ML19354E082
| ML19354E082 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 01/16/1990 |
| From: | POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
| To: | |
| Shared Package | |
| ML19354E043 | List: |
| References | |
| NUDOCS 9001250334 | |
| Download: ML19354E082 (9) | |
Text
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-l, ATTACHMENTI l
PROPOSED TECHNICAL SPECIFICATION CHANGES REGARDING AUGMENTED INSERVICE INSPECTION OF - -
MAIN STEAM AND FEEDWATER PIPING WELDS .,
(JPTS-89 013) t i
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l New York Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 DPR 59 9001250334 900116 Y PDR ADOCK0500ggg3 1
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JAFNPP 3.6 (cont'd) 4.6 (cont'd)
F. StructuralIntegrity F. StructuralIntegrity The structural integrity of the Reactor Coolant System shall Nondestructive irgions shall be padviirmd on the ' 'l be maintained at the level required by the original ASME Boiler and Pressure Vessel Code Class 1,2 and 3 acceptance standards throughout the life of the Plant. cornpormnts and supports in accordance with the requirements of the weld and support insennce ;nspec.,tui .
program. This insennce inspection program is based on ..
an NRC approved edition of, and addenda to, Section XI of the ASME Boiler and Pressure Vessel Code which is in effect 12 months or less prior to the beginning of the inspection interval.
G. Jet Pumps Whenever the reactor is in the startup/ hot standby or run - _
modes, all jet pumps shall be operable. If it is determined Whenever there is recirculation flow with the' reactor in the :
that a jet pump is inoperable, the reactor shall be placed in startup/ hot standby or run modes, jet pump operab8 sty a cold condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. shall be checked daily by veniying that the follomng conditions do not occur simultaneously:
Amendirent No. 96,13I 144
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JAFNPP .
3.6 and 4.6 BASES (cont'd) not required to be operable (reactor coolant temperature less in addition, visual inspection in accordance with the approved -
than or equal to 212'F and the reactor ' vessel vented or the ASME code will be made during penosc pressure and reactor vessel head removed). Permitting physics testing and operator training under these conditions would not place the ostatb Ms d did Wems.u N ' W~mNe plant in an unsafe condition.
M& the @ m d h M W piping system within the drywell. The i.ispection period is F. StructuralIntegrity based on the observed rate of defect growth from fabgue A pre-service inspection of the ASME Code Class 1 sWes WM by N AEC.
components was performed after site erection to assure the These studies show that thousand of stress cycles, at stresses system was free of gross defects. An initial inspection program beyond any expected to occur in a Reactor Coolant System, as detailed in Appendix F of the FSAR was developed and were required to propagate a crack. The test based on an approved edition of the ASME Code. -
The program has been expanded to include the requirements i of later, approved ASME Code editions and addenda as far as "
practicable. The importance of these inspections is recognized, and efforts to develop practical new alternative methods of assuring plant inservice integrity will continue. This -
inspection program should assure the detection of problem areas well before they represent a significant impact on safety.
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Amendment No.96, tikf 153
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ATTACHMENT ll l
SAFETY EVALUATION FOR PROPOSED l TECHNICAL SPECIFICATION CHANGES REGARDING ,
AUGMENTED INSERVICE INSPECTION OF MAIN STEAM AND FEEDWATER PIPING WELDS
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- (JPTS-89-013) i E
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New York Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 DPR-59
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Attachment ll
. SAFETY EVALUATION 'l ~
Page 1 of 5
' l. DESCRIPTION OF THE PFEO'OSED CHANGES
- This application for an amendment to the James A. FitzPatrick Technical Specifications deletes the augmented inservice inspection program imposed on the main steam and feedwater piping a welds. The proposed changes to the Technical Specifications are
A. Section 4.6.F.2, page 144; delete the following specification:
- 2. An augmented inservice inspection program is required for those high stressed circumferential piping joints in the main steam and feedwater lines larger than 4 inches in diameter, where no restraint against pipe whip is provided. The augmented in-service inspection program shall consist of 100 percent inspection of these welds per inspection interval.
B. Bases Section 3.6 and 4.6, page 153; delete the following paragraph:'
Several locations on the main steam lines and feedwater lines are not restrained >
to prevent pipe whip in the event of pipe failure at these locations. :The physical layout within the drywell precludes restraints at these points. Unrestrained high stress areas have been identified in these lines where breaks could result in pipe _ .
whip such that the pipe could impact the primary containment wall. Augmented inservice inspection of these weld locations shall be performed during each inspection period, ll. PURPOSE OF THE PROPOSED CHANGES Existing Specification 4.6.F.2 requires 100% of the carbon steel pipe welds inside the drywell on the Main Steam and Feedwater Systems (total of 34 welds) to be examined every 10 years. This i augmented inservice inspection (ISI) frequency was required by the Atomic Energy Commission staff in their Safety Evaluation Report (Reference 1), because it was not practicable to backfit pipe whip restraints at these locations on the main steam and feedwater high energy piping. As required by the augmented ISI inspection program, all 34 welds (12 on main steam and 22 on feedwater) have been examined at least once, in all cases, no flaws were noted.
This proposed technical specification change will delete the requirement for augmented inservice inspection. Future inspections of the affected welds will be in accordance with the standard Fitzpatrick ISI program which implements the requirements of 10 CFR 50.55a and 1
ASME Section XI (Reference 2). That is,25% of the affected welds will be inspected every 10 years. The proposed change will reduce radiation exposures by approximately 20 rem during a 10 year ISIinterval.
On October 27,1987 (Reference 3), the NRC modified General Design Criteria 4 (GDC 4) in 10 CFR 50, Appendix A, by allowing the use of leak before-break (LBB) technology to eliminate l from plant design bases the dynamic effects associated with high energy pipe ruptures. The
! modified rule permits the removal of pipe whip restraints and jet impingement barriers in operating nuclear power plants. In general, the LBB technology uses fracture mechanics analyses to show that high energy pipe flaws (cracks) are detectable, either by normal ASME .
' Attachment II
' SAFETY EVALUATION -
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inservice inspection techniques or by leakage monitoring systems, long before the flaws can -
grow to critical or unstable sizes and lead to large break areas such as a double-ended guillotine pipe wpture.
l A leak before break evaluation has been performed for the Authority by Structural Integrity .
Associates. A report of this evaluation, entitled " Evaluation of Leak Before-Break for Feedwater i
and Main Steam Piping Inside Containment at the James A. Fitzpatrick Nuclear Power Plant," is. ,
enclosed as Attachment lli. This evaluation concludes that the main steam'and feedwater piping systems comply with the general criteria of NUREG 1061, Volume 3 (Reference 4) and that augmented inservice inspections of weld locations which are unrestrained against postulated 3 pipe breaks are not necessary.
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lil. IMPACT OF THE PROPOSED CHANGES k
Fracture mechanics analyses (Attachment Ill), coupled with leak detection systems and ASME Section XI ISI, demonstrate that the probability of a main steam or feedwater pipe rupture (s) is i
extremely low. The leak-before-break evaluation (Attachment 111) was performed on the 24 inch .
l main steam piping and the 12.75 inch and 18 inch feedwater piping. Weld locations with the . '
least favorable combination of high stress and material properties were analyzed in each pipe size for bounding considerations. The leak-before-break evaluation is based on detecting pipe leaks, through leak detection systems or inservice examinations, and that detectable cracks -
(flaws) are inherently stable (i.e., will not rupture). Indirect pipe failure mechanisms such as water hammer, erosion / corrosion, and fire are also evaluated.
Leak detection l
The calculations assumed a leak detection limit of 5 gpm which is consistent with the design basis of the Fitzpatrick plant's leak detection systems (FSAR Section 4.10 and Technical -
Specification 3.6.D). The lower bound leak detection capability is generally considered to be 0.5 gpm which provides a leak detection margin of 10 as specified in NUREG-1061, Volume 3.
A long part through-wall flaw which is detectable by ultrasonic means is bounded by a 5 gpm leaking crack. Fatigue crack growth analyses of 360 part through-wall cracks was performed to assess the margin against rupture for pipes with long subsurface flaws. The results show that an assumed initial flaw with a depth of 15% of the pipe wall will not grow to a depth exceeding the -
l critical flaw depth in 40 years.
Crack stability 1
A postulated through-wall flaw which will leak at 5 gpm can be doubled in length and will remain stable under normal operating plus safe shutdown earthquake loads.
A loading safety factor of 1.4 is maintained for the highest stressed (limiting) location in each pipe size as specified in NUREG-1061, Volume 3.
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Attachment ll SAFETY EVALUATION -J Page3 of 5 q Other mechanisms Water hammer is not expected to have an adverse effect on the Integrity of the main steam and . $
feedwater piping inside the containment. A review of past ISI inspection reports notes that there ;
have been no water hammer induced pipe support deficiencies on the main steam and feedwater systems inside containment.
An erosion / corrosion Inspection program has been implemented at the Fitzpatrick plant. The j '
feedwater system is included in this program to detect wall thinning.
l The drywell is inerted with nitrogen gas during power operation. Postulated fires can not occur l on or near the main steam and feedwater piping within the containment.-
The carbon steel welds on the main steam and feedwater piping inside containment are not susceptible to intergranular stress corrosion cracking. All 34 welds have been examined at least once, and no flaws or defects have been noted. .
The proposed technical specification changes are administrative in nature. They do not involve -
any phys! cal modification to the plant; nor do they introduce any new failure modes. The changes reduce the frequency of weld inspections and eliminate dynamic effects from the .
design basis of main steam and feedwater high energy piping.
IV. EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATION Operation of the James A. FitzPatrick Nuclear Power Plant in accordance with the proposed amendment would not involve a significant hazards consideration as defined in 10 CFR 50.92, sinceit would not:
- 1. involve a significant increase in the probability or consequences of an accident previously . l evaluated. A leak before-break evaluation was performed on the 24 inch main steam p! ping
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and the 12.75 inch and 18 inch feedwater piping. Weld locations with the least favorable i
combination of high stress and poor material properties were analyzed in each pipe size for bounding considerations. This evaluation concludes that the main steam and feedwater-piping systems comply with the general criteria of NUREG-1061, Volume 3 and that -
augmented inservice inspections of the carbon steel pipe welds inside the drywell on the main steam and feedwater piping are not necessary. These fracture mechanics analyses, coupled with leak detection systems and ASME Section XI ISI, demonstrate that the probability of main steam and feedwater piping rupture (s) remains extremely low. The .
, propossd changes are administrative in nature and can not increase the consequences of postulated accidents. '
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- 2. create the possibility of a new or different kind of accident from those previously evaluated.. l The proposed changes are administrative in nature. They do not involve any physical modification to the plant; nor do they introduce any new failure modes. The changes reduce the frequency of weld inspections and eliminate dynamic effects from the design basis of main steam and feedwater high energy piping.
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l Attachment il .
, SAFETY EVALUATION :
Page 4 of 5 '
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- 3. involve a significant reduction in the margin of safety. The application of leak before-break q l technology to exclude dynamic effects from the design basis of main steam and feedwater i
- 'high energy piping is allowed by 10 CFR 50, Appendix A, General Design Criterion 4.
Fracture mechanics analyses, coupled with leak detection systems and ASME Section XI '
ISI, demonstrate that the probability of main steam and feedwater piping rupture (s) is extremely low. Any uncertainties associated with flaw geometry, analytical procedures, or i
leak detection are conservatively accounted for in accordance with NUREG 1061, Volume 3.
Significant safety margins are applied to taak detection, piping loads, and leakage crack sizes, such that the margin between the leakage crack size and the critical (unstable) crack size is a factor of two.
The proposed technical specification changes significantly reduce worker radiation exposures with an insignificant impact on offsite risk.
V. IMPLEMENTATION OF THE PROPOSED CHANGES Implementation of the proposed changes does.not impact the Fire Protection Program at the ,
FitzPatrick plant, nor will the changes impact the environment.
VI. CONCLUSION -
The changes, as proposed, do not constitute an unreviewed safety question as defined in 10 CFR 50.59. That is, they:
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- a. will not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis -
report; 1
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- b. will not create the possibility of an accident or malfunction of a type different from any evaluated previously in the safety analysis report;
[ c. will not reduce the margin of safety as defined in the basis for any technical specification; l
and
- d. involve no significant hazards consideration, as defined in 10 CFR 50.92.
Vll. REFERENCES
- 1. USAEC " Safety Evaluation of the James A. FitzPatrick Nuclear Power Plant" (SER), dated November 20,1972, pages 5-4 and 5-5. t
- 2. ASME Boiler & Pressure Vessel Code, Section XI,-1974 Edition through the Summer 1975 Addenda. ,
l Attachment il t.
SAFETY EVALUATION Page 5 of 5 :
- 3. Federal Register, Volume 52, No. 207, pages 41288 41295, Final rule amending General Design Criterion 4, October 27,1987.
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- 4. NUREG 1061, Volume 3, " Report of the U.S. Nuclear Regulatory Commission Piping Review Committee, Evaluation of Potential for Pipe Breaks," November,1984.
- 5. USAEC
- Supplement 1 to the Safety Evaluation of the James A. FitzPatrick Nuclear Power ,
l Plant" (SER), dated February 1,1973.
l USAEC " Supplement 2 to the Safety Evaluation of ti = ., vnes A. FitzPatrick Nuclear Power 4 1-6.
Plant" (SER), dated October 4,1974, t
- 7. Structural Integrity Associates, Inc. Report No. SIR 86-033, Revision 1,
- Evaluation of Leak-Before-Break for Feedwater and Main Steam Piping Inside Containment at the James A. .;
Fitzpatrick Nuclear Power Plant," April,1988. l
- 8. NSAC 141, Nuclear Safety Analysis Center, " Lead Plant Application of Leak Before-Break to !
High Energy Piping," January,1989.
- 9. NSAC 114, Nuclear Safety Analysis Center, " Applying Leak Before-Break to High Energy l Piping," November,1987.
- 10. EPRI NP-4991, Electric Power Research Institute," Application of the Leak Before-Preak l
Approach to BWR Piping," December 1986.
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