ML19332C076
| ML19332C076 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 11/07/1989 |
| From: | Long W Office of Nuclear Reactor Regulation |
| To: | Doherty J AFFILIATION NOT ASSIGNED |
| References | |
| RTR-REGGD-01.099 NUDOCS 8911220288 | |
| Download: ML19332C076 (3) | |
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L DISTRIBUTION
' Docket ho. 50-263.
JUidamd@
~ EJ0kDAti Mr. John F. Doherty, J.D.
PD31 GRAY FILE ECRIMES
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CHOLAHAN-ACRS(10)
Wellesley Hills, MA.- 0?l81 JZWOLlhSKI PSHUTTLEWORTH fl
Dear Itr. Doherty:
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SUBJECT:
TRAtlSMITTAL OF MONTICELLO NUCLEAR GENERATitiG PLANT LICEhSE At'Et4DMENT APPLICATI0ll i
The enclosures ere forwarded in response to your October 19th ittter to John Thona. Ycur letter requests a copy of iorth(rn States Power Company's
-application for amendment dated March 31, 1989.
b The license amendment was issued on November 11, 1909. The an.endment changed
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the Monticello Technical Specifications to:
4 (1) revise the teactor vessel pressure vs. temperature (P/T)' curves for consistency with Pevision 2 of Regulatory Guide 1.99;
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addrequirem(ntsforaugrentedinserviceinspectionofpiping);and susceptible tonintergranular stress corrosion cracting (IGSCC 1
(3) ievise the r(quirements for the periodic Type A containtient integrattt' leak rate test (CILRT) to permit the use of the mass poirt ir:st netaod approved by the Coradssion in a recent charge 10.10 3~
CFR Tart 50, Appendix J, as published in the IEDERAL REGISTER, on.
November 15, 1988 (53 FR 45981).
For your inforsation, I have also enclosec a copy of car Safety Evaluation for this arendment.
If your desire additioral informsticn, please contact me on (301) 40?-1323 between the hours of 7:00 a.m. and 4:00 p.m.
Sincerely, Original signed by William O. Long, Project Manager Project Directorete 111-1 Division of Reactor Projects - III, IV, V & Special Projects Office of Nuclear Reactor Regulation
Enclosures:
[
(1)
D. Musolf, Northern States Power Company letter dated March 31, 1989 (2) NRC Safety Evaluation cc:
See next pagt n\\
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UNITED STATES 3"
NUCLEAR REGULATORY COMMISSION j
g WASHINGTON, D. C. 20655 g
3 November 9, 1989 3..... J
. Docket No. 50-263 i
.Mr. John F. Doherty, J.D.
18 Maugus Avenue #6 Wellesley Hills, MA. 02181
Dear Mr. Doherty:
SUDJECT: TRANSMITTAL OF MONTICELLO NUCLEAR GENERATlhG PLANT LICENSE Al1ENDMEliT APPLICATIO!1 e
The enclosures are forwarded in response to your October 19th letter to John Thoma. Ycur letter requests a copy of flerthern States Power Company's application for amendment dated March 31,1989.
l The license amendment was issued on November 11, ic89. The an.endment changed l
the Monticello Technical Specifications to:
(1) revise the reactor vessel presrure vs. temperature (P/T) curves for consistency with Revision 2 of Regulatory Guide 1.99; add requirements for augmented inservice inspection of piping); and (2) susceptible tc intergranular stress corrosion cracking (IGSCC (3). revise the requirenents for the pertdic Type A containment integrated leak rate test (CILRT) to permit tlie use of the n. ass point test method approved by the Consiission in a recent change to 10 CFR Part 50, Appendix 0, as published in the FEDERAL REGISTER, on
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November 15, 1988 (53 FR 45981).
For your information, I have also enclosed a copy of our Safety Evaluation for this amendn;ent.
If your desire additict.a1 information, please contact me on (301)492-1323 between the hours of 7:00 a.m. and 4:00 p.m.
1 L
Sincerely.
Willian,O. Long, Project 'anager Projetf Directorate 111-1 Divisicn of Reactor Projects - III, IV, V & Special Projects Office of 14uclear Reactor Regulation
Enclosures:
(1)
D. Musolf,florthern States Power Company letter dated March 31, 15T9 (2) NRC Safety Evaluation cc:
See next page
-p.
c Mr. Thomas M. Parker Mon'ticello Nuclear Generating Plant Northern States Power Company CC:.
Cerald Charnoff, Esquire Shaw, Pittman, Potts and Trowbridge' 2300 N Street..HW Washington, D. C.
20037 U. S. Nuclear Regulatory Comission Resident Inspectors' Office Box 1200 Monticello, Minnesota 55362 Plant Manager Monticello Nuclear Generating Plant Northern States Power Company Monticello, Minnesota 55362 Russell J. Hatling Minnesota Environmental Control
. Citizens Association Energy Task Force 144 Melbourne Avenue, S.E.
Minneapolis, Minnesota 55113 Dr. John W. Ferman Minnesota Pollution Control Agency 520 Lafayette Road St. Paul, Minnesota 55155-3898 Regional Administrator, Region 111 U. S. Nuclear Regulatory Comission 799 Roosevelt Road Glen Ellyn, Illinois 60137
. Commissioner of Health Minnesota Department of Health 1
t 717 Delaware Street, S.E.
Minneapolis Minnesota 55440
- 0. J. Arlein, Auditor Wright County. Board of Comissioners t
10 NW Second Street Buffalo, Minnesota 55313-i
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p 414 Nicoliet Matt Minneapohs Minnesota MADI Telephone (6th 330 5500 i
March:31,.1989 10 CPR Part 50 j
Section 50.90 j
Director of Nuclear Reactor Regulation U S Nuclear Regulacory Commission Attn:
Document Control Desk >
Washington, DC 20555-t MONTICELLO NUCLEAR GENERATING P1 ANT Docket No. 50 263 License No. DPR 22 l'
License Amendment Request Dated March 31, 1989 Regulatory Guide 1.99. Revision 2. and Miscellaneous Changes e
Attached is a request for a change in the Technical Specifications, Appendix A of the Operating License, for the Monticello Nuclear Gener-ating Plant.
This request is submitted in accordance with the pro-visions of 10,CFR'Part 50, Section 50.90.
This request propuses changes in three areas:
- a. Revised pressure temperature limit curves to meet the requirements of Regulatory Guide 1.99 Revision 2,
" Radiation Embrittlement of Reactor Vessel Materials."
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The proposed changes are submitted in response to e
NRC Generic Letter 88 11.
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- b. A new tequirement for augmented inspection of piping susceptible to intergranual stress corrosion cracking
-(ICSCC) as required by NRC Ceneric Letter 88 01.-
- c. A change in the requirements for Type A contalument leakage rate tests to permit the use of the " mass point" method as approved by the Commission in a recent change to 10 CFR Part 50, Appendix J.
Exhibit A contains a description of the proposed changes, the reasons for requesting the changes, and a Significant Hazards Determination.
' Exhibit B contains pages of the Monticello Technical Specifications
- revised to show the proposed changes.
Retyped pages are included in Exhibit C.
Ceneral Electric has prepared a report which supports the propon.'
I changes in the Monticello pressure temperature limitation curves.
A copy of this report, SASR 88 99, Revision 1, January, 1989, is attached as Exhibit D.
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Direct':r cf NRR' March 31, 1989' Northom States Power Company j
Page 2 1
Please cantact us if you have any questions related to this request or if addittenal information is required to supplement it.
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David Musolf Manager Nuclear Support Services c: Regional Administrator III, NRC Sr Resident inspector, NRC NRR Project Manager, NRC i
G Charnoff Minnesota Pollution Cont:o1 Agency (State Contact)
I Attn: Dr J W Ferman Attachments i
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r UNITED STATES NUCLEAR REGULATORY COMMISSION NORTHERN STATES POWER COMPANY MONTICELLO NUCLEAR CENERATING PLANT DOCKET NO. 50 263 REQUEST FOR AMENDMENT TO OPERATING LICENSE DPR 22 LICENSE AMENDMENT REQUEST DATED MARCH 31, 1989 Northern States Power Company, a Minnesota corporation, requests auth-orization for changes to Appendix A of the Monticello Operating Li-cense as shown on the attachments labeled Exhibits A, B, C, and D.
Exhibit A describes the proposed changes, reasons for the changes, and includes significant hazards evaluations.
Exhibit B is a copy of the Monti:ello Technical Specifications incorporating the proposed changes. Exhibit C are retyped pages of the Monticello Technical Specifications including the proposed changes.
Exhibit D is a report prepared by the General Electric Company in support of some of the proposed changes.
This letter contains no restricted or other defense information.
NORTHERN STATES POWER COMPANY By Dd 9Mu a David Musolf
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Manager Nuclear Support Services on this 8 M day of f b d
, M P @ before me a notary public in and for said County;-personally appeared David Musolf, Manager Nuclear Support Services, and being first duly sworn acknowledged that he is authorized to execute this document on behalf of Northern States Power Company, that he knows the contents thereof, and that to the best of his knowledge, information, and belief the statements made in it are true and that is is not interposed for delay.
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Exhibit A Monticello Nuclear Generating Plant License Amendment Request Dated March 31, 1989 O
Description and Evaluation of Proposed Change to Appendix A of Operating License DPR 22 l
. Pursuant to 10 CFR Part 50, Section 50.90, the holders of Operating '
L License DPR 22 hereby propose the following changes to Appendix A, l'
' Technical Specifications:
- 1. Compliance With Regulatory Guide 1.99, Revision 2 Pressure Temperature Limitation Curveh l
Proposed Changes 1
Revise the wording of Specifications 3,6.B.1, 3.6.B.2, 3.6.B.3, 4.6.B.1, and 4.6.3.2 en page 122 as shown in Exhibit B.
Revise the bases on pages 145 and 146 as shown in Exhibit B.
Replace Figures 3.6.1 through 3.6.4 on pages 133 136 with the re-vised figures shown in Exhibit B.
Revise the titles of the figures on List of Figures page "v" as shown in Exhibit B.
Reason for Changes General Electric has' revised the pressure tamperature limitation
. curves for Monticello to conform to the requirements of Regulatory Guide 1.99, Revision 2.
The new analysis also removes an overly con-servative assumption for the initial RTNDT temperature of weld mate-rial in the vessel beltline region and allows for separate temperature monitoring of the vessel beltline and bottom head regions.
Refer to Genera 1' Electric Report SASR 88 99, Revision 1, ' January,1989 "Imple-mentation of-Regulatory Guide 1.99, Revision 2 for the Monticello Nuclear Generating Plant," which is provided in Exhibit D.
l L
The proposed Technical Specification changes on pages "v", 122, 133-j 136, and 145-146:
- a. Revise the List of Figures to reflect new titles for Fi ures
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6 3.6.1 through 3.6.4.
- b. Revise the pressure temperature limit curves to conform to Regulatory Guide 1.99. Revision 2, and the latest General Electric procedure,
- c. Add a new requirement to monitor vessel beltline temperature as well as bottom head temperature during vessel hydrostatic and leahage tests.
This permits the cooler botto= head tem-perature to be compared to the less limiting "RPV Remote from Core Beltline" limits of Figure 3.6.2 during vessel testing.
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[w EXHIBIT A 2
- d. Clarify the Technical Specifications to note that analysis of l'
the first vessel specimen capsule has been completed.
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- e. Add a clarifying statement in the bases that small rates of temperature change during vessel hydrostatic and leakage testing do not affect the permissible pressure temperature operating range,
- f. Update the basis to describe the General Electric procedure for determining the new pressure temperature limitations.
l Safety Evaluation and Determination of f'
Sis:nificant Hazards Considerations i
The proposed changes to Appendix A of the Operating License have been evaluated to determine whether they constitute a significant hazards considerr.tien as required by 10 CTR Part 50, Section 50.91 using the standards provided in Section 50.92. This evaluation is provided below:
- 1. The proposed amendment will not involve a significant increase in the probability or consequences of an accident previously evaluated.
The proposed amendment would revise the Technical Specifications to conform to current NRC requirements for protection of the reactor pressure vessel from radiation induced embrittlement. A number of minor clarifications and updates to the Technical Specifications are also proposed.
The proposed Technical Specification changes result in more conservative operating limits for the reactor ves-sel. The new operating limits will provide increased margins of protection for the reactor vessel from non ductile failure.
Failure of the reactor vessel is not a design basis accident.
Through design conservatisms, reactor vessel failure has a vanish-ingly low probability of occurrence and is not considered in the safety analyses. The new proposed pressure temperature operating limits will add additional conservatism making reactor vessel fail-ure even less credible. Therefore, this change cannot increase the probability or consequences of any previously evaluated accident.
- 2. The proposed amendment will not create the possibility of a new or different kind of accident from any accident previous 1v evaluated.
l l
The proposed Technical Specification changes deal exclusively with l
reactor vessel pressure-temperature limitations. No other compo-l nent or plant system is affected.
The new pressure temperature limit curves will be adjusted for reactor vessel fluence in a more conservative manner than the existing curves.
Therefore, there is
7 EXHIBIT A 3
no possibility of a new or different kind of accident being created from any accident previously evaluated.
- 3. The proposed amendment will not involve a significant reduction in the earnin of safety.
The proposed Technical Specification changes will increase margins of safety by adjusting reactor vessel pressure temperature limitation curves in a more conservative manner. The proposed changes conform fully to the recommendations of the NRC Staff contained in Regulatory Guide 1.99, Revision 2.
Therefore, the proposed changes will not reduce marg' ins of safety, but will increase them.
The Commission has provided guidance'concerning the application of the Standards for determining whether a significant hazards consideration exists by providing certain examples of amendments that are considered not likely to involve significant hazards considerations.
These ex3m-pies were published in the Federal Register en March 6, 1986.
Changes proposed in this Licer.se Amendment Request are representative of example (vii).
They are related to a change in the Technical Specifications to conform to changes in NRC regulations where the change results in very minor changes to facility operations clearly in keeping with the regulations.
As previously described, the Technical Specification changes reflect the guidance contained in NRC Regulatory Guide 1.99, Revision 2.
The overall effect of the changes is to improve plant safety by increasing margins of protection against'non ductile failure of the reactor ves-sel.
- 2. Augmented Inspection of Piping Suscoptible to Intergranual Stress Corrosions Cracking (ICSCC)
Proposed Chances Add new Specification 4.15.A.2 as follows:
- 2. Welds in austenitic stainless steel piping four inches or larger in nominal diameter containing reactor coolant at a temp-erature above 200 degrees F during power operation, including reactor vessel attachments and appurtenances, not meeting the requirements of NUREG 0313, Revision 2, for IGSCC Category A weldments shall be included in an augmented inspection program meeting the requirements of NUREG 0313, Revision 2.
Refer to Exhibit B, page 229f and new page 229ff.
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l EXHIBIT A
.4 Reason for Changes l
This chan5e is required to implement the requirements of NRC Ceneric Generic 1 Letter 88 01 at the Monticello Nuclear Generating Plant.
l Letter 88 01 required Technical Specification changes to be proposed by licensees for augmented inspections of piping systems which may be susceptible to ICSCC damage.
As noted in our letter dated July 28, 1988, substantially all piping welds. susceptible to ICSCC have been eliminated. Two susceptible welds remain. The proposed Technical Specification change will re-quire these welds to be inspected more frequently in accordanca with the requirements of NUREG 0313. Revision 2, " Technical Report on i
Material Selection and Processing Guidelines for WR Ccolant Pressure boundary Piping.'
Safety F.'ea3u.stion and D7 termination of Significatst Untards Considerations The proposed changes to Appendix A of the Operating 1.icense have been evaluated to determine rhether they constitute a significant hazards consideration as required by 10 CFR Part 50, Section 50.91 using t.hs standards provided in Section 50.92. This evaluation is provided below:
- 1. The proposed amendment will not involve a significant increase in the probability or consequences of an accident previous 1v evaluated.
The proposed amendment would revise the Technical Specifications to i
add new requirements for augmented inspection of piping susceptible to ICSCC. These inspections conform to NRC requirements contained in Generic Letter 88 01 and NUREC 0313, Revision 2.
They provide additional assurance that piping degradation due to ICSCC will be detected prior to failure. Probability of pipe break accidents are significantly reduced through this program of augmented inspection.
The probability or consequences of previously evaluated pipe break accidents are therefore reduced by the proposed change.
- 2. The proposed amendment will not create the possibility of a new cr different kind of accident from any accident previous 1v evaluated.
The proposed Technical Specification changes add additional requirements for inspection of piping susceptible to ICSCC. Only pipe-break accidents are affected by the proposed change, and the proposed change will reduce the possibility of this type of accident. No other plant components or systems are affected.
Therefore, there is no possibility of the creation of a new or different type of accident.
EXH181T A 5-
- 3. The proposed amendment will not involve a significant reduction in the martin of safety.
The proposed Technical Specification changes will increase margins of safety by providing for augmented inspection of piping suscep.
tible to ICSCC degradation. The proposed changes conform fully to the recommendations of the NRC Staff contained in Generic Letter 88-01 and NUREC 0313 Revision 2.
Therefore, the proposed changes will not reduce margins of safety, but will increase them.
The Commission has provided guidance concerning the application of the Standards for determining whether a significant hazards consideration exists by providing certain examples of amendments that are considered not likely to involvs significant harards considerations. These exsm-pies were published in the Federal Register on March 6, 1985.
Changes proposed in this License Amer.dment Request ara reptasentative of example (vii). They are related to a change in the Technical Specifications to conform to changes in NRC regulations, where the change results in very minor changes to facility operations clearly in keeping with the regulations.
As previously described, the Technical Specification changes reflect the guidance contained in NRC Ceneric Letter 88 01 and NUREC 0313 Revision 2.
The overall effect of the changes is to improve plant safety by increasing margins of protection against non ductile failure of the reactor vessel.
- 3. Use of Mass-Point Method for Type A Containment Leak Rate Tests Proposed Changes Revise Specification 4.7.A.2.b to read
- b. The primary containment leakage rates shall be demonstrated at the following test schedule and shall be determined in conformance with the criteria, methods, and provisions of Appendix J of 10 CFR Part 50:
l Refer to Exhibit B page 159.
Reason for Change This change deletes.the specific reference to ANSI N45.4-1972 fer the Type A containment leak rate test precedure.
Recent changes to Appen-p
EXHIBIT A 6
dix J to 10 CFR Part 50 published it the Federal Register on November 15, 1988, recognize the mass point method of ANSI /ANS 56.8 1981 as bein5 an acceptable procedura for tests having a duration of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> The mass point method is a superior method of computing or more.
overall containment leak rate. Adopting this change will revise the Technical Specifications to permit the use of the new and superior methodology.
i The revised Specification will simply reference the requirements of Appendix J instead of specifying a particular test procedure.
Safety Evaluation and Determination of
- Significant Hazards Considerations r
The propored changes to Appendix A of the Operating License have been 3
eveluated to determine whether they constitute a significant hazards censideration as required by 10 CFR Part 50, Section 50.91 using the standards proeided in Section 50.92. This evaluation is provided below:
- 1.,The proposed amendment will not involve a significant increase in the probability or consequonees of an accident previousiv evaluated.
The proposed amendment would revise the Technical Specifications to recognize recent changes in Appendix J to 10 CFR Part 50.
New and superior methodology can be used to compute integrated containment leak rate during during Type A leak rate tests.
Increase assurance of containment integrity will be provided by using the new methodology.
The proposed change involves only a containment leakage rate test procedure. No analyzed accident is affected by the change,
'2. The proposed amendment will not create the possibility of a new or different kind of accident from any accident previousiv evaluated.
The proposed Technical Specification changes deal exclusively with containment leak rate testing procedures.
No plant components or plant systems are affected. Therefore. there is no possibility of a new or different kind of accident being created from any accident previously evaluated.
- 3. The proposed amendment will not involve a significant reduction in the margin of safety.
The proposed Technical Specification changes will increase margins of safety by allowing the use of a new and superior method of com-puting c:ntcin=ent integr:ted Icahage rate. The preposed change a~
a
EXHIBIT A 7-l a
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will conform the Technical Specifications to recent changes in j
Appendix J to 10 CFR Part 50.
l The Commission has provided guidance concerning the application of the Standards for determining whether a significant hazards consideracion exists by providing certain examples of amendments that are considered not likely to involve significant hazards considerations.
These exam.
pies were published in the Federal Register on March 6, 1986.
7 a
Changes proposed in this License Amendment Request are representative of example (vii). They are related to a change in the Technical Specifications to conform to changes in NRC regulations, where the change results in very minor changes to facility operations clearly in keeping with the regulations.
I As previously described, the Technical Specification changes reflect tho 'geidance contained in Appendix J to 10 CFR Part 50. Ths overall effect of the changes is to improve plant safety by improving containment leak rate test procedures.
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i Exhibit B 1,
License Mendment Request Dated March 31, 1969 Docket No. $0 263 License No. DPR 22 Exhibit B consists of revised 'pages for the Monticello' Nuclear Gener.
ating Plant Technical Specifications annotated to show the proposed wording changes:
Pages: y 122 133 134 135 136 145 l*
146
- c59 219ff (new page) 3
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Firure No.
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4.1.1 T Tactor Craphical Aid in the Selection of 44 an Adequate Interval Between Tests 4.2.1 System Unavailability 75 i
3.4.1 Sodium Pentaborate Solution Volume. Concentration 97 Requirements 3.4.2 Sodium Pentaborate solution Temperature Requirements 98 t
.5 $.1 Single Loop operation Surveillances Fever /Tiow Curve 114b N W,m 133 3.6.1 re Beltline Operating Limits Curve Adjustment P
.vs. Fluence 134 3.6.2 Minimum Temperature vs. Pressure for Pressure Tests 135 3.6.3 Minimum Temperature vs. Pressure for Mechanical Heatup or Cooldown Without the Core Critical l
136 3.6.4 Minimum Temperature vs. Pressure for Core Operation 4.6.2 Ch or e Stress Corrosion Test Results (d 500'T 137 3.7.1 Differential Pressure Decay Between the Drywell 191 and Verve 11
-l 3.8.1 Monticello Nuclear Generating Plan: Site Boundary 1968 for Liquid Effluents 3.6.2 Monticello Nuclear Generating Plan: Site Boundary 19th for caseous Effluents j
3.11.1 MAPTAC Limits 215a p
3.11.2 MAPPAC Limits 215b 7
3.11.3 Power Dependent MCPR Limits 215c 3.11.4 MCPR Limits 215d T
I 6.1.1 NSP Corporate Organizational E,elationship to On Site 234 Operating Organization 6.1.2 Monticello Nulear Generating Plant Functional 235 Organization for on Site Crperating Group 9 C C,,
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4.0 S11RVEILLANCE REQUIREMENTS 1.0 LIMITING CONDITIONS FOR OPERATION ReactorVesselTeheratureandPressore B.
11 Reactor Vessel Temperature and Pressure 1.
During in-service hydrostatic or leak 1.
During in-service hydrostatic or leak test-testing when the vessel pressure is ing, the reactor vessel shell temperatures above 312 psig, the following temper-specified in 4.6.B.1, except for the reactor atures shall be recorded at least every vessel bottom head, shall be at or above the 13 ninutes.
l F ure temperatures shown on the two e ryes, "RPV-Core Be t p1TMe,3.6 g 5 i ichqe the dashed curve Reactor vese*1 shell adjacent 1
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i core beltline a.
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( t emperature adjustment from Figure 3.6.1.
The to shel3 flange.
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teactor vessel bottom head terperature shall b.
Reactor vesse) bottom head.
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be at or above the temperatures shown on the lid curve of Figure 3.6.2, "RPV Remote from Reactor vessel shell or coolant i
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Core Beltline," with no adjustment f temperature representative of the mini-r
\\ Figure 3.6.1.
mum temperature of the beltline re ion.
2.
Test specimens representing the reactor vessel, 2.
During heatup by non-nuclear means bcse weld, ar.d weld heat affected zone metal (except with the reactor vessel shall be installed in the reactor vessel vented), cooldown following nuc1 car adjacent to the vessel wall at the core mid-shutdown, or low level physics tests plane level. The matetial semple program the reactor vessel shell and fluid shall conform to I.STM F. 185-66. Samples shall
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temperatures specified in 4.6.A shall te withdrawn at one fourth and three fourths he at or above the higher of the service lifc.
Analysis of the first sample
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temperatures of Figure 3.6.3 where the G.1te.Jetps=f== tin of gj dashed curve, "RPV Core Beltline "
shall includ a os.t a
--i is increased by the expected shift in t e materini c 2s strice.
(Note: Analysis o the first semple has been completed. The g ",3[*')
2 ItT from Figure 3.6.1.
Figure 3.6.1 core beltline temperature adjust-NDT nt curve reflects the chemistry da a bt ined).
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- 3. Ituring all operation with a critical reactor, other than for low level 3.
Neutron flux v res siall be installed
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3 physics tests or at times when the in the reactor vessel adjacent to the I --(jp],{;
1 reactor vessel is vented, the reactor reactor vessel wall at the core mid-vessel shell and fluid temperatures plane 14 vel. The wires shall be removed
-I1 specified in 4.6.A shall be at or and tested during the first refueling
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above the higher of the traperatures cutage to experisentally verify the
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of Figure 3.6.4 where the dashed curve, calculated value of neutron fluer.ce at "RPV Core Beltline," is increased one fourth of the helt11ne shell thickness from by the expected shift in RTHDT that is used to determine the NDTT Figure 3.6.1.
shift t' rom Figure 3.6.1.
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A General Electric Company procedure, designed to evaluate fracture toughness requirements for older values on an equivalent basis 7
plants where information may be incomplete, was used to estirsate RTNDT to the new requirements for plants which have construction permits af ter August 15, 1973.
j t
V m
145 1.6/4.6 MASES
- 4-4 y-
/
' Itis e s 3.6 and 4.6 - Continued:
to fast fracture toughness of all ferritic steels gradually and uniformly decresses with exposures The neutrons above a threshold value, and it is prudent and conservative to &ccount for this in the operation of the reactor pressure vessel. 'Two types of information.are needed in this analysis: 1) A relationship hetween the changes in fracture toughness of the reactor pressure vessel steel and the neutron fluence
{
(Integrated neutron flux), and 2)
A measure of the neutron fluence at the point of interest in the reactor pressure vessel wall.
relationship of predicted adjustment of reference temperature versus fluence and the copper and nickel The content of the core beltline materials give in regulatory Guide 1.99, RevIs! >n 2, was used to define the f
core heitline temperature adjustment versus fluence shown on Figure 3.5.1.
l A relationship between full power years of operation and neutron fluence has been experimentally the reactor vessel. The vessel pressurization temperatures at any time period can be determined for determined from the thermal energy output of the plant and Figure 3.6.1 ured in conjunction with Figure 3.6.2 (pressure tests) Figure 3.6.3 (mechanical beatup or cooldown following nuclear shutdown), or Figure d
2
(
During the first fuel cycle, only calculated neutron fluence 3.6.4 (operation with a critical core).
At the first refueling, neutron dosimeter wires which were installed adjacent to the s
vessel wall were removed to experimentally determine che neutron fluence versus full power years of values were used.
first This experimental result was updated by testing additional dosixtry removed with the g
l operation.
surveillance capsule.
,C J..)
Iteactor vessel material samples are provided, however, to verify the re.lationship expressed by Figure Three sets of mechanical test specimens representing the base metal, weld metal, and weld heat
~
3.6.1.
{U C affected zone (IIAZ) metal have been placed in the vessel end can be removed and tested as required.
An Ms analysis and report will be submitted to the Commission on all such surveillance specimens removed from the reactor vessel le accordance with 10 CFR 50, Appendix II, lacluding information obtained on the level of Integrated fast neutron irradiation received by the specimens and actual vessel material.
W I
I Q.3 I
- L, f%.
146 1.6/4.6 I!ASES
3.0 LIMITitiG CO?IDITIOllS FOR OPERATIOfi 4.0 ANCP ENT N
The primary containment leakage rates h.
When Primary Cor:tainment Integrity is required, 9.
leakage rates shall be limited to:
shall be demonstrated at the following test schedule and shall be determined 1.
An overall integrated leakage rate of less in confermancM with the criteria, methods, than or equal to fa, 1.2 percent by weight and provisions of 10 CFR Part 50:
of the containment air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at Pa, 42 psig.
d 1.
Three Type A overall integrated 2.
A combined leakage rate of less than or equal containment leakage rate tests shall to 0.6La for all penetrations and valves, be cenducted at 40110 month intervals
- except for main steam isolation valves, duting shutdown at >Fa during each subject to Type B and C tests when pressurized 10-year service period. The third to Pa.
test of each set shall be conducted during the shutdown for the year 3.
imss than or equal to 11.5 scf per hour for plant inservice inspection.
any one main steam isolation valve when tested at 25 psi.
2.
If any periodic Type A test fails to meet 0.75La, the tsst schedule for Ulth the measured overall integrated primary subsequent Type A tests shall be containment leakage rate exceeding 0.75La, or the reviewed an1 approved by the Commission.
measured combined leakage rate for all penetrations If two consecutive Type A tests fail to and valves, except main steam isolation valves, meet 0.75La, a Type A test shall be la performed at least every 18 months until nJ subject to Type B and C testing ' exceeding 0.6La, or
___. J the measured leak rate exceeding 11.5 scf per hour two consecutive Type A tests meet 0.75La,
.-- for any one main steam isolation valve, restore at which time the above test schedule I
leakage rates to less than or equal to these values may be resumed.
p--
I : ' {
prior to increasing reactor coolant system temper-ature above 212* F or, alternatively, restore 3.
All Type A test leakage rates shall be measured leakage rates to within these limits within calculated using observed data converted Cta-one hour or be in at least flot shutdown within the to absolute values. Error analyses next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in Cold Shutdown within the shall be performed to select a balanced g
y following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Integrated leakage measurement system.
L1 I~
- D:e second test of the second 10-year service period may be conducted during the 1989 refueling outage.
159 1.7/4.7 6
u-__
_ _ _ _ _. ~ _. _,
2 s
3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILIANCE REQUIREMENTS f
/2. Welds in austenttic stainless steel piping P
four inches or larger in diameter containing reactor coolant at a temperature above 200 degrees F during power operation, including reactor vessel 3
attachments and appurtenances, not meeting the requirements of NUREC-0313, Revision -
4 2, for IGSCC Category A weldments shall be included in an augpeented inspection program meeting the. requirements of NUREG 0313, Revision 2.
l n
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C)o G_Q i
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L 3.15/4.15 g gg
Exhibit C License Amendment Request Dated March 31, 1989 Docket No. 50 263 License No. DPR 22 Exhf. bit c consists of retyped pages for the Monticello Nuclear Gener-i
' ating Plant Technical Specifications showing the final form of the proposed wording changes:
Pages: v 122 133 134 135 136 145 146 159 l
229ff (new page) h n
9 4
i B
B t
.9 r
r;-
UET.L.
O
{
1.IST OF T7 CURES.
'Tirure No.
Pare No.
4.1.1 "M" Factor. Craphical Aid in the Selection of 44 an Adequate Interval Between Tests 4.2.1 System Unavailability 75 3.4'.1 Sodium Pentaborate solution Volume Concentration 97 Requirements 3.4.2 Sodium Pentaborate Solution Temperacure Requirements 98 l
.3.5.1 Single Loop Operation Su illances Power / Flow Curve 114b 3.6.1 Core Beltline Ope. rating Limits curve Adjustment 133 vs. Fluence 134 l
3.6.2 Minimum Temperature vs. Pressure for Fressure Tests 135 3.6.3
' Minimum _ Temperature vs. Pressure'for Mechanical Heatup or Cooldown Without the Core Critical 136 3.6.4 l Minimum Temperature vs, Pressure for Core Operation 4.6.2 Chloride Stress Corrosion Test Results @ 500*F 137 j
1 3.7.1 Differential Pressure Decay Between the Drywell 191 and Wetwell 3.8.1 Monticello Nuclear Generating Plant Site Boundary 196g for. Liquid Effluents
=3.8.2 Monticello Nuclear Generating Plant Site Boundary 198h for Caseous Effluents 3.11.1 MAPFAC Limits 215a p
i 3.11.2 MAPFAC Limits 215b
{
F 3.11.3 Power Depe1 dent MCPR Limits 215e 3.11.4 MCPR Limits 215d p
6.1.1 NSP Corporate Organitational Relationship to On-Site 234 Operating Organitation
-6.1.2 Monticello Nuclear Generating Plant Functional 235 Organitation for On-Site Operating Group i
V
=. _
3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILIANCE REQUIREMEKfS B.
Reactor Vessel Temperatore and Pressure B.
Reactor Vessel Temperature and Pressure 1.
During in-service hydrostatic or Icak test-1.
During in-service hydrostatic or leak Ing, the reactor vessel shell temperatures testing when the vessel pressure is specified in 4.6.B.1,'except for the reactor above 312 psig, the following temper-vessel bottom head, shall be at or above the atures shall be recorded at least every temperatures shown on the two curves of Figure 15 minutes.
3.6.2, where the dashed curve, "RPV Core Belt-a.
Reactor vessel shell adjacent line," is increased by the core beltline temperature adjustment f rom Figure 3.6.1.
The to shell flange.
reactor vessel bottom head temperature shall be at or above the temperatures shown on the b.
Reactor vessel bottom-head.
solid curve of Figure 3.6.2, "RPV Remote from c.
Reactor vessel shell or coolant core Beltline," with no adjustment from temperature' representative of the mini-l Figure 3.6.1.
enus temperature of the beltline region.
1 2.
During heatup by non-nuclear means 2.
Test specimens representing the reactor vessel, (except with the reactor vessel base weld, and weld heat affected zone metal vented), cooldown following nuclear shall be installed in the reactor vessel' shutdown, or low level physics tests adjacent to the vessel wall at the core mid-the reactor vessel shell and-fluid plane level. 1ms material sample program shall conform to ASTM E 185-66. ' Samples shall temperatures specified in 4.6.A shall be withdrawn at one fourth and three fourths be at or above the higher of the temperatures of Figure 3.6.3 where the service life. Analysis of the first sample dashed curve, "RPV Core Beltline,"
shall include a quantitative determination of is increased by the expected shift in the material chemistries.
(Note: Analysis of RT fr m Figure 3.6.1.
the first sample has been completed. The NDT Figure 3.6.1 core beltline temperature adjust-
- 3. During all operation with a critical ment curve reflects the chemistry. data obtained)._
reactor, other than for low level physics tests or at times when the 3.
Neutren flux wires shall be installed reactor vessel-is vented, the reactor in the rer,etor vessel adjacent to the vessel shell and fluid temperatures reactor vessel wall at the core mid-specified in 4.6.A shall be at or plane level. The wires shall be removed above the higher of the temperatures and tested during the first refueling of Figure 3.6.4 where the dashed curve, outage to experimentally verify the "RPV Core Beltline," is increased calculated vante of neutron fluence at fr m ne f urth of the beltline shell thickness-by the expected shift in RTNDT that is used to determine the NDTT Figure 3.6.1.
shift from Figure 3.6.1.
122 3.6/4.6 i
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A.. Iteact or Coolant llentup and Gooldown HeatingLandL The vessel has been analyzed for stresses caused by thermal and pressure transients. idered in cooling transients throughout plant life at uniform rates. of 100*F per hour were cons for stress-temperature range of 100.to 546* F and were shown to be within.the requirements lutensity and fatigue limits of Section' III of -the ASME Boller and Pressure Vessel Code..
the td During reactor operation, the temperature of the coolant in an idle recirculation loop.is expec eRequir
~
reactor coolant temperature unless it is valved out of' service.
in an idle loop to be within 50*F of the reactor coolant temperature before the pump is.
to remain at temperature at the reactor vessel nozzles and bottom temperature statted assures that the change in coolant head region are within the conditions analyzed for the reactor vessel thermal and pressure transients.
i d has a During hydrostatic pressure testing, a coolant heatop or cooldown of 20*F in any one-hour per o negligible ef fect on the reactor operating limits of Figure 3.6.2.
Itenctor Vessel Temnerature and Pressure n.
?
d cooldown Operating limits on the reactor vessel pressure and temperature during normal heatup ant 50, Appendix G nnd during inservice hydrostatic testing were established using 10 CFR Par iler and Pressure Vessel and Appendix G of the Summer 1916 or later Addenda to Section III cf the A 24 inches ll other reactor vessel -
at t.he flange-to-vessel junction and one-quarter of the material thickness at aFor the purpose of setting these operating Code.
~
locations and discontinuity regions can be safety accommodated.HDT, of the vessel mate 965 limits the reference temperature, RTin accordance with requirements of the code to which this vess Edit:f on including' Summer 1966 Addenda).
for older A General Electric Company procedure, designed to evaluate fracture toughness requirementsvalues on an equi plants where information may be incomplete, was used to estimate RTNDT August 15, 1973.
to the new requirements for plants which have-construction permits after t
t 4
145 1.6/4.6 BASES L
e m
m-
.m s
<me
.e 4
s nases 3,6 and 4.6 - Contintted:
fast The fracture toughness -of all ferritic steels gradually and uniformly decreases with' exposure to.for this in the operat neutrons above a threshold value, and it is prudent and conservative to account
- 1) A relationship Two types of information are needed in this analysis:-
of the reactor pressure vessel.
between the changes in fracture toughness of the reactor pressure vessel' steel and the neutron fluence A measure of the neutron fluence at the point of interest ~ In the (int egrated neutron flux), and 2) reactor pressure vessel wall.
d nickel; The relationship of predicted adjustment of reference temperature versus fluence and the t
content of the core heltiine materials give core beltline temperature adjustment versus fluence sl.own on Figure 3.6.1.
i t lly A relationship between full power years of operation and neutron fluence has been exper men determined for the reactor vessel. thermal energy output of the plant and Figure 3.6.1 used in conjunction wit h Figure i
3.6.2 (pressure tests), Figure 3.6.3 (mechanical heatup or cooldown following nuclear -shutdown),
determined from the During the first fuel cycle only calculated neutron fluence 3.6.4 (operation with a critical core).At the first refueling, neutron dosimeter wires which were installed adjacen full power years of values were used.
vessel wall were removed to experimentally determine the neutron fluence versu operation.
surveillance capsule.
by Figure Reactor vessel material samples are provided, however, to verify the relat 3.6.1.
d as required. An affected zone (llAZ) metal have been placed in the vessel and can be removed and teste l
i ens removed from analysis and report will be submitted to the Commission on all such surveil ance spec minformation obtained on the the reactor vessel in accordance with 10 CFR 50, Appendix 11, including l vessel material.
of integrated fast neutron irradiation received by the specimens and actua 4 3 1
146 3.6/4.6 BASF.S b
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4.0 SURVEILIANCE REQUIREMENTS 3.0 LIMITING CONDITIONS FOR OPERATION b.
The primary containment leakage rates h.
When Primary Containment Integrity is' required, shall be demonstrated at the following leakage rates shall be limited to:
test schedule and shall be determined in conformance with the ' criteria, methods, 1.
An overall integrated leakage rate of less and provisions of 10 CFR Part 50:
than or equal to La, 7.2 percent by weight of the containment air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at Pa, 42 psig.
1.
Three Type A overall integrated containment leakage rate tests shall 2.
A combined leakage rate of less than or equal be conducted at 40110 month intervals
- to 0.61a for all penetrations and valves, during shutdown at >Pa during each except'for main steam isolation valves, 10-year service perTod. The third subject to Type B and C tests when pressurized test of each set shall be conducted to Pa.
during the shutdown for the 10-year Less than or equal to 11.5 scf per hour for plant inservice inspection.
3.
any one main steam isolation valve when tested 2.
If any periodic Type A test fails to at 25 psi.
meet 0.75La, the test schedule for subsequent Type A tests shall be With the measured overall integrated primary reviewed and approved by the Commission.
contalument leakage rate exceeding 0.75La, or the If two consecutive Type A tests fail to measured combined leakage rate for all penetrations meet 0.75La, a Type A test shall be and valves, except main steam isolation valves, performed at least every 18 months until subject to Type B and C testing exceeding 0.6La, two consecutive Type A tests meet 0.75La, or the measured leak rate exceeding 11.5 scf per hour at which time the above test schedule for any one main steam isolation valve, restore leakage rates to less than or equal to these values may be resumed.
prior to increasing reactor coolant system temper-3.
All Type A test leakage rates shall be ature above 212* F or, alternatively, restore calculated using observed data converted measured leakage rates to within'these limits withi to absolute values. Error analyses n
one hour or be in at least 110t Shutdown within the shall be performed to select a balanced 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in Cold Shutdown within the integrated leakage measurement system.
next following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
i
- The second test of the second 10-year service period may be conducted during the 1989 refueling outage.
159
~1. 7/4. 7
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- ' t A -
3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILIANCE REQUIREMENTS 2.
Velds in austenttic stainless steel pipin6:'
four inches or-larger in diameter' containing reactor coolant at a temperature above 200 degrees F during-power operation, including reactor vesself attachments and appurtenances, not meeting the requirements of NUREG-0313. Revision 2, for IGSCC Category A weldments shall be included in-an augmented inspection-program meeting the requirements of NUREC-0313. Revision 2.
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Exhibit'D,
},
License = Amendment' Request Dated March 31,-1989
[
. Docket No. 50 263 -License No. DPR.22
- Exhibit D consis'es of General Electric Report SASR 88 99, Revision 1, January.1989,- " Implementation.of Regulatory Guide 1,99, Revision 2,
- for the Monticello Nuclear Generating Plant.". This_ report was pre.'
pared-to. support changes proposed in this License Amendment Request.
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SASR 88 99 DRF 137 0010 January 1989' Revision 1 u
IMPLEMENTATION OF-REGULATORY GUIDE 1.99, REVISION 2 FOR THE MONTICELLO NUCLEAR' GENERATING PLANT 1
i
- i
'I i A W
\\
l Prepared by:
T. A. Caine, Senior Engineer Materials Monitoring &
Structural Analysis _ Services j
.i
/1BAuA/n Verified by:
C. J. Papandrea, Engineer Materials Monitoring &
-j Structural Analysis Services
'1
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.i Reviewed by:
S. Ranganath, Manager Meterials Monitoring &
S:ructural Analysis Services M
W' GENuclear Energy qg@if.@7l1ge g
P
i IMPORTANT NOTICE RECARDING
. CONTENTS OF THIS REPORT
' PLEASE READ CAREFULLY k
This-report was prepared by General Electric - solely for the use The information contained in this of Northern States Power Company.
report is believed by General Electric to be an accurate and true representation' of the facts known,' obtained or provided to General Electric at the time this report was prepared.
The only undertakings of the General Electric Company respecting are contained in the contract governing information in this-document Northern States Power. Company Purchase Order E62730MQ Task 88 03;'and shall be construed as changing-said nothing contained in this document contract.
The use of this information except as defined by said or for any purpose other than that for which 'it is intended,
- contract, any such unauthorited use, is not authorized; and with respect to neither General Electric Company nor - any <of the contributors to this document makes any representation or warranty (express or implied)' as to~~
the completeness, accuracy or' usefulness of the information contained in this document or that such use of such information may i
not infringe. privately owned rights:- nor do they assume any responsibility for liability or damage of any kind which may result from such use of such information.
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a TABLE OF CONTENTS
! ZA< 7. ".' 5;.37_I,i, 2.,
1-1
1.0 INTRODUCTION
21 2.0:
SUMMARY
31 OF BELTLINE MATERIALS 3.0 INITIAL RTNDT 3*1 3.1 Plate Initial RTHDT-3*1
= 3.2 Wald Initial RTNDT 3-2 Uncertainty 3.3 Initial RTNDT 41 4.0 IRRADIATION SHIFT 4-1.
C.1 Rev 2 Methods 4-2 4.2 Limiting Beltline Materials 4-2 Irradiation Shift Versus Fluence 4.3 5-1 5.0 PRESSURE-TEMPERATURE CURVES 51 Application of the P T Curves 5-2 5.1 Bottom Head Monitoring During Pressure Tests
'5.2 6-1
6.0 REFERENCES
7 A1 SUGGESTED REtISION TO THE UPDATED FINAL APPENDIX A SAFETY ANALYSIS REPORT B1 SUGGESTED REVISION TO THE TECHNICAL APPENDIX B SPECIFICATION i
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1.0 INTRODUCTION
In 1984, 'GE provided Northern States Power (NSP) with pressure-temperature (P T) curves updated to the 1983 requirements of 10CFR50, Appendix G.
The rep' ort [6 1). showed the weld material to be the limiting material _ in the beltline region, because of the high initial reference temperature of nil ductility transition (RTNDT) assumed.
The high initial RTNDT, 40'F, was reported in_ a 1979 CE report on the Monticello surveillance program ' [6-2).
The-value of 40'F ~ was established from the Monticello vessel purchase specification requirements because there were no Charpy data available for the veld material heats used in the Monticello beltlire welds.
In 1987, NSP obtained data from Alloy Rods to use as a basis for reducing the vulue of the assumed initial RTNDT of the beltline welds.
based on the data GE revised the estimate of weld initial RTNDT, provided by NSP, and revised the pressure-temperature curves for Monticello, still usin5 ReSulatory Guide 1.99, Revision 1 to predict NDT shife due to irradiation- (6-3].
RT In May 1988, Revision 2 to ' Regulatory Guide 1.99 (Rev 2) was issued by the NRC.
Rev 2 presents different methods for predicting RTNDT shift, typically increasin5 shifts for M'Rs.
NSP has requested that the P-T curves and RTNDT shift relationships reported in [6 3) be revised to reflect the methods of Rev 2.
In further discussions between GE and NSP, it was agreed that reductions in heating prior to pressure testing might be possible by monitoring the bottom head te=peratures against a separate PT curve from that used for the beltline and the rest of-the vessel, and that GE would address the bottom head monitoring option in this analysis.
Therefore, the objectives of this report are as follows:
1.
Determine the limiting beltline material and calculate 'the irradiation shift as a function of fluence using Rev 2 as a basis.
1-1 f
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-'2.
Revise the pressure temperature' curves, if necessary, to reflect value of the limiting beltline material.
the initial RTNDT
, Include the necessary discussion and technical bases to explain
- 3.
and justify monitoring bottom head temperatures separately from beltline temperatures during pressure testing.
4 Modify : the updated final safety ~ analysis report (tJFSAR) and.
Technical Specification inputs Yrom (6-3) to reflect the changes due to this evaluation.
4 B
1 p
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+
2.0
SUMMARY
The pressure-temperature curves governing vessel operating limits for.Monticello are revised to reflect implementation of Regulatory -
Cuide.1.99,-Revision 2.
Initial RTNDT values are documented, as well as their uncertainty, if appropriate.
The approach in Revision 2 to and Margin are determining quarter thickness
- fluence, ARTNDT presented.. Predictions for adjusted reference temperature (ART) for 32 effective full power years of 'o'peration are made, showing that conditions for ART specified in 10CFR50 Appendix C are met.
The' place material'(place 1-15) is shown to still be the limiting
. beltline material.
The relationship of irradiation shift versus fluence' was revised.-
The new curve is calculated according to Regulatory Guide 1.99, Revision 2, with the chemical composition of l
the limitin5 P ate.
The pressure temperature curves consist of the curves for the region. remote from the beltline, which are not affected by the change in ' the - re5ulatory guide, and the belrline curves.
The beltline, including the irradiation shif t per Revision 2, is more limiting than the non-beltline'for conditions of interest such as the pressure test.
'The pressure test curve for the region remote from the beltline may be ur.ed for allowable bottom head temperatures, rather than using the more restrictive limits of the beltline curve. This may allow NSF to lower pressure teste temperatures 20*F to 30*F, offsetting some of the impact of the increase due to Rev 2.
Suggested revisions to the UFSAR and Technical Specification from are modified to reflect the chan5es of this analysis.
The modified versions are in Appendices A and B.
Changes from the text previously provided to NSP are indicated with a bar in the margin.
2-1 C
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1 4F BELTLINE MATERIALS 3.0 INITIAL RTNDT The.' methods, used to-determine initial RTNDT values of the beltline - plate sacerialsL and the resulting RTNDTs -were presented in-The most recent determination
[6-1),; and are reviewed briefly below.
in (6-3) is reviewed here as well.
of ' beltline. weld initial RTNDT
- Rev 2 -includes a new ' requirement to - determine the uncertainty of-NDT values, ag, which is ' addressed in this section.
- initial RT 3.1 PLATE INITIAL RTNDT Charpy testing of plates during fabrication ~ demonstrated 30 ft lb impact. energy in longitudinal specimens.
A GE procedure shown in
[6-1) was used on these data to calculate equivalent RTNDT values to those that would be determined with post-1972 (current) ASME Code test methods. ' The GE procedure basically requires a 2'/ft-lb adjustment on test temperature to raise Charpy energy data to the 50 f t-lb - level, to account -for the-and then ? requires - a 30*F addition to RTNDT lonS tudinal and transverse Charpy specimen i
difference between orientation.
The CE procedure was derived as a: conservative bound of data for
}
r 24 plates.of SA 533 low alloy steel tested as part of the work l
Bulletin 217 [6-4) and 22 reported in Welding Research Council (WRC) plant-specific place data sets retrieved from GE Quality Assurance o
' records.
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The vessel beltline seam welds were made using a shielded metal As shown in [6 2), the welding rods used arc welding (SMAW) process.
for the seam welds were provided by Alloy Rods Company.
Since the l'
records provided with the welding rods did not. include Charpy test values for the data, it is not possible to deter =ine specific RTNDT
+
Monticello beltline weld heats.
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- 3 During, fabrication, there was some material certification tests, which included dropweight and j
referred to-as " Category B" tests,.
Charpy testing of E8018NN weld material used in the Monticello vessel.'
V81"'
The dropweight and Charpy; data, shown in ~ (6-3 }. support a RTNDT et
.',0'F.
1 The first capsule of the Monticello surveillance program was removed an6 tested in 1984 ( 6_- 5 ).
The test results. included Charpy data for' Irradiated weld metal which indicated.a RTNDT of e75'F in the irradiated condition. Therefore, the weld metal heat, or heats, used~
to make the surveillance weld had an initial RTHDT of -75'F or lower.
Vnile this information provides justification for a lower veld-RTNDT, it cannot be used as a valid basis for estimating the initial veld RT because many different veld metal her.ts could have been
Therefore, NSP contacted Alloy Rods Corpora: ion to - obtain a data base of E8018NM RT values for NDT values, shown in [6-3), ranging statisti. cal use.
There are 45 RTNDT from -30'F to -90'F.
Assuming a normal distribution, Mean RTNDT = -65.6'F, l-og - 1?.7'F, l
whera ay is the standard deviation.
M ERTAINU 3.3 INITIAL RTNDT l
a Mar 5 n term to be added to the i
The methods of Rev 2 include The Margin term includes a component for calculatad value t.RTNDT.
uncertainty in initial ETHDT'
'I.
Rev 2 discusses determination of a7 for two categories of initial RTNDT, ceasured values and generic mean values.
For generic mean values, ey is si= ply the standard deviadion eticulated for - the data set used to co=pute the mean.
For eensured values, requirements for deter =inatien of e7 are sc:ewhat vague.
3*2
m Rev 2 states', "If a measured value of initial RTNDT'for the material
~
in question is availabis, ag is to be estimated from the precision of values derived from tho' test method.""
CE's position for. RTNDT seasured data is that ag is zero, as explained below.
Y' The Charpy curves fit to surveillance data, which ultimately data for development of Rev 2, were best estimate provided the ARTNDT fits.
An idealized example is provided as curve #1 in Figure 3-1.
is based on the t
However, the ASME Code-approach to determining RTNDT lowest value of three specimens exceeding the required limits of impact energy and lateral expansion.
A visualization of a Charpy curve drawn on the basis of the Code RTNDT approach is shown as curve
- 2 in Figure 3-1.
In comparing curves #1 and #2, it is clear that
-curve #2, which is based _on the lowest value rather than the mean Therefore, value, provides a conservative estimate of initial RTNDT.
from measured data is the ASME Code method of determining-RTNDT conservative, and og - O'F is appropriate.
discussed in Section 3.1, a procedure was used to modify the As measured data for plates to make the data equivalent to current methods.
As already mentioned, that procedure yields conservative RTNDT estimates.
Furthermore, the procedure operates on the lowest Charpy energy, not the mean or average energy for a given test values based Therefore, for the case of equivalent RTNDT temperature.
on measured Charpy dcta, 07 - 0*F is appropriate.
k In the Rev 2 draft which was circulated after editing to incorporate 8
public co=ments, the text stated, "ag, the standard deviation for the g
if a measured value cf initial may be tchen cs ter initial RTNDT, for the material in question is available."
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130 -
120 -
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40 80 120 160 200 Charpy Test Temperature (F) 4
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Appr ach Figure 3-1.
Comparison of Surveillance Data Fit and RTNDT
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4.0 1RRADIATION SHIFT-
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4.1 REV 2 METHODS Rev i-provides methods to predict RTNDT shift due to irradiation, as did Regulatory Guide 1.99, Revision 1.
In addition, Rev 2 has methods for determining uncertainty margin to be included as part of.
the shift, and has methods for determining fluence attenuation.
The shift correlation in Rev 2 consistr of two terms:
SHIFT - ARTNDT + Margin (4-1).
NDT - [CF)*f(0.28 0.10 log f)
(4-2)
+ a3 )0. 5 (4,3).
2 2
Margin - 2(ay where f - fluence (n/cm ) / 1019, and 2
CF - chemistry factor, provided in Rev 2 tables.
In previous shif t calculations with Rev 1, the 1/4 T fluence was used as one of the inputs to regulatory guide calculation.- However, Rev 2 now requires that the vessel inside surface fluence, fsurf, be used as. one of the inputs to the regulatory guide calculation.
The surface fluence is then attenuated to the vessel 1/4 thickness depth (1/4.T) as part of the regulatory guide calculation by the expression surf (* 0.24x)
(4,4) f
-f x
where x - distance (inches) into the vessel from the inside surface.
- The minimum thickness assumed for the vessel in the beltline region is 5.06 inches,. neglecting cladding.
A ratio of surface f;uence to 1/4 ~ fluence of 1.30 is reported in (6-51 The 1/4 T fluence previously assumed for 32 EFPY [6-3) was 1E 2
3.9x10 n/cm, so based on the ratio from (6-5), the surfsee fluence,
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has been calculated, Margin is determined based on ag Once ARTNDT
-and aA.
Rev 2 has set values of a3 of 17'T for plate and 28'F for
~
. However, a3 need not be greater than 0,5*ARTNDT' veld.
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The adjusted reference' temperature (ART) is - the - initial RTNDT
' ~ ~
-plus'the Rev 2' shift in Equation 4 1:
ART = Initial RTNDT + ARTNDT + Margin (4 5) 4.2 LIMITING BELTLINE MATERIAL L
The limiting beltline material, from the perspective of brittle fracture, is the material with the highest ART at a given time.
The information on plate and weld chemical compositions from (6 2),
summarized in Table 4 1.
initial RTNDT values and 32 EFPY shifts are As shown, the plate material (plate 1-15) is the limiting beltline material.
- 4. 3 IPJtADIATION SHIFT VERSUS FLUENCE The Rev 2 shift correlation, Equation 4-1, for the chemical composition ~ of plate 1-15 is plotted versus 1/4 T.
fluence in 7
Figure 4-1.
The fluence shown is the value attenuated from the inside surface fluence, per Equation 4-4 For example, the attenuated 1/4 T fluence for 32 EFFY is:
18
-0.24(5.06/4) f1/4 T - 5.1x10 e
18 2 at 32 EFFY.
f1/4 T - 3.8x10 n/cm 4=2
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DETERMINATION OF LIMITING BELTLINE MATERIAL?
a.
Material Chemistry
' Initial
- NDT
- Martin bR_I.
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identification ICu INI Factor NDT Plates:
1-14 0.17 0.58 125.3 0*F.
0*F 92.1*F 34*F 126.1* F ;l 1
1-15 0.17 0.58 125.3 14*F 0*F 92.1*F 34*F 140.1*F E
t-16 0.14 0.56 98.2 0*F O'F 72.2*F 34*F
' 106.2*F e,
1-17 0.17 0.50 118.5 6*F 0*F 87.1*F.
.34*F
. i 127.1*F ai Uc1ds:
l 6
Limiting 0.10 O.99 134.9
-65.6*F 12.7'F" 99.1*F 61.5"F.I 95.0*F; b'
(h Case NDT values computed for a 32 EFPY fluence at the vessel ART I
Inside surface of 5.1x10 n/cm2 (3.8x10 n/cu2 at 1/4 T).
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Cu limit accepted by NRC in correspondence following surveillance b
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weld chemical analysis.
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v 5.0 PRESSURE TEMPERATURE CURVES The pressure temperature curves used in the Monticello Technical Specification consist of two separate curves:
a curve remote from the beltline and a curve for the beltline.
The curves remote from the
'f beltline are based on the feodwater nozzle limits with a RTNDT 40*F.
Thr.se curves apply to all vessel regions-except the. beltline region, and are not affected by the changes to RTNDT shift associated with Rev 2.
The beltline curves apply - to the plates and. welds discussed in Section 4, and must be calculated taking into account the Rev 2 RTNDT shifts.
The P-T curve plots for Monticello are shown in
. Figures 5-1 through 5-3.
L 5.1 APPLICATION OF THE P-T CURVES Figure 5-1 is the pressure-temperature curve for pressure tests, referred to as curve A.
Figure 5-2 is, the curve for non-nuclear heatup and cooldown, called Curve B.
Core critical operation is go'verned by the pressure-temperature limits in Figure 5-3, called Curve C.
The beltline region curves are calculated according to Appendix G of ASME Code Section III, assuming a 1/4 T flaw.
The beltline values correspond to the initial, or zero EFPY, plate properties (RTNDT - 14'F).
The actual beltline limits for a given period of operation are determined by adding the appropriate shift
'from Figure 4-1, based on Rev 2, to the beltline P-T curves.
L The vessel operating limits are determined as the combination of l
' beltline and non-beltline limits that result in the highest l-temperature for a givt.n pressure.
For a leak test at 1000 psis, the non beltline temperature of 162*F is no longer governing compared to the shifted beltline curve.
Assuming that Monticello has accu =ulated about 12 EFPY of operation, the leak test temperature, based on the beltline curve, is 95'T + 96'F Shift - 191*F.
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t 5.2 BOTTOM HEAD MONITORING DURING PRES $URE TESTS-V
. While. the -beltline curves are limiting 'for pressure tese conditions,' the non beltline limits can still bn applied to the bottom
-head region.
It is likely that, during leak and hydrostatic pressure-testing, the bottom head temperature may be significantly cooler than the higher elevations of the vessel.
This condition can occur when the recirculation pumps are operating at low speed, or are off, and injection through the control rod. drives is used to pressurize the vessel.
Monitoring the bottom head separately from. the beltline region may reduce the required pressure test temperature by as much as 20*F to 30*F.
Some hypothetical temperatura demonstratin5 the potential benefit of separate bottom head monitoring are shown in Figure 5 4..
.The Technical Specifications currently require that all vessel
.y temperatures be above the limiting conditions on the P-T curve.
That would mean the bottom head would have to be heated above 191*F at 12 EFPY, as shown in case (a) of Figure 54 The bottom head q
temperature reading would likely be the limiting reading on the vessel during the test.
If the required temperature for the bottom head were only 162*F, the limiting reading would probably be near.the beltline, as shown in case (b), and the actual. vessel temperatures could be lowered compared to case (a).
One condition on monitoring the bottom head separately is that it must. be demonstrated that the vessel beltline temperature can be monitored during pressure testing.
An experiment has been conducted 1
at a Bk'R/4 - which showed that thermocouples on the vessel near the feedwater no::les, or temperature measurements of recirculation pump inlet water provide good estimates of the beltline temperature during pressure testing.
NSP may need to confirm this before implementin5 separate monitoring of the bottom head.
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T t
a) Bottom llend Mcnitored by b)- Bottom llead Monitored Beltline Linits Separately from Beltline Figure 5-4.
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i 6.0 RETERENCES
_.. :., a-
"~"~
~
!=
61 "Monticello Nuclear Generating Plant' Final 'Sa'fety Analysis and Technical Specificaticus Revised to 10CTR$0 Appendix C, 19 8 3, .
L.
General Electric Company, September 1984 62 "Monticello Nuclear Generating Plant Information on Rea(. tor Vessel Material Surveillance Program," CE Report NEDO 24197, Revision 1, October 1979.
63
" Revision of Pressure. Temperature Curves to Reflect Improved Beltline Ueld Toughness Estimate for the Monticello Nuclear Generating Plant," CE Report SASR 87 61, Revision 1, December 1987.
64 Hodge, J.M.,
" Properties of Heavy Section Nuclear Reactor Steels " Velding Research Council Bulletin 217, July 1976.
l 6-$
" Examination. Testing and Evaluation of Irradiated Pressure Vessel Surveillance Specimens from the Monticello Nuclear Generating Plant,"
Bactelle Columbus Laboratories Report BCL.585 64 2, Revision 1, November 1984 f
4 9
r l'
l
APPENDIX A i
SUCCESTED REVISION TO THE UPDATED FINAL SAFETY ANALYSIS RTFORT.- --- 4 4--
i I
Subsection 4.2.3.2 of the Monticel'lo Updated Final. Safety Analysis Report (UFSAR) should be replaced with the text in.this Appendix.
Revisions of this text from the version in [6 3) are Pa5 nation of the UTSAR 1
indicated with revision bars in the margin.
should be changed to accommodate this revision.
I i
?
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. 5 1
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l 9
0 l
l.
A1
r MONTICE1.1.0 4.2.3.2 Fracture Teutbress of Reector Pressure Vessel 4.2.3.2.1 Coweliance with 10CFR50 Accendix 0. May 1983 A maj or condition necessary for full compliance to Appendix G is satisfaction of the requirements for the Susemer 1972 Addenda to Section 111 of the ASME Code.
This is not possible with components which were purchased to earlier Code requirements.
(The Monticello reactor pressure vessel (RPV) was manufactured to the 1965 Edition of the ASME Code, to and including the Summer 1966 Addenda.)
Ferritic materials complying with 10CFR50 Appendix G must have both drop weight tests and Charpy V Notch (CVN) tests with the CVN specimens oriented transverse to the principal material working The CVN tests l
direction to establish the reference temperature RTNDT.
nust be evaluated against both an absorbed energy and a lateral expansion criteria.
The maximum acceptable RTNDT must be determined in accordance with the analytical procedures of the ASME Code Section III, Appendix G.
Appendix G of 10CFR50 requires a minimum of 75 f t lbs upper shelf CVN energy for unirradiated beltline materials, and at.least 50 f t lbs upper shelf CVN energy at the end of life.
It also requires at least 45 ft lbs CVN energy and 25 mils lateral enpansion for boltinB material at the lower of the preload or lowest service temperature.
By comparison, materials for the Monticello RPV vere qualified by drop weight tests and/or longitudinally oriented CVN tests (both not required), generally at only one temperature, confirming that the material nil ductility transition temperature (NDT) is at least 60'F below the lowest service temperature.
There was no upper shelf CVN energy requirement on the Monticello beltline materials.
The bolting material was qualified to a 30 f t lb CVN energy requirement at 60'T below the minimum preload temperature.
A;
^
MONTICELLO From the above comparison it can be seen that the fractura tsughness l
testing performed on the Monticello RPV material cannot be shown to comply directly with~ 10CTR$0 " App'eidd c.7 SoMer, ~ to' EleterEne7.'.71~..
l
'~
operating limits in accordance with 10Cnt$0 Appendix c, estimates of 5
values of values and the hi hest RTET the beltline materials RTET all other materials were made, as explained in Para 5raph 4.2.3.2.1.1.
The method for developing these operating limits is described in-f Paragraph 4.2.3.2.1.2.
On the basis of the last paragraph on page 19013 of the July 17, 1973 Tederal Retister, the following is considered an appropriate method of compliance.
4.2.3.2.1.1 Method of cemellance The intent of the proposed special method of compliance with Appendix 0 of the ASME Code is to provide operating limitations on 1
pressura and temperature based on fracture toughness. These operating limits assure that a margia of 6afety against a non ductile failure of this vessel is essentially the same as a vessel built rm the Summer 1972 Addenda.
The specific temperature limits for operation are based on 10CTR50, i
g Appendix G, May 1983.
4.2.3.2.1.2 Methedr of Obtainine liaitt Bared en Fracture Teuchners Operating limits which define minimum metal temperatures versus reactor -pressure during nermal heatup and cooldown, and during inservice hydrostatic testin5, were established using the methods of Appendix C of Section III of the ASME Boiler and Pressure Vessel Code, l
up to and including the Summer 1976 Addenda, f
The vessel and discontinuities such as the RPV flanges, nozzles and i
l bottom head penetrations were evaluated and the operating limit curves I
are based on the limiting location.
The boltup limits for the flange l'
and adjacent shell regions are based on a minimum metal temperature of l
RTg 7 + 60'T.
The maximum through wall temperature Ersdient from A3
. ~, -,
+-
m 1
HONTICELI0 l
The f
continuous heating and cooling at 100'T per hour was considered.
safety factors were as specified in the ASME Code Appendix C.
I' For the purpose of setting these operating limits the reference l
temperature, RTNDT, was determined from the toughness test data taken in accordance with requirernents of the Code and the Ceneral Electric RPV purchase specification to which the Monticello RPV was dasigned and manufactured.
These toughne ss test data, Charpy V Notch (CVN) and/or dropweight nil ductility transition temperature (NDT) were analyzed to establish compliance with the intent of 10CTR50 Appendix C.
Because all toughness testing needed for strict compli-ance 4th
.*.yy ncix C was not required at the time of RPV procurement, some toughness results are not available.
For example, longitudinal CVNs, instead of transverse, were tested, usually at a single test temperature of +10*F or +40'F, and only against an absorbed energy criteria.
Also, at the time, either CVN or drop weight testing was permitted; therefore, in some cases both tests were not performed as i r, currently required.
To substitute for this absence of certain were derived for the vessel data, toughnes s property correlations materials in order to operate upon the available data to give a of 10CFR50 conservative estimate of RTNDT, compliant with the intent Appendix C criteria.
These toughness correlations vary, depending upon the specific material
- analyzed, and were derived from the results of VRC Bulletin 217, " Properties of Heavy Section Nuclear Reactor Steels",
In the case of vessel and from toughness data for other BWR reactors.
i plate material (SA 533 Grade B, Class 1),
the predicted limiting l
toughness property is either NDT or transverse CVN 50 ft lb temperature minus 60'F, whichever is greater.
L i
As a matter of practice where NDT results are missing, NDT is estimated as the longitudinal CVN 35 ft lb transition temperature.
h However, for the Monticello vessel plates, "no break" dropweiE t information was availatla at Z specified temperaturas, se the nil conse: tatively taken as 10 ductility transformation temperature was degrees belew tne "no brean" test temperature.
The transverse CYN A-
MONTICE1.L0
.- (
estimated from longitudinal CVN 50 ft.lb transition temperature was data in the following manner.
The lowest longitudinal CVH ft.lb values was adjusted to derive a longitudinal CVN 50 ft lb transition If the temperature by adding 2'T per ft lb to the test temperature.
actual data' equalled or exceeded 50 f t lb, the test temperature was used.
Once the longitudinal 50 ft.lb temperature was derived, an additional 30'T was added to account for the orientation change frem 50 ft lb.
For f,orgings (SA 508 longitudinal 50 f t.ib to transverse Class 2), the predicted limiting property is the same as for vessel was estimated in the same way.
plates, and the RTNDT For the vessel weld metal the predicted limiting property is the CVN 50 ft lb transition temperature minus 60'F, as CE experience indicates the dropweight NDT values are typically 50'F, or lower for these materials.
The C"N 50 f t lb temperature would be derived in the same way as for the vessel plate material, except the 30*F addition for orientation effects was omitted since there is no principal working direction in veld metal.
If NDT valuss were available, they would v uld be taken as the higher of NDT also be considered and the RTNDT or the 50 ft lb transition temperature minus 60*F.
However, no data, either Charpy or dropweight, were available on the fracture toughness The veld of the specific weld materials used in the Monticello RPV.
data for E8018NM rod supplier, Alloy Rods Corporation, provided RTNDT weld metal, which was used to determine a statistically conservative CI The data showed a mean RTNDT estimate of the beltline veld RTNDT.
65.6'F, with a standard deviation of 12.7'F.
For vessel weld heat affected, rene (HAZ) material the RT was NDT assumed the same as for the base material as ASME Codo veld procecure treatment indicate qualification test requirements and post veld heat this assumption is valid.
l Original closure bolting material (SA-540 Grade B24) toughness test L
requirements were for 30 f t lb at 60'T below the boltup temperature.
are for 45 f t lb and 25 mil I
current 10C??.50 Appendir. O ret'.:irements 1cteral expansien (MLE) at the preload or lowest service temperature, including boltup.
All closure stud materials meet L5 ft lb absorbed A5 l
1 l
L6
~r.g:,ia h;Lu
=
cnergy ct 010'F but cils 10ttr:1 cxp:nsicn results vero not rcpertod.
Since total compliance with current requirements could not be shown, the original requirements were used to establish the boltup temperature.
The purchase agreement for Monticello closure stud material was for 30 f t lb at +10'F, and no deviations were reported.
Calculated Values of Initial RTNDT 4.2.3.2.1.3 The methods of Subsection 4.2.3.2.1.2 were used to calculate initial
'l values for the core beltline ' plates and welds, closure flange
. RTNDT
- region, nozzles and other discontinuities, and closure bolting 1"
material.
The calculation methods conservatively estimate RTNDT' order to meet the intent of 10CFR50 Appendix G criteria.
The core beltline plate and weld RT values are presented in NDT Table 4.2.3.2 1.
Regulatory Guide 1.99, Revision 2 requires that the standard deviation og be estimated for each beltline material.
For the beltline welds, cy is estimated from the data set evaluation performed.
For the beltline plates, actual measured values from each plate in question were obtained, and a conservative method of from that data was used.
Therefore, ay for the determining RTNDT beltline plates is estimated to be O'F.
Adjusted CVN data for the closure flanges and adjacent plates gave an RT of
+10'F.
Based on the NDT requirement in the purchase NDT agreement, the RTgg7 of the velds at the closure flanges was +10'F also.
Calculations for the nozzles and discontinuities gave an RT )7 g
of +40*F based on purchase agreement NDT requirements.
The closure bolting material RT )7 used was +10'F, based on the purchase agreement g
(-
NDT requirement.
4.2.3.2.1.4 Effect of Ner:1es and Discontinuities en Operatine limits The minimum temperature for boltup and pressurization was established by addin5 50'T to the RT fr the closure flange region, the NDT critical location during boltup.
7 l
A6
'l r
n, -
MONTICELLO I
The 60'T temperature added for boltup, a requirement of the ASME Code f
applicable to the original Monticello design work, provides 60'T
{
f margin not required by 10CFR$0 Appendix C for boltup and pressurization up to 20% of the system hydrostatic test pressure.
Above 20% test pressure, 10CFR$0 Appendix C Paragraph IV.A.2 requires the closure flange region to be 90'T above RT for hydrostatic NDT for normal operation, pressure tests and leak tests,120'F above RTNDT when the core is critical. The 90'T requirement and 160'T above RTNDT is met by adding 30'T at 20% test pressure to the 60*F boltup margin.
temperatures required at 10% test pressure for non nuclear heatup The 120*F and cooldown and core critical operation exceed the respective 4
and 160*F nargins required by 10CFR$0 Appendix C.
The effect of the nozzle and bottom head discontinuities was considered by adjusting the results of Bk'R/6 reactor discontinuity analyses to the Monticello reactor.
The adjustment was made by increasing the minimum temperatures required by the difference between values.
The nortle and bottom hea'd the Monticello and BiiR/6 RTNDT of +40'F.
adjustments were based on an RTNDT 4.2.3.2.2 Pressure Tercerature Limits 4.2.3.2.2.1 Limit Curves The baris for setting operational limits on pressure and temperature for normal, upset and test conditions for the RPV is described in Section 4.2.3.2.1.2.
The fracture toughness analysis was done for the normal heatup or cooldown rate of 100*F/ hour and it also included the effects of cold
+
The vater injections into the no::les and other operation transients.
temperature gradients and thermal stress effects are included.
The results of the analyses are a
set of operating limits ~ on Figure 4.2.3.2 1.
including pressure testing (curves labeled A).
non nuclear heatup or cooldown (curves labeled B) and core critical operation (curves labeled C).
A7
-e--
a
MONTICELLO 4
4.2.3.2.2.2 Temeerature timits for Boirue L
A minimum temperature of 70'T is required for the closure studs.
A sufficient number of studs may be tensioned at 70'T to seal the closure flange 0 rings for the purpose of raising reactor water level about the closure flanges in order to assist in warming them.
The i
flanges and adj acent shell are required to be warned to minimum temperatures of 70'T before they are stressed by the full intended F
bolt preload.
The fully preloaded boltup limits are shown in Figure 4.2.3.2 1.
4 4.2.3.2.3 Irradiation Effects l
i Estimated maximum changes in RT for 32 effective full power years NDT I
(EFPY) of fluence at the one quarter thickness (1/4 T) depth of the vessel beltline materials are shown in Table 4.2.3.21.
The updated
. predicted peak 32 ETPY fluence at the 1/4 T depth of the RPV beltline, based on the methods in Regulatory Guide 1.99, Revision 2,
is 18 2
3.8x10 n/cm.
Irradiation shifts, including Margin, were calculated in accordance with the rules of Regulatory Guide 1.99, Revision 2.
Results show that beltline plate 1-15 is limiting through 32 EFPY.
Since predicted adjusted reference temperatures (ART) are less than 200'F, provisions to permit thermal annealing of the RPV in accordance with Paragraph IV.B of l'0CTR50 Appendix 0 is not required.
The predicted 32 ETPY shift in RT fr plate 1 15, shown in NDT Figure 4.2.3.2 2 (based on the neutron fluence at 1/4 of the vessel wall thickness) was added to the core beltline limits to arrive at the curves A',
B' and C' in Figure 4.2.3.3 1.
The predicted shift in the core beltline RTNDT shown in Figure 4.2.3.2 2 is based on the results of flux vire measurements of fluence versus full power years of operation toEether with the relationships between fluence and Cu and Ni given in Regulatory Guide 1.59, Revision 2.
MONTICELLO i,4 4
4.2.3.2.4 oeeratine Procedures e
......... _.......... ;.c.;.y
. in. Para.
7*
By comparison of the pressure versus temperature limits f
graph 4.2.3.2.2 above, with intended normal operating procedures for the most severe upset transient, it is shown that the limits will'not be exceeded during any foreseeable upset condition. Reactor operating
[
procedures have been established such that actual transients will not be more severe than those for which the vessel design adequacy has been demonstrated.
Of the design. transients, the upset condition producing the most adverse temperature and pressure condition anywhere in the vessel head and/or shell areas yields a minimum fluid Scram temperature of 250'F and a maximum pressure peak of 1180 psige automatically occurs with initiation of this event, prior to the reduction in the fluid temperature, so the applicable operating limits given by Figure 4.2.3.21 Curves A.
For a temperature of 250*F, are the maximum allowable pressure exceeds 1300 psig for the intended margin against non ductile failure.
The maximum transient pressure of 1180 psig is therefore within tl.e specified allowable limits.
b 9
i A9
.- ~
g Tabla 4.2.3.2-1 DETERMINATION OF LIMITING SELTLINE MATERIAL' tfaterial Chemistry Initial.
RT ANNDT
- Marrin Mtl '
Ident:I fIcat f or, ICu It[1-Factor NDT Plates:
1-14 C.17 0.58 125.3 0*F 0*F 92.1*F 34*F 126.1*F 1-15 0.17 0.58 125.3 14*F 0*F 92.1*F 34*F 140.1*F 1-16 0.14 0.56 98.2 0*F 0*F 72.2*F 34*F 106.2*F e
1-17 0.17 0.50 118.5 6*F 0*F 87.1*F 34*F 127.1*F velda:
C Limiting 0.10 0.99 136.9
-65.6*F 12.7'F 99.1*F 61.5*F.: l 95.0*F b
Case
- ART values computed for a 32 EFFY fluence at the vessel
'i ilDT I8 I8 inside surface of 5.1x10 n/cm2 (3.8x10 n/cm2 at 1/4 T).
I' Cu limit accepted by 1.'RC in correspondence following survelliance weld chemical analysis.
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APPENDIX B l
i
- succESTED REVISION.TO THE TECHNICAL SPECITICATION'
....~..
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The following listed technical specification changeh are l
recommended for Monticello.
Tech Seee Pare (s)
Rectacament Text 1
i 122 B2 133 through 136 B.3 through B 6 Revisions to the text are indicated by margin bars.
Figura 3.6.1 has l
been revised to reflect the shift relationship presented in this report.
Figures 3.6.2 through 3.6.4 are unchanged from the curves presented in (6 3), but are included in this appendix for information.
I The current technical specification pagination ar.d figure designation should be revised, if necessary.
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MONTICELID -
3.6 LIMITING CONDITIONS FT)R OPERATION 4.6 ' SURVEILIANCE REQUIREMENTS B. Reactor Vassol Tempercture and Proceure B. R2acter Vescal Temperctura and Pressure
?*
1.
During in-service hydrostatic or leak 1.
During in-service hydrostatic.or leak testing, the reactor vessel shell testing when the vessel pressure is above temperatures specified in 4.6.B.1, 312 psig,~the following temperatures.shall except for the reactor vessel bottom-be recorded at least every 15 minutes.
head, shall be at or above the higher of the temperatures shown on the two curves a.
Reactor vessel shell adjacent to shell of Figure 3.6.2 where the dashed curve, flange.
- RPV Beltline Region", is increased by the core beltline temperature b.
Reactor vessel bottom head.
adjustment from Figure 3.6.1.
The reactor vessel bottom head temperature c.
Reactor vessel shell or coolant shall be at or above the temperatures temperature representative of the shown on the solid curve of Figure minimum temperature of the beltline 3.6.2, "RPV Reseote from Core Beltline",
region.
with no adjustment from Figure 3.6.1.
{
2.
During heatup by non-nuclear means 2.
Test specimens representing the reactor (except with the reactor vessel vented),
vessel, base weld, and weld heat affected cooldown following nuclear shutdown, or rone metal shall be installed in the reactor low level physics tests the rea : tor vessel adjacent to the vessel vall at the vessel shell and fluid temperatures core midplane level. The material sample specified in 4.6.A shall be at or above program shall conform to ASTM E185-66.
the higher of the temperatures of Figure Samples shall be withdrawn at one fourth and 3.6.3 where the dashed curve, three fourths service life. Analysis of the "RPV Beltline Region", is increased by first sas.ple shall include a quantitative l
the core beltline temperature determination of the material chemistries.
adjustament from Figure 3.6.1.
(Note: Analysi= of the first sample has been completed. The Figure 3.6.1 core beltline temperature adjustment curve reficcts the chemistry data obtained).
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NONTICELID 3.6 LIMITINC CONDITIONS FOR' OPERATION 4.6 5URVEILIANCE REQUIRENENTS B. Reactor Vessel Temperature end Pressure B. React:;r V;csel Teeperf.ture and Pressure 3.-
During all operation with a critical 3.
Neutron flux wires shall be installed in the
= ' '
reactor..other than for low level reactor vessel adjacent to the reactor vessel physics tests or at times when the wallget the core midplane level.; 1he wires reactor vessel la vented, the reactor.
shall be removed and tested during the first vessel shell and fluid temperatures refueling outage to experimentally verify s
specified in 4.6.A shall be at or above the calculated value of neutron fluence at the higher of the temperatures of Figure one fourth of the beltline shell thickness 3.6.4 where the dashed curve, "RPV that is used to determine the core beltline '
Belt 11ne Region *, is increased by the temperature adjustment from Figure 3.6.1.
~
core beltline temperature adjustment (Note: this surveillance requirement has fror Figure 3.6.1.
been completed and the core beltline -
temperature adjustment shown in Figure 3.6.1 now reflects the flux wire experimental.
results).
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"N HONTICELIA -
Baser 1.6 and 4.6:
A.-
Ec.jictor Coolant Heatun and Cooldown The vessel has been analyzed for stresses caused by thermal and pressure transients. Heating and cooling -
transients throughout plant life at uniform rates of 100*F per hour were considered in the temperature range of 100 to 546*F and were shown to be within the.wquirements for stress intensity and fatigue limits of Section III of the ASHE Boiler and Pressure Vessel Code.
During reactor operation, the temperature of the coolant in an idle recirculation loop is expected to remain at reactor coolant temperature unless it is valved out of service. Fequiring the c%1 ant tempera-in an idle loop to be within 50*F of the reactor coolant temperature before the pump is started ture assures that the change in coolant temperature at the reactor vessel nozzles and bettom head region are
'E within the conditions analyzed for the reactor vessel thermal and pressure transients.
t-During hydrostatic pressure t.esting, a cc.olant heatup or cooldown of 20*F in any one hour period has a negligible ef fect on the reactor operating liraits of Figure 3.6.2.
B.
Reactor Vessel Temperature and Pressure 4
Operating limits on the reactor vessel pressure and temperature during normal heatup and cooldown and during in-service hydrostatic testing were established using 10CFR50 Appendix G, May 1983 and Appendix G
?
These of the Summer 1976 or later Addenda to Section III of the ASME Boiler and Pressure vessel Code.
operating limits assure that a large postulated surface flaw, having a depth of 0.24 inches at the flange-I to-vessel junction and one-quarter of the material thickness at all other reactor vessel locations and discon'tinuity regions can be safely accommodated. For the purpose of setting these operating limits the In gyg., of the vessel material was estimated fre,m impact test data taker.
reference temperature, RT accorda6ce with requirements of the Code to which this vessel was designated and manufactured (1965 Edition including Summer 1966 A.hlenda).
MONTICEI.1D -
I'.;ises 3.6 rnd 4.6 - Contimied A General Electric Company procedure, designed to evaluate fracture toughness requirements for $1 der.'
plants where information may be incomplete, was used to estimate RT values on an equivalent basis to NDI the new requirements for plants which have construction permits after August 15, 1973.
fracture toughness of all ferritic steels gradually and uniformly decreases with exposure to fast The neutrons above a threshold value, and it is prudent at:d conservative to account for this in the operation of the reactor pressure vessel. Two types of information are needed in this analysis: 1) A relationship between the changes in fracture tougimess of the reactor pressure vessel steel and the neutron fluence (integrated neutron flux), and 2) A measure of the neutron fluence at the point of interest in the reactor, pressure vessel vall.
T The relationship of predicted adjustment of reference temperature versus fluence and the copper and nickel content of the core beltline materials given in Regulatory Guide 1.99,'. Revision 2, was used to i.
l t
define the core beltline temperature adjustment versus fluence shown on Figure 3.6.1.'
A relationship between full power years of operation,and neutron fluence has been experimentally, deter-mined for the reactor vessel. The vessel pressurization temperatures at any time period can be determined i.,.
from the thermal energy output of the plant and Figure 3.6.1 used in conjunction nith Figure 3.6.2 (pres-sure tests), Figures 3.6.3 (mechanical heatup or cooldown following nuclear shutdown), oriigure 3.6.4 5-(operation with a critical core). During the first fuel cycle, only calculated neutron fluence values g.
At the first refueling, neutron dosimeter wires which were installed adjacent to the; vessel were used.
~r vall were removed to experimentally determine the neutron fluence versus full power years of operation.
This experimental result was updated by testing additional dosimetry removed with the first surN111ance
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nasas 3.6 rnd 4.6~- Continutd)
Reactor vessel material samples are provided, however, to verify the relationship expressed by Figure 3.6.1.
Three sets of ~ mechanical test specimens representing; the base metal, weld metal, and weld heat affected zone (llA2) metal have been placed in the vessel and can be removed and tested as repired._ An analysis and report will be submitted to the Commission on all such surveillance specimens removed from-the reactor vessel in accordance with.10CFR50, Appendix H, Including information obtained on the level of integrated fast neutron irradiation received by the specimens and actual vessel material.
temperature'plus The requirements for cold bolt-up of the reactor vessel closure are based on the RTNDT 60*F which is derived from the requirements of the ASME Boiler and Pressure Vessel Code to which the vessel was built. The RT temperature of the closure flanges, adjacent head and shell material, and NDT stud material is a maximum of 10*F.
The minimum temperature for bolt-up is therefore 10' + 60* - 70*F.
[
The neutron radiation fluence at the closure flanges is well below 10 n/cm2 (E>l NEV)'and therefore 17 radiation effects will be minor and will not influence this temperature.
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NUCLEAR REGULATORY COMMISSION s
maimotow. o. c. rous SAf,J,,TY,JVA(UATIONBY.THEOFFICEOF.WUCLEARREACTOR. REGULATION i
V RELATED TO AMENDMENf N0. m :T0 JACILITY OPERATING. LICENSE NO..DPR.22 NORTHERh $TATFS POWER COMPANY MONTICEL(QNUCLEARGENHATING. PLANT opC.v1 lng,Jg:n}
1.0 INTRODUCTION
By lette-dated March 31,'1989, Northern States Power Company (the licensee) requested an amendment to the Technical Specifications appended to Facility Operating License No. DPR-22 for the Monticello Nuclear Generating Plant.
The proposed amendment would:
(1) revise the reactor vessel pressure vs. temperature (P/T) curves for consistency with Revision 2 of Regulatory Guide (RG) 1.99; 1
(2) add requirements for augmented inservice inspection (ISI) of piping susceptible to intergranular stress' corrosion cracking (IGSCC); and (3) revise the requirements for the periodic Type A containment integrated leak rate test (CILRT) to permit the use of the mass point test method.
A discussion of the proposed changes and the NRC staff's evaluation and findings relative to each are addressed in Section 2 of this Safety Evaluation.
1 2.0 DISCUSSIE R D,E,VM UAT,10N 2.1.1 Revised Pressure /T g erature Limits Pressure / temperature limits are included in facility technical specifications l
for the purpose of precluding conditions conducive to brittle fracture of reactor coolantsystem(RCS) materials. The fracture toughness of RCS materials is a function of the material chemistry and decreases as irradiation accumulates.
Revision 2 of Regulatory G#1de 1.99, issued May 1988, specifies a more accurate and more conservative means than previously used for predicting the effects of irradiation damage on RCS materials. Generic Letter 08-11. "NRC Position on Radiation Embrittlement of Reactor Vessel Materials and Its Impact on Plant Operations" requesteo licensees of operating reactors to reanalyze their P/T limits using the revised criteria.
In response to Generic Letter 88-11, the licensee applied to revise the P/T. limits in the Monticello Nuclear Generating Plant Technical Specifications, Section 3.6.
The NRC staff evaluated the proposed changes using the following NRC regulations and guidance: Appendices 1
l
. -. - ~ - - -... -- ~. -.. -
]
l 1 G and H of 10 CFR Part 50; the ASTM Standards and the ASME Code, which are referenced in Appendices G and H; RG 1.99, Rev. 2; Standard Review Plan (SRP)
Section 5.3.2; and Generic Letter 88-11.
Appendices G and H of 10 CFR Part 50 i
define specific requirements' for fracture toughness and reactor vessel material j
surveillance that must be considered in setting P/T limits.
SRP Section 5.3.2 i
describes an acceptable method for constructing the P/T limits.
Appendix G of
]
10 CFR Part 50 specifies fracture toughness and testing requirements for. reactor i
vessel materials in accordance with the ASME Code and, in particular, that the beltline materials in the surveillance capsules be tested in accordance with Appendix H of 10 CFR Part 50.
Appendix H of 10 CFR Part 50 requires the licensee to establish a surveillance program to periodically withdraw surveillance capsules from the reactor vessel.. Appendix H refers to the ASTM Standards which, in turn, require that the capsules be installed in the vessel before startup and that they contain test specimens made from plate,Jweld, and heat-affected-zone (HAZ) materials of the reactor beltline.
These tests define the extent of vessel embrittlement at the time of surveillance specimen capsule withdrawai in terms of the increase in reference temperature.
Appendix G also requires the licensee to predict the effects of neutron irradiation on vessel embrittlement by calculating the adjusted reference temperature (ART) and Charpy upper shelf energy (USE).
All surveillance capsules contain Charpy impact specimens and tensile specimens made from base metal, weld metal, and HAZ metal.
The licensee has removed one surveillance capsule.
The results from capsule 117C 3991 3-1 were reported in a Battelle-Columbus Laboratories Report BCL-585-84-2, Revision 1.
t The licensee used the method in RG 1.99, Rev. 2, to calcuir,te an ART of 140 degrees F. for the limiting plate material (I-15) at 32 d fective full power years (EFPY) at 1/4T (T= reactor vessel thickness) in tne Monticello beltline.
l The ART was calculated using Section 1 of RG 1.99, Rev. 2, because only one I
surveillance capsule has been withdrawn from the Monticello reactor pressure i
vessel.
The NRC staff performed a similar calculation and verified the I
licensee's ART to be conservative (see Table 1).
Substituting the ART of 140 l
degrees F. into equations in SRP 5.3.2, the staff verified that the proposed i
P/T limits for heatup, cooldown, and hydrotest meet the bsitline material requirements in Appendix G of 10 CFR Part 50.
In addition to beltline materials, Appendix G of 10 CFR Part 50 also imposes P/T limits based on the reference temperature for the reactor vessel closure flange materials.
Section IV.2 of Appendix G states that when the pressure exceeds 20%
of the preservice system hydrostatic test pressure, the temperature of the closure flange regions highly stressed by the bolt preload must exceed the reference temperature of the material in those regions by at least 120 degrees F for' normal operation and by 90 degrees F. for hydrostatic pressure tests and leak tests.
Based on the flange reference temperature of 10 degrees F. the staff has determined that the proposed P/T limits satisfy Section IV.2 of Appendix G.
Section IV.B of Appendix G requires that the predicted Charpy USE at end of life l
'be above 50 ft-lb.
Based on data from surveillance capsule 117C 3991 G-1 withdrawn at 7.63 f.FPY, the lowest measured irradiated Charpy USE of the material tested is 109 f t-lb for the beltline plate I-15.
To estimate the USE at 32 EFPY, l
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.3 thestaffsybtractedthefluencetowhichthesurveillgncecapsulewasexposed, 2.9E17 n/cm,-from the fluence at 32 EFPY, 5.1E18 n/cm.
The staff then referred to Figure 2 of RG 1.99, Rev. 2, for the predicted decrease in USE and calculated that the USE, at 32 EFPY, for the I-15 beltline plate material would be 85 ft--lb.
.This value is greater than 50 ft-lb and, therefore, is acceptable.
JIn' addition to revising the P/T limit curw s to meet RG 1.99, Rev. 2 criteria, the licensee also proposes that, during pressure tests, the reactor vessel bottom head temperature be monitored separately from the beltline region.
This
-would facilitate pressure testing by reducing the amount of non-nuclear heatup required for pressure testing.
In order for this to be acceptable, it is
.necessary that operators ~have the capability to monitor the beltline region temperature separately, :The licensee hm advised the staff that this capability is provided by redundant resistance temperature detectors (RTDs).
The staff concludes that the proposed P/T limits for the RCS for heatup, cooldown, leal. test, and criticality are valid through 32 EFPY because the limits conform
.to the' requirements of Appendices G and H of 10 CFR Part 50. The licensee's submittal-also satisfies Generic Letter 88-11 because the licensee used the method in RG 1.99, Rev. 2, to calculate the ART.
Hence, the proposed P/T limits may be incorporated ~into the Monticello Technical Specifications.
2.1.2 Auomented Inservice Inspection for IGSCC Genehic' Letter 88-01requestedlicenseestodescribetheirplansforreplacement, inspection, repair and 1.eakage detection of piping susceptible to intergranular
-stress corrosion cracking (ISGCC).
Among the items specified to be incleded in.the licensees' responses to Generic Letter 88-01 is an application to change
- the Technical Specifications to include a requirement that the Inservice Inspection Program for piping covered by the scope of the Generic Letter be in conformance with the staff positions'on schedule, methods and. personnel, and sample expansion. -The licensee's March 31, 1989 application proposed to invoke NUREG-0313,' Revision 2, as a reference for the augmented ISI requirements.
In a
' letter dated. September 27, 1989, the licensee revised the application to cite Generic Letter 88-01'as the reference for.the new augmented ISI requirements, f
The use of Generic letter 88-01 as the reference is consistent with the Model Technical Specifications provided in Generic Letter 88-01 and is acceptable,
'~
l.
(Note: The NRC staff will issue a separate evaluation of the licensee's complete
-response to Generic Letter 88-01 in the near future.
This change serves only tr i.
implement.that portion of the generic letter relating to Technical Specification
' requirements for augmented ISI.)
- 2.1.3 CILRT Test Method ll L
The proposed amendment would change Technical Specification 4.7.A.2.b to delete l
the requirement that the test method for the CILRT be in accordance with the i
1972 revision of ANSI N45.4.
ANSI N45.4-1972 specifies use of either the total time or point-to point method of containment integrated leak rate testing.
An exemption was issued to the licensee on October 21, 1988 to permit use of the superior " mass point" method pending revision of 10 CFR Part 50, Appendix J.
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-i t Appendix J'has since been revised and now permits use of the mass point method l(when used for a period of at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) as an alternative to the total time and point-to point methods specified by ANSI N45.4-1972.
The amendment-would bring the Technical Specifications into consistency with the revised 10 CiR Part 50,' Appendix J and is, therefore, acceptable.
3.0 ENVIRONMENTAL CONSIDERATION
. Pursuant to 10 CFR 51.21, 51.32 and 51.35, an environmental assessment and finding of no significant impact was published in the Federal Register on-September. 28, 1989.
Accordingly, based upon the environmental assessment, the Commission has determined that issuance of this amendment will not have'a significant effect on the quality of the human environment.
14.0' CONCLUSION We have concluded, based on the considerations discussed above, that (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be
~
conducted in compliance with the Commission's regulations, and (3) the 1=suance of this amendment will not be inimical to the common defense and
. security or to the health and safety of the public.
Principal Contributors: W 'Long J. Tsao (Section 2.1.1)
Dated:
November 2, 19,89 l.
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'i TABLE 1-The'NRC' Staff Calculated Adjusted Reference. Temperature for the Limiting.
Reactor Beltline Material at Monticello Nuclear Generating Plant Limiting Beltline Material-Plate material' Code No.t I-15
+
Copper Content 0.17%
~
Nickel Content 0.58%
Initial Reference Temperature 14 degrees F V-I
' Reactor Vessel Beltline
-Thickness (in.)
-5.06
- Chemistry Factor Used-t in Calculation 125.3 t
NeutronFhuence(n/cm)at32EFPY-2 At ID-0.51E19 At 1/4T 0.38E19
-At 3/4T 0.21E19 Fluence Factor
.v At ID 0.812
.At 1/4T 0.732 At 3/4T 0.575 Margin 34 at 1/4T
. ART: at 1/4T' at~ 32 EFPY:
140 degrees F.
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