ML20058M169
| ML20058M169 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 12/14/1993 |
| From: | Barkhurst R ENTERGY OPERATIONS, INC. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20058M173 | List: |
| References | |
| W3F1-93-0099, W3F1-93-99, NUDOCS 9312200185 | |
| Download: ML20058M169 (11) | |
Text
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V Entergy Operations, Inc.
w2:~ENTERGY im ia L A 'ni60-0751
' y 504.' 3 U,1 Ross P. Barkhurst 0;.,.m c e
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4n W3F1-93-0099 A4.05 PR e
December 14, 1993 Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C.
20555
Subject:
Waterford 3 SES Docket No. 50-382 License No. NPF-38 Technical Specification Change Request NPF-38-148 Gentlemen:
The attached description and safety analysis supports changes to the Waterford 3 Technical Specifications that will:
1.
remove the reactor vessel material specimen withdrawal schedule from the Waterford 3 Technical Specifications, and 2.
update the Waterford 3 Reactor Coolant System pressure-temperature (P-T) curves.
The first proposed change (above) is being submitted in an effort to eliminate unnecessary duplication of update submittals associated with the Waterford 3 reactor vessel material specimen withdrawal schedule.
Currently, since the withdrawal schedule is part of the Technical Specifications, updates to the schedule have to be submitted as a license amendment request (before implementation). Additionally,Section II.B.3 of Appendix H to 10 CFR Part 50 also requires the submittal of a proposed change to the schedule prior to implementation.
The proposed change to remove the withdrawal schedule from the Technical Specifications is consistent with guidance provided in NRC Generic Letter 91-01, " Removal of the Schedule for the Withdrawal of Reactor Vessel 180023 9312200185 931214 II h' Ob PDR ADOCK 05000382
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v Technical. Specification Change Request NPF-38-148
-W3F1-93-0099 Page 2 December 14, 1993 Material Specimens f rom Technical Specifications." The schedule is currently included in the Waterford 3 Final Safety Analysis Report (FSAR)
Table 5.3-10.
The proposed changes to. the specimen pull schedule,.will be incorporated in the FSAR.
The second proposed change (P-T Curve update) to the Waterford 3 Technical Specifications is being submitted as a result of Combustion Engineering's reanalysis utilizing inputs from the testing results of the Reactor Vessel specimen material pulled April 11, 1991. A copy of the P-T Curve and Reactor Vessel specimen withdrawal schedule analysis report is provided as Attachment C of this submittal.
In addition to the current Technical Specification change request, Reactor Vessel material specimen withdrawal schedule updates are being' submitted to the NRC under separate cover letter in accordance with Sect. ion II.B.3 of Appendix H to 10 CFR Part 50.
The proposed changes described herein have been evaluated in accordance with 10CFR50.91(a)(1) using criteria in 10CFR50.92(c) and it has been determined that the proposed changes involve.no significant hazards considerations. The Plant Operating Review and Safety Review Committees have reviewed and accepted these proposed changes based on the foregoing evaluation.
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Technical ' Specification Change Request NPF-38-148 W3F1-93-0099 Page 3 December 14, 1993
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i Should you have any questions or comments, please contact Paul Caropino at (504) 739-6692.
q Very truly yours, h
R1 JA R.P. Barkhurst i
Vice President, Operations Waterford 3 RPB/PLC/dc Attachments:
Affidavit NPF-38-148 cc:
J.L. Milhoan, NRC Region IV j
D.L. Wigginton, NRC-NRR I
R.B. McGehee N.S. Reynolds NRC Resident Inspectors Office l
Administrator Radiation Protection Division (State of Louisiana)
American Nuclear Insurers
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g UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION-In the matter of
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Entergy Operations, Incorporated
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Docket No. 50-382 Waterford 3 ' Steam Electric Station
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R.P. Barkhurst, heing duly sworn, hereby deposes and says that he is Vice President Operations - Waterford 3 of Entergy Operations, Incorporated; that-he is duly authorized to sign and file with the Nuclear Regulatory Commission the attached Technical Specification Change Request NPF-38-148; that he is familiar with the content thereof;. and that the matters set forth therein are:
true and correct to the best'of his knowledge, information and belief.
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'i dJt R.P. Barkhurst Vice President Operations - Waterford 3 -
l STATE OF LOUISIANA
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PARISH OF ST. CHARLES
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Subscribed and sworn to before me, a_ Notary Public in.and for the Parish.and State above named this
/@" day of D E C E"' ut
, 1993.
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Notary Public q
'1 My Commission expires O*"
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g, DESCRIPTION AND SAFETY AWALYSIS OF PROPOSED CHANGE NPF-38-148-This proposed change modifies the following Waterford 3 Technical Specifications (TS) as described below.
1.
TS 3.4.1.3 the double asterisks footnote is revised as follows:
- A reactor coolant pump shall not be started with one or more of the Reactor Coolant System cold leg temperatures less than or equal to 285 272 of unless...
2.
TS 3.4.1.4 the double asterisks footnote is revised as follows:
- A reactor coolant pump shall not be started with one or more of the Reactor Coolant System cold leg temperatures less than or equal to 285 212 of unless...
- 3. Surveillance Requirement 4.4.8.1.2 is revised as follows:
4.4.8.1.2 The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine, changes in material properties, at the intervals required by 10 CFR Part 50 Appendix H in accordance with the Reactor Vessel material surveillance program - withdrawal schedule in Table d.' 5 FSAR Table 5.3-10...
4.
Surveillance Requirement 4.4.8.1.2 (a & b) is revised as follows:
4.4.8.1.2 The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine changes in material properties, at the intervals required by 10 CFR Part 50 Appendix H in accordance with the Reactor Vessel material surveillance proaram - withdrawal schedule in Table 4r4-5 FSAR Table 5.3-10.
The results of these examinations shall be used to update Figures 3.4-2 and 3.4-3.
%e-adjusted-eeference temper +twe-result 4eg frem neuteen irradiatien shall be calculated based en the greater ef the fellewicy av Asteal-shift in the RTg as measured by impact testing ef 88114!
014&-weld-metalt br Predicted shift in RTg fer E8018/B0CA ::cid metal as deteveined by4egulatery-Guide-1,-99, " Effects ef Re;idual - Element +-en-Predicted Radiatien-Damage te React-er Vessel Materials."
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5.
Figure 3.4-2 is replaced.
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6.
Figure 3.4-3 is replaced.
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7.
TS Table 4.4-5 is deleted and replaced with a blank page stating
" Table 4.4-5 This Table has been deleted".
l 8.
TS 3.4.8.3 is revised as follows:
Applicability: Mode 4 when the temperature af any RCS cold leg is less than or equal to 335222 oF, MODE 5, and...
9.
Bases, page B 3/4 4-7 paragraph 2 wil' 'e revised as follows:
i The reactor vessel materials have been tested to determine their initial RTno7; the results of these tests are shown in Table B 3/4 4-1.
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Reactor operation and resultant fast neutron (E greater than 1 MeV).
NOT.
Therefore, an i
irradiation will cause an increase in the RT adjusted reference temperature, based upon the fluence, copper and j
nickel content of the material in question, can be predicted using FSAR Table 5.3-1 and the recommendations of Regulatory Guide 1.99, Revision 2 1, "E-f4eets of Residual Elements en Predicted Radiat4en Damage te Reaeter-Vessel Materials."
" Radiation Embrittlement of Reactor Vessel Materials" The heatup and cooldown limit curves...
i 10.
Bases, page B 3/4 4-7 paragraph 3 will be revised as follows:
The actual shift in RTwor of the vessel material will be established periodically during operation by removing and evaluating, in accordance with ASTM E185-82 and 10 CFR Part 50 Appendix H, reactor vessel material i
irradiation surveillance specimens installed near the inside wall of the reactor vessel in the core area. The surveillance specimen withdrawal.
schedule is shown in FSAR Table 4,4-6 5.3-10...
11.
Bases, page B 3/4 4-10 paragraph 3 will be revised as follows:
l The OPERABILITY of the shutdown cooling system relief valve or an RCS l
vent opening of greater than 5.6 square inches ensures that the RCS will
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be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the RCS cold legs are less than or equal to 285 272 of...
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12.
Bases, page B 3/4 4-10 paragraph 4 will be revised as follows:
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The restrictions on starting a reactor coolant pump in MODE 4 and with the reactor coolant loops filled in MODE 5, with one or more RCS cold i
legs less than or equal to 285 272 oF, are'provided...
13.) Bases, page B 3/4 4-10 paragraph 5 will be revised as follows:
he--automat 4c inclatien-setpcint of the shat-dewn-cooling inclation valve: is suff4e4ently high to preclude inadvertent' i:clation of the shutdown cooling relief valves during a prc :ure tr-ancient.
Existina Specification i
See Attachment A Proposed Specification See Attachment B Backaround r
(For Items 1.), 2.), fe.), 11.) and 12.) above]
-l Selection of 2/2 of is the LTOP alignment temperature is consistent with the l
original value (285 oF) selected. The Appendix G curve (P-T Curve) that j
ABB/CE generated (during analysis of reactor vessel specimen test results) is j
less restrictive than the original curve..This is due to improved analysis i
techniques, including linear elastic fracture mechanics.
The original curve crossed the 2500 PSIA pressure line at 285 of, while the proposed ~ curve crosses at 272 oF.
Requiring the alignment of the LTOP system at this' I
temperature ensures that pressure relief protection of the P-T curve is I
available at all modes of operation. This is consistent with the intent of Branch Technical Position RSB 5-2.
Although 272 0F_ provides additional margin-j for the plant, it is a result of removing unnecessary conservatism from the i
P-T Curves.
[For Items 3.), 7.) and 10.) above]
i The NRC issued Generic Letter 91-01, " Removal of the Schedule for the Withdrawal of Reactor Vessel Material Specimens from Technical Specifications", dated January 4,1991. - This generic letter informed licensees of a case (Joseph M. Farley Nuclear Plant) wherein the NRC approved c
p a request to remove the subject schedule from the Technical Specifications for that plant.
The NRC classified this as line-item ~ improvement consistent with the Commission Policy Statement on Technical Specification improvements.
The l
purpose for the allowed change was to eliminate an unnecessary duplication of requirements.
In addition to the Technical Specification >,quirements for reporting upgrades to the subject schedule,Section II.B ~
>f Appendix H to 10 i
CFR Part 50 requires tLe submittal to, and approval by, NRC of a proposed withdrawal schedule beto.e implementation.
Entergy is requesting NRC approval-for adopting the aforementinned line-item Technical Specification improvement.
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[For Item 4.)]
Supplement 1 of the staffs' Safety Evaluation Report (SER NUREG-0787)
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subsection 5.3.1.2 dated October 1981, discusses an exemption to the 10CFR50 Appendix H (II.b) requirements that required Waterford 3 to evaluate weld metal by use of actual impact tests of the weld metal 88114/045 or the predicted shift of weld metal E 880818/ BOLA per Regulatory Guide 1.99.
These requirements were incorporated into Technical Specification as 4.4.8.1.2 -items a and b.
This specific exemption was no longer required as a result of the
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1983 changes to Appendix H.
This fact appears in supplement 8 of NUREG-0787 r
subsection 5.3.1.2.
Therefore, this proposed change removes TS 4.4.8.1.2a and b as an administrative change.
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[For Items 5.) and 6.) above]
j The first Waterford 3 Reactor Vessel material test specimen capsule was pulled i
during Refuel 4 (April 11, 1991). The specimen capsule was sent to Babcox and
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Wilcox Nuclear Systems (B&WNS) for testing.
The resulting report stated that j
the current Pressure-Temperature curves may be extended to 10.5 EFPY based on the fluence calculations performed for capsule W-97 and the Charpy. impact test j
results.
It was further reported that based on current operations, Waterford j
3 should reach 8 EFPY during cycle 7 in early 1995.
Current Technical j
Specification Figures 3.4-2 and 3.4-3 require that the curves be updated prior j
to 8 EFPY of operations.
Test results were submi+.ted to the NRC on November 25, 1992 (W3F1-92-0369).
Note that the NRC had granted an extension beyond the one year requirement for submitting test results. In Entergy Letter W3F1-i 92-0369, a commitment was made to submit proposed changes to the. curves.and-schedules to the NRC as a Technical Specification amendment request by
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December 31,'1993.
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[For Item 9.)]
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Waterford 3 informed the NRC in Letter W3P88-1980 that the plant had applied the methodology of Regulatory Guide 1.99, Rev. 2.
This was in response to i
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Generic Letter GL'88-II, that encouraged Licensees to use the methods described in Revision 2 of Regulatory Guide 1.99 to predict the effect'of neutron radiation on reactor vessel materials.
This proposed Technical Specification change reflects the above commitment.
[For Item 13.)]
-Deletion of wording associated with the automatic isolation setpoint of the.
i shutdown cooling isolation valves is justified since the Shutdown Cooling Auto 1
Closure Interlock feature was removed in 1991 via DC-3260. The proposed i
Technical Specification change is consistent with the' November 1988 revision of Branch Technical Position RSB 5-2 "0verpressurization Protection of Pressurized Water Reactors While Operating at Low Temperatures." Branch j
Position 10 indicates that if pressure relief is from a low pressure system, I
not normally connected to the primary system, the over pressure protection function should not be defeated by interlocks which would isolate the pressure system from the primary coolant system. (See BlP ICSB3). The Waterford 3 j
design change (DC-3260) was completed in response to Generic Letter 88-17.
The change was formally submitted to the NRC in letter W3P90-0234 dated July 25, 1990, associated with Technical Specification 4.5.2.d.1 change request.
The current proposed deletion should have been included with that earlier Technical Specification change.
Descriotion The requested Technical Specification change at Waterford 3 involves _the-
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deletion of Table 4.4-5 and deleting the reference to the table number in a surveillance requirement (Surveillance 4.4.8.1.2) to avoid duplication of.
l submittal requirements. Future updates to the Waterford 3 Reactor Vessel material surveillance program withdrawal schedule will continue to be
. submitted to the NRC for approval prior to implementation in accordance with the Section II.B.3 of Appendix H to 10 CFR 50 requirement.
The withdrawal schedule is currently part of the Waterford 3 FSAR (Table 5.3-10) and will be updated.
-1 The requested Technical Specification change also submits proposed changes to-the Reactor Coolant System Pressure / Temperature-curves. A report describing-development of the requested curve updates is provided as Attachment D.
Safety Analysis 1
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i The propose'd change described above shall be deemed to involve a significant hazards consideration if there is.a positive finding in any of the following areas:
1.
Will operation of the facility in accordance.with this proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
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Response
No.
Although the Reactor Vessel material specimens withdrawal schedule will-d be removed from the lechnical Specifications,.the Technical Specifications bases will continue to provide background information on the use of the data-obtained from material specimens. Also, updates to the schedule will continue to be' submitted to the NRC for approval prior.
to implementation.
Operating the plant in accordance with the new, updated P-T Curves will l
assure preserving the structural integrity of the reactor vessel over the life of the plant. The pressure and temperature limits were developed in accordance with 10 CFR 50 Appendix G requirements.
Removing the requirements associated with the previous exemption to Appendix H (TS 4.4.8.1.2 items a & b) is purely an administrative change.
-j Therefore, the proposed changes will not significantly increase the probability or consequences of any accident previously evaluated.
j 2.
Will operation of the facility in accordance with'this proposed change create the possibility of a new or different type of accident from any-l 4
accident previously evaluated?
Response
No.
Removal of the Reactor Vessel material specimen schedule from the Technical Specifications has no impac.t on accidents at the plant, j
Updates to the schedule will -still be required to.be submitted to the
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NRC prior to implementation per Section' II.fL3 of Appendix H to 10 CFR. Part 50.
Also, updates.to the P-T Curves will not create a new or different type accident. The reactor vessel beltline P-T limits were revised ' applying the general guidance of the ASME Code, Appendix G procedures with the i
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necessary margins of safety for heatup, cooldown and inservice hydro test conditions.
The change to TS 4.4.8.1.2 items a & b is purely administrative.
Therefore, the proposed changes will not create the' possibility of a new-or different kind of accident from any accident previously evaluated.
3.
Will operation of the facility in accordance with this proposed change involve a significant reduction in a margin of safety?
Response
No.
Removal of the schedule for Reactor Vessel material specimen withdrawal from the Technical Specifications does not impact the margin of safety.
The schedule will continue to receive NRC review and approval prior to implementation of updates to the schedule.
Updates to the P-T Curves are provided to preserve the margin to safety to assure that when stressed under operating, maintenance, and testing the boundary behaves in a non-britt h manner and the probability of rapidly propagating fracture is minim ued.
The change to TS 4.4.8.I.2 items a & b is purely administrative.
Therefore, the proposed changes will not result in a significant reduction in the margin of safety.
Safety and Sionificant Hazards Determination Based on the above safety analysis, it is concluded that:
(1) the proposed change does not constitute a significant hazards consideration as defined by 10 CFR 50.92; and (2) there is a reasonable assurance that the health and safety _ of the public will not be endangered by the proposed change; and (3) this action will not result in a condition which significantly alters the impact of the station on the environment as described.in the NRC final environmental statement.
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