ML20031G898
| ML20031G898 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 10/20/1981 |
| From: | Poindexter C BALTIMORE GAS & ELECTRIC CO. |
| To: | Clark R Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8110260165 | |
| Download: ML20031G898 (12) | |
Text
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BALTIMORE GAS AND ELECTRIC
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CLARLES CENTER P.O. BOX 1475 BALTIMORE, MARYLAND 21203 cwais H Poisorx rca v<rr c.,0cNr
&MW MeI%I I NOW,IM'NG AND CONSTHuCfiON Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission
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Washington, D.C. 20555 g
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Mr. Robert A. Clark, Chief
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Division of Licensing A
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Calvert Cliffs Nuclear Power Plant Ib Unit No.1, Docket No. 50-317 Pressurized Thermal Shock to Reactor Pressure Vessel
References:
(a)
NRC letter dated 8/21/81 from D. G. Eisenhut to A. E.
Lundvall, Jr. same subject (b)
W.S. Pellini and P.P. Puzak, " Fracture Analysis Program Procedures for the Fracture Safe Engineering Design of Steel Structure NRC Report 5920," March,1963.
(c)
F.J. Lass, R.A. Braz, Jr., and J.R. Hawthorne,
" Significance of Warm Prestress to Crack initiation
., ring Thermal Shock," NRC/NUREG Report 8165, September,1977.
Gentlemen:
The referenced letter requested that certain information be provided for Calvert Cliffs Unit I (CCNPP-1) within 60 days, and that additional information be provided within 150 days.
Our letter of September 24, 1931 acknowledged your request. This letter provider our 60-day responses for CCNPP-1, using the item numbers from the referenced letter.
1.
RT NDT values for plates and welds of the cylindrical portion of the CCNPP-1 reactor vessel are provided in Enclosure 1.
form or " maps" showinc. the locations of inlet / outlet nozzles and welds. RTThe values for each plate and weid are indicated within a rectangle on the figures NDT Sheet (1) of Enclosure (1) lists the locations and provides the designations assigned to each location on the Figures.
NDT shif t predictions and the values of the peak vessel fluence atS used in the RT ok%
811026016S 811020 PDR ADOCK 05000317
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PDR F
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2 the I.D. for both the end of this year (12/31/81) and after four more calendar years of continued operation (thru 12/31/85). The methods used are summarized on Sheet (2).
a.
Initial R TNDT values are shov/a on Shees (3) of Enclosure (1).
b.
Adjusted RT>
values at the inner surface of the CCNPP-1 vessel predic'ted for the end of this year (12/31/81) are shown on Sheet (4) of Enclosure (1).
2.
In response to yoc request for the rate of change of RTND condition, Enclosure (2) provides a " map" of the predicted R~2-from the current I NDT values at the inner surface of the CCNPP-1 vessel af ter four more calendar years of continued operation, using the assumptions and methods described in Enclosure (1).
a 3.
We do not consider it appropriate to define an upper limit RTNDT value for 1
continued operation for the reasons discussed in item 4 below.
4.
The capability of a reactor vessel to withstand the effects of pressurized thermal shock cannot be represented independent of system geometry, system i
transient resoonse, and material behavior. The simple RTNW limit requested in reference (a) is inappropriate. Stress, as a function of tin;e ar'd intensity, is stro:: gly dependent upon the system response to a given tr.snsient or accident.
Since stresses are not imposed uniformly throughout the vessel RT nT must be N
determined as a function of position within the pressure vew L Tfils requires identification of variations in bot'. fluence and material coraposition.
The current issue of reactor vessel thermal shock involves rapid cooldown transients which cannot be adequately addressed by the simple fracture analysis technique based on the work of Pellini and Puzak (reference (b)). This work involved test experience with - essentially isothermal materials, vhich is a valuable concept and useful design tool for materials which experier.ce slowly varying temperatures. Research on the effects of thermal shock on reactor i
vessel integrity indicate that the " phenomenon of warm prestress should be considered in predictions of reactor vessel integrity and that this phenomenon may form a key element upon which to base assurance of vessel integrity..."
(Reference c).
Adoption of a simple RTNDT limit would not permit proper consideration of the key element of warm prestress or the other factors discussed above.
l The pressurized thctmal shock evaluation program presently underway for the CCNPP-1 reactor vessel considers in detail the effects of warm prestress, the values of the RT at all critical locations around the vessel, and the variation througYthe vessel wall. This program will result in a determination of RTNDT of the number of effective full power years of plant operation during which vessd pressure boundary integrity can be assured for the events which result in the most challenging pressure-temperature transients.
5.
Both termination of high pressure safety injection (HPSI; (HPSI pumps and charging }wmps), and control of feed rate are available to our operators to prevent overpressurizing the reactor vessel in the event of thermal shock.
Current procedures provide operator guioance to prevent overpressurization.
More detailed information with regard to HPSI termination criteria and feedwr.ter throttling criteria are proviued in Enclosure (3). Both the HPSI and the feedwater operating criteria are included to minimize the total reactor coolant system cooldown and subsequent repressurization while still ensuring that the core remains cooled.
Emergency Operating Procedure training for operators addresses these criteria and includes discussion of the conditions in the primary system during such a transient or accident.
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A two phase program to review operating procedures in greater detali has been initiated.
Phase I is a generic effort being performed by Combustion Engineering, Inc. (CE) for the CE Owners Group (CEOG). During this phase, CE will review existing procedures in light of the CEOG generic guidelines currently under development and information generated in related analytical work. Where it it warranted, revisions to existing procedures will be proposed and evaluated.
EvaluaticA will include engineering analysis, review of proposed changes by our staff, and validation of the altered procedure on a simulator. During Phase 2, detailed procedural changes will be made to the CCNPP emergency procedures for changes developed in Phase 1, including the retraining of the operating staff.
The status of this near-term effort will be provided in our 150-day response.
More detailed criteria for minimum acceptable procedures to prevent cold i
repressurization have not yet been established. These will have to be very transient-specific and it has become clear that they can only be addressed af ter the detailed analyses of transients or accidents by class have been completed (requested in Attachment I to Reference (a)).
Changes indicated by these analyses will be implemented in a deliberate and carefully controlled manner to preclude adverse interaction with other safety factors.
3 If you have any questions or comments concerning our approach in this matter, please let us know.
Very truly yours, J
STATE OF MARYLAND; CITY OF BALTIMORE: TO WIT:
Mr. C. H. Poindexter, being duiy sworn, states that he is Vice President of the Baltimore Gas and Electric Company, a corporation of the State of Maryld; that he executed the foregoing Amendment for the purpose therein set fortl; that the statements made in said Amendment are true and correct to the best of his knowledge, information and belief; and that he was authorized to execute the Amendment on behalf of said corporation.
.s Witness my Hand and Notarial Seal:
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voasam cc:
- 3. A. Biddison G. F. Trowbridge D.Jatfe R. E. Architzel I
Sheet 1 CALVERT CLIFFS UNIT #1 Designation on Figure Longitudinal Seam Welds -
Upper Shell Course 1-203 A, B and C Intermediate Shell Course 2-203 A, B and C Lower Shell Course 3-203 A, B and C Girth Seam Welde -
Upper to Intermediate Shell 8-203 Intermediate to Lower Shell 9-203 Shell Course Plates -
Upper D-7205-1, 2 and 3 Intermediate D-7206-1, 2 and 3 Lower D-7207-1, 2 and 3 A_ssumptions Residual Chemistry - as measured or:
Lappe~
0.25%
Best Estimate Paospl.orus 0.012%
Best Estimate Nichel (as measured)
Initial RT
- as measured or NDT
-20 F (20 upper bound for submerged arc welds)
Branch technical position MTEB 5-2 for plates Peak Vessel ID Fluence 18 4.77 EPPY 7.05 x 10 n/cm' (12/31/81) 19 2
7.97 EFPY 1.18 x 10 n/cm (12/31/85)
(assumes 0.8 c.f.)
Sheet 2 TRANSITION TEMPERATURE SHIFT METHODOLOGY For All Plate Materials -
Regulatory Guide 1.99 (Rev. 1):
(40 + 1000 (Cu
.08) + 5000 (P -.008) )
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10 For All Weld Materials -
Regulatory Guide 1.99 (Rev. 1) if Nickel content is more than 0.30%
Modified Regulatory Guide 1.99 if Nickel content less than 0.30%:
(90 + 600 (Cu
.24) )
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Enclosure (1)
Sheet (4) rRESSURE VESSEL VIEW ADJUSTED RT NDT
- F AT ti.77 EFPY CALVERT CLIFFS UNIT I Or etANT nom ER 04 12-31-81 3
SUTLET INLET INLET GUTLET INLET INLET.
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Enclosure (2)
PRESSURE WESSEL VIEW e
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ADJUSTED RT NDT F AT 7.97 EFPY Or stANT Nu=sER 04 CALVERT CLIFFS UNIT 1 12-31-85
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ISTLET INLET INLET GUTLET INLET INLET SUTLET mw I
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8-20 3l 52 )
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100 N
2-203 0-72C6-2 0-7206-3 0-T208-1 N
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'I AZluuTHAL LOCATlON m3T DEEREES e
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Sheet (1)
EXCERPTS FROM EMERGENCY OPERATING PROCEDURES Emergency Operating Procedure No. 5, Loss of Reactor Coolant (EOP-5)
Control of High Presure Safety injection Supplementary Action, item 2 2.
Af ter any SIAS6 p rate ECCS until RCS hot leg and cold leg temperatures are at least 50 F (Subcooled Margin Monitor or calculated value XC 913. or
... comparing lowest pressurizer pressure and incore thermocouples highest RCS Temperature) below saturation temperature for the existing RCS pressure and pressurizer level has been restored to greater than 0" indicated level,...
Supplementary Action, Item 8 CAUTION Do not override the automatic actions of Engineered Safety Features unless specifically directed to do so by this procedure or unless unsafe plant conditions would result from continued operation of the equipment (Such as violation of the reactor vessel pressure / temperature relationship limit).
Control of Feedwater Immediate Action, Item 2 Carry out the Immediate Actions of EOP-1: Reactor Trip.
EOP-1, immediate Action, Item 4:
Verify that steam generator feedwater regulating valves shut, and their bypasses open to 5% flow.
Immediate Action, Item 7 Verify either main or auxiliary feed operating to restore steam generator levels.
Supplementary Action, Item 6 If main feedwater flow is decreased to below 5% of full flow and concurrently the steam generator water level drops below -26", the auxiliary feedwater system must be used to feed the steam generators until level increases above -26".
Feed rate must be controlled to prevent excessive cooldown and depressurization of the RCS.
Do not exceed Technical Specification limits.
Enelcsure 3 sheet (2)
EOP-1, Supplementary Action 9 If main feed pumps are available, maintain steam generator level with the feed regrilating valves bypass valves. If main feed pumps are not available, establish Aux. Feed within 10 minutes and maintain steam generator level as per 01-32.
Technical Specification 3.4.9.1 is applicable at all times:
The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the limit lines shown on Figure 3.4-2....
i
Sheet (3)
Emergency Operating Procedure No. 4, Steam Line Rupture (EOP-4) i The immediate Actions of this procedure require the operator to isolate the steam generators by closing the following valves:
Main Steam isolation, Feed Isolation, Top and Bottom Blowdown, Main Steam Isolation to the Auxiliary Feed Pumps, and Drain isolation upstream of the MSIV's Feedwater is restored to the steam generators per Supplementary Action item 6 Open the main steam isolation (s) to the auxiliary feed pumps (s) (MOV-4070/MOV-4071) from the unaffected steam generator (s) and feed the unaffected steam generator (s) with the auxiliary feed pump (s)in accordance with 01-32 (Auxiliary Feed System).
Do not add water to a dry steam generator except in an emergency when no other heat sink is available.
Safety Injection is reset in accordance with Supplementary Action, item 9.
4 When pressurizer leve! is restored arid in the event of an upstream rupture, containment pressure has been reduced below 4 psig, block the pressurizer pressure signal and reset SIAS, CSAS, and CIS.
Additionally, the following actions are taken per EOP-1, Reactor Trip:
j Immediate Action, Item 4:
Verify that steam generator feedwater regulating valves shut, and their bypasses open to 5% flow.
(
Supplementary Action 9 t
if main feed pumps are availasle, maintain steam generator level with the feed regulating valves bypass valves. If main feed pumps are not available, establish Aux. Feed within 10 minutes and maintain steam generator level as 3
per 01-32.
i Technical Specification 3.4.9.1 is applicable at ali times:
i The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the limit lines showr on Figure 3.4-2....
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