirement for specific limits on lead factor from Appendix H, paragraph II.C.2, is justified based on information gained from operating experience.
Sufficient data from plant surveillance programs have been 4
generated and adequate radiation damage estimating techniques are available (Regulatory Guide 1.99) to provide information to compensate for relatively small inaccuracies that may result from higher lead factors and to ensure that adequate safety margins are maintained.
Additionally, the restriction against welded attachment to the vessel wall can be eliminated, provided adequate precautions are taken in the design and fabrication of the welded attachment.
We have reviewed seve'ral OL applications that incorporate welded attachments to the vessel wall and have concluded that the attachments do not cause degradation of the vessel.
Implemen-tation of the indicated ASME Code rules will provide reasonable assurance that the vessel will maintain adequate safety margins.
Finally, there is some value in the elimination of the require-ment for a maximum lead factor of three and elimination of the restriction on welded attachments, Decause this action may, in some instances, increase the level of system integrity.
This conclusion is supported by recent operating experience on reactors in which the capsule holders were mechanically sup-ported from a flange area to comply with the restriction against welding and still provide the required lead factor.
Fluid flow between the capsule and the capsule holder tube during plant 5
c.
operation caused the holder tube to fail from vibration, and pieces of the capsule dropped into the vessel.
This resulted in shutdown of several plants.
The amount of radiation exposure that repair crews would have incurred in rebuilding the attachments precluded continuing the irradiation surveillance program at those plant sites.
6
AFFDrDcx 0 -Ps4cTema Torazarass Ra4csazuzarTo
- 27. FaaCTVm2 ToUeMKane a*QUIaEMEWre A. De pressure. retaining cornponenta of the reactor coolant pressure boundary that are made of ferittle Instartais shall meet the following nquiremerits for frseture tough.
neas during system hydratatte testa and any condition of normal operation. Incteding an.
ticipated operational occurrences:
- 1. De materials shall Eneet the acceptance standards of paragtsph NB-2330 of the ASMI Code, and the requirementa of sectiona IVA2. 3 and 4 and !YJ5. of this appendas.
- 2. For vessela, eactustee of botting or cGet fasteners:
- a. Calculated stress Intenalty factere shall be lower than the reference stress intensity factore by the margins spectSed in the ASME Code Appendix 0, " Protection Apinat Non.
Ductile Fallure'*. *1te ca!culation procedures shall comply with the procedures specited in the ASME Code Append 12 0, but additional and alternative procedures may be used if the Commission determines that they provide equivalent margins of safety against fracture.
maitng appropriate allowance for all uncer.
tainties in the data and analyses.
- b. For norzles, fiactes and shell regtons
- near georcetric d.tseontinutttes, the data and procedures required in additton to those spectN in the ASME Code shall provtdo mare of safety cornparable to those re.
cr.are or shells and heads remote from dia.
.ontinuitles.
- c. Whenever ther core is critical, the metal j temperature of the teactor vessel anall be high enough to provide an adequate margin of protection against fracture taking toto account such f actore as the potential ter overstress and thermal shoca during antics.
pated operational occurrences in the control of reactivity. In no caae when the core la critical (other than for the purpose of low.
level phystes tests) shall the tecipersture of 11 e reactor vessel be less than the mintmum permlasible temperature for the inservice syv.
tem hydrostatic pressure test nor less than
[ 40'F above that temperrture required by
, l ecction IV.Ah.
- d. If there is no fuel in the reactor during
. the initial preoperational system lettage and hydtcetatic pressure tests, the minimum per.
mtssible test tetnperature shall be deter.
mined in accordance with paragraph 02410 of the ASME Code except that the factor cd safety applied to each term making up the calculated stress Lntensity iactor may be re.
duced to 14. In no case shall the test tem.
perature be less than RTsev+60*F.
i. ;p
- shes
- 3. Matartsia tw ptptag f
. : : m. pum;. and valve.c.- _ 3,and
.v+--
2_
=
2-
- r_*s
- - - - " ; - ~ 7_ -; C:
...-o W,Matertala for botting and other f aa+.an.
ftheASMECode,paragraphsNB-2332and
.._2,.
_2
.re m -
- h*22 m"5 2*M*= "Ru r'='nta ofs-NB-2333, respactively,
k 1 Kl~ ~ %~Lw. ' _. ' '., J.G
] Z. C.'L_,__._. -.. _~ J._~. M. u. - _
,y
- ~ * = w-n rs t. ee
_-.-+-em newer.
- B. Reactor vessel beltline matertals shall have min! mum upper..helt etteriry, as deter.
Arrastz 1!-Rzacrom Vase:L Marsatat Stavan.Wres Paootau Raertazasants
- 3. rwraconcTsow he purpose cf the snatorial survettlance program requtrod by thta appendiz La to monitor changes in the fracture toughness properttee of ferrttte rnatorials in the re.
actor veuel bottline it en of water cooled power reactors reeu; tics frer4 their espesure to neution irradtation a.nd the thermal en.
vironment. Under this program, fracture toughnesa test data are obtained from mate.
rtal specimena withdrawn periodically frorn the reactor vessel. nese data wt:1 permit the determination of the condittona under which the vessel can be operated with ade.
. quate marstr.a et safety agattat fracture I throughout Sta sertice Itf e.
II. avavs2Laawrs raccaau ca;-taza A. No material eurve!!!ance pregram is re.
quired for reactor vessels for which is can be conservatively demonstrated by analyt.
leal methods, applied to expertmental data and teste performed on comparable vease e.
maktcg appropriate allowances for all un.
' certatntles in the measurements. that the peak neutron fluence (DIMeV)st the end of the design life of the vessel wul not escaed
/
10" n/cm'.
B. Reactor vessete constructed of ferritte materials which do not meet the conditions of section II.A. ahau have thett beltline re.
giona monitored by a surveulance preg-arn complying with the American Society for 74 sting and Matartats ( ASTM) Standard Rec.
ommended Practice for Survetnance Teste for Nuclear Reactor vessels. AST'.i Designation:
W 185-73.5 except me modtaed by this sp.
pendit.
C. ne surveulance program shall meet the following requirements:
- 1. Surveulance specimens sha!! be taken from locaticna alongside the fracture tous t.
nsaa test specimens required by section 1.1 of Appendiz O. ne specimen types s! Al comply with the requirementa 0:; sestion IIIA of Appendix 0 (except that drop we!ght specimena are not requtredt.
- 2. 8, :~eutancef capsules -- -
^J :
_ _ - - "' - ; n- :r sha!! be located near M --* -Seehed-ee the Inalde vessel wn!!
In the bet *Jine rtiton, so that the,, me--
\\
n....a ot gortsEth,guJo, t.,%g specimen ug ha that recets Wpaset inner surface) an pd.cai e,nvirarsen*%. cic as specimen irradiation history duplicates to 2.m. _
.m he design and location of the capsules shall the extent practicable, within the physical permli in.ersion of espisoement capauses.
^ecoimied tradauen,upe,'y constraints of the system, the neutron n _.., _ _, _... _. g spectrum, temperature history, and maximum
=---m ma, he us,d in a4diuon to the required numner of sur.
neutron fluence experienced by the reactor v.uianc. c.psulee specteed tn.ection rr.C1 vessel inner surface.
If the capsule holders
- 3. The required number of survettlance capsules and thett withdrawai nchedulu are are attached to the vessel wall or to the as fostown:
- a. For reactor vessets for whlen it can b.
vessel cladding, construction and inservice weld' ection of the attachments and attachm conservauvety demonstrated ny erperimental insD data and testa performed on comparable ves.
s shall be done according to the require-aet steet. maung appropriato nuovancer for an un..ertatattes in the meuurementa. that the acjusted reference temperature estab.
ments for permanent structural attachments,,
unbed.n accordance wtta action Ill.B. will to reactor Vessels given in the ASME Code, m not exceed 100*F at the end of the ee.wice urettme of the ructor vuus, at least three Sections III and XI.
surveulance capaules shall be provided for subsequent withdrawal as follows:
a r.doctave March 1.1973. Copies may be ob.
2Defined in Paragraph II. A. of Appendix G satned rrom th. Amancan society ror Test-
. ing and Matettals.1915 Race St. Phundet.
po 10 CFR Part 50.
pbia. ra. In03. eitn.r u m.eparata or <when avausble) ma part of the 1973 Annual Book of ASnt Standarsia Part 30 and also in Part St. Coptes are avausolo for inspect 9a at the Comminston e Puhite Document Room.1717 E St. NW. Washington. D.C.
g
DRAFT CONGRESSIONAL LETTER Enclosed for the information of the Subcommittee are copies of a Notice of Proposed Rule Making to be published in the FEDERAL REGISTER.
'The amendment of 10 CFR 50 comprises a limited revision of Appendices G and H, Fracture Toughness and Reactor Vessel Surveillance Requirements.
The purpose is to lift some restrictions that have proved to be unneces-sary and to delete some technical detail by substituting references to National Standards.
The amendment to Appendix G deletes the specific fracture toughness require-ments for bolting and substitutes similar requirements from the ASME Boiler and Pressure Vessel Coda (ASME Code). The amendment to Appendix H deletes restrictions on the location and method of attachment of material surveillance capsules in the reactor vessel and substitutes general requirements that reference the ASME Code.
The amendments would reduce significantly the number of exemptions to Appendices G and H in current applications for operating licenses, and thereby save considerable time and effort on the part of applicants and reactor vessel vendors and the NRC staff as well.