ML20040B250

Responds to NRC 811218 & 0821 Requests for Addl Info Re Pressurized Thermal Shock Phenomenon.Continued Operation Justified Since Evaluations Show Reactor Vessel Will Retain Integrity Throughout Life of Plant

ML20040B250
Person / Time
Site:	Maine Yankee
Issue date:	01/21/1982
From:	Randazza J Maine Yankee
To:	Clark R Office of Nuclear Reactor Regulation
References
MN-82-08, MN-82-8, NUDOCS 8201250311
Download: ML20040B250 (48)

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,4,gj (207) 023-3521 January 21, 1982 %.~,,rM MN-82JJ8 kWLlm United States fluclear Regulatory Corr <nission D Washinqton, D. C. 20555 B Attention: Of fice of Nuclear React or Pegulation ,' ' ? rg, Division of Licensing ( /d[7 Operating Peactors Branch #3 y ,. p / l,y j Mr. Robert A. Clark, Chief Suhject: Peactnr Vessel Pressurized Thermal Shock N. i, N,f< Peferences: 1. License DPR 36, Docket 50-309 2. Letter MYAPCo tn USNPC dated Cecerrber, 31, 1981, " Response to Item II.K.2.13 of NUREG-0737 ' Thermal Mechanical Report'" 3. Letter Southern California Edison to USNRC dated Decerrher 31,1981, " Transmittal of CEN-189, ' Evaluation of Pressurized Thermal Shock Effects Due to Small Break LOCA's with Loss of Feedwater for Combustion Engineering NSSS's,' December, 1981" 4 Letter USNRC to MYTPCo dated Decerher 18, 1981, recuesting additional information on pressurized thermal shock. 5. Letter MYAPCo to U5NPC dated Novenber 2, 1981, "Peactor Vessel Pressurized Thermal Shock" 6. Letter MYAPCo to USNRC dated September 29, 1981, "Peactor vessel Pressurized Thermal Shock" 7. Letter HYAPCo to USNPC dated September 3, 1981, " Maine Yankee Vessel Fluence Calculation" 8. Letter llSNRC to MYAPCO dated August 21, 1981 9 Letter MYAPCo to USNfC dated May 18, 1981 on Pressurized Thermal Shock. 10. " Analysis of the Maine Yankee Reactor Vessel Second Accelerated Surveillance Capsule" WCAP-9875, EPRI Research Project PP 1021-3 Topical Report, March, 1981 11. Letter MYAPCO to USNRC dated March 30, 1981, " Maine Yankee Vessel Fluence Calculation" 12. letter MYAPCo to USNFC dated March 17, 1981, " Reactor Vessel curveillance Capsula Analysis Peport - Errata" 13. Letter MYAPCo to USNRC dated February 27, 1981, "Peactor Vessel Surveillance Capsule Analysis Report" h0N 14 t etter MYAPCo to USNFC dated October 27, 1977, "Peactor s Vessel Material Surveillance Program" l 15. Letter MYAPCo to USNRC dated October 22, 1975 on / /l "thirradiated Properties of Maine Yankee Vessel Materials and Analysis and Evaluation of First Accelerated Surveillance Capsule" i I 8201250311 820121 PDR ADOCK 05000309 P PDR

i MAINE YANKEE ATOMIC POWER COMPANY U.S. Nuclear Regulatory Commission January 21, 1982 Attn: Mr. Pobert A. Clark, Chief Page Two

Dear Sir:

This letter is in response to the NRC's requests (Pef. 4 and 8) for additional information concerning the pressurized thermal shock phenomenon as it applies to the Maine Yankee Plant. Maine Yankee has previously submitted considerable information on this matter (Ref. 7, 5, 6, 7, 9, 10, 11, 12, 13, 14, and 15), and plans to develop additional information. Justification for Continued Operation: All available evaluations show that the Maine Yankee reactor vessel will 1 retain its integrity throughout the entire design lifetime of the plant (no cracking initiated) even if operator action less beneficial than that called for in plant procedures currently in use is assumed. Extrapolation of the maximum RTNDT for the limiting vessel material, including the critical weld, shows that the Maine Yankee reactor vessel material will not exceed a maximum RTNDT of about 300cF during the plant design lifetime even if the rate of vessel fluence accumulation is unchanged. Continued operation is justified because the Maine Yankee reactor vessel is capable of maintaining its inteority in credible pressurized thermal shock i situations both at the present time and for the foreseeable future without plant modifications or heroic operator actions. Contents of this Letter This letter includes the following appendices: ) Appendix A - Description of Maine Yankee program for further analysis or evaluation of PTS and response to four "150 day questions" of Pef. 8. Appendix B - Response to " request for additional information" of, Pef. 8. Appendix C - Response to " evaluation... and request for additional information" Ref. 4. l

M AINE Y ANK EE ATOMIC POWER COMPANY U.S. it> clear Regulatory Commission January 21, 1982 Attn: Mr. Robert A. Clark, Chief Page Three Appendix 0 - Pesults of Maine Yankee plant specific fluence calculations and RTNDT mapping. Appendix E - Results of Maine Yankee plant specific evaluation of vessel integrity under MSLB and A00 conditions. We trust this Information is satisfactory. If there are any cuestions, please do not hesitate to call. Sincerely, MAINE YANKEE ATOMIC POWER COMPANY ~ ohWCN$*j -- e 0 John B. Randazza Vice President JRR/bjp STATE OF MAINE ) )ss COUNTY OF KENNEBEC ) Then personally appeared before me, John B. Randazza, who, being duly sworn, did state that he is a Vice President of Maine Yankee Atomic Power Company, that he is duly authorized to execute and file the foregoing request in the name and on the behalf of Maine Yankee Atomic Power Company, and that the statements therein are true to the best of his knowledge and belief. Aa/sb $m t N Notary Public 'M 6 4

Appendix A DESCPIPTION OF MAINE YAf! VEE PROGRAM FOR FURTHER ANALYSIS OR EVALUATION OF PTS AND PESPOtEE TO FOUR "150 DAY QUESTI0t4S" of Ref. 8 This appendix describes Maine Yankee's plans for PTS activities in 1982 and beyond. Maine Yankee Pressurized Thermal Shock Prooram (beyond 1/1/82) Submittal of CE report on MSLB and A00. o Appendix E is a report prepared by Conbustion Engineering describing an investination of anticipated operational occurrences (A00's) and main steam line breaks (MSLB's) and their impact with respect to PTS on the Maine Yankee vessels. We are reviewing this report and will provide any further information soon. Plant specific model development. O Maine Yankee has begun upon a joint program with the Electric Power Research Institute (EPPI) to develop a consistent linked set of codes that will be used to evaluate transient and accident behavior of the Maine Yankee reactor and assess the inteority of the reactor vessel under these conditions. The licensed Maine Yankee steam line break model, based on RETRAN 01, will be upgraded to the level of RETPAtl 02 and further developed to provide capability to mndel transients and accidents over the extended time periods of interest in PTS analysis. The RETRAtl mndel will drive 3D flow and temperature calculations (COMMIX) for the cold leg and downtomer regions of the coolant system, and generate time dependent 30 vessel wall temperature distributions. The 3D temperature distributions from COMMIX analysis and pressure profiles from RETRAN will feed a vessel integrity evaluation code ( APAQUS) which will develop vessel wall stresses and apply fracture mechanics principles to determine vessel integrity. Application of plant specific models for further investigation of o vessol inteority. Once the plant specific model development described above is completed and validated, the models will he used for selection of limiting events, analysis of plant behavior and vessel integrity, evaluation of the need for and efficacy of system modifications, development of improved operation strategies, and investigation of the ef fects of multiple failures and operator errors. Development of operations strategy o Following analysis of the effects of operations strategy on vessel integrity, prudent changes to plant operating procedures will be Identified and implemented. A-1

Appendix A (Cont.) Procedures and training o As procedure changes are implemented, training on the changes and bases for them will be incorporated in the onsite retraining program and simulator sessions. During the code development period, we will also work to resolve o ouestions your staff might have about information submitted thus far. Prnaram Schedule We expect development of the methods described above will take 9-12 months, and application is likely to continue for another year or two. Responses to "150 day ouestinns" nf Pef. 8 NRC C0tCEPT: Peduction of further neutron radiation damage at the beltline by replacement of outer fuel assemblies with dummy assemblies or other fuel management changes; MAINE YANKEE PESPONSE: Low-leakage fuel management is planned for Cycle 7, and subseouent cycles with expected decreases in both the vessel average and peak location fluence rates. Preliminary calculations indicate a Cycle 7 fluence rate reduction of approximately 10-15% for the vessel average and 15-20% for the peak location relative to the historical fluence rates used in the present analysis. These benefits are considered typical of low leakage fuel management within the constraints of three hatches and extended cycles (i.e.,14-15 month refuelings). Low-leakage fuel management without these constraints might be expected to yield maximum fluence rate reductions in the 30-35% range. Use of dummy assemblies or high burnup assemblies at the core periphery appears unnecessary, and would likely reduce operating margin or necessitate a derate. These remain viable options and could be implemented if necessary in the future. NRC CONCEPT: Reduction of the thermal shock severity by increasing the ECC water temperature: MAINE YANKEF PESPONSE: The analyses reported in CEN-189 (Ref. 3) indicated good mixing and heatup of the cold injection water with the system water, using a conservative empirical calculation based on the preliminary results of EPRI sponsored tests. Since good mixing is expected, the sensititivy to the temperature of the ECC water is less than may have been originally thought. In the analyses performed to date, the ECC water supply has been assumed to be at the Tech. Spec. limit of 400F. Plant operating experience indicates that this is conservative. A-2

l Appendix A (Cont.) As we have previously indicated, the refueling water storage tank (RWST) l at Maine Yankee is eauipped with heaters used to meet the RWST minimum l temperature recuirements. This heater system could be utilized to maintain the tank temperature at a higher level. Maine Yankee will operate the heater system to maintain a higher tank temperature not to exceed 800F. NRC C0tCEPT: Recovery of RPV touchness by in-place annealing (include the basis for demonstrating that your plant meets the requirements in 10 CFR 50 Appendix G IV C); MAINE YANKEE RESPONSE: Maine Yankee believes that in-place annealing of the reactor pressure vessel is a feasible option. However, upon consideration of the materials properties of the vessel and the ability of the vessel to maintain its integrity under PTS conditions as indicated by all availahle studies, we do not believe this option needs to be exercised at this time. We will continue to keep informed of industry studies of the enoineering aspects of annealing. We believe that annealing would require considerable preparation and technical support to properly address all relevant issues. NRC C0tCFPT: Design of a control system to mitigate the initial thermal shnck and control repressurization. MAINE YANKEE RESPONSE: If it is determined that the operations strategy now in use (maintain pressure at or near saturation pressure plus an overpressure eauivalent to 500F subcooling) is or will be inadecuate, attention will be focused on possible control systems designs. Evaluation of control strategy is scheduled to proceed after a plant I specific evaluation system is in place, and this evaluation will include consideration of whether automation is necessary. This evaluation will consider the effects of failures and operational errors as they apply to existing control systems as well as new or modified control systems if such systems are indicated as necessry. Discussion of implementation schedules of NRC concepts to assure continued inteority. As discussed above, Maine Yankee will adopt a low leakage cor> design beginnina with cycle 7 (startup late 1982). Increasing RWST water temperature is being implemented. i j A-3 l

Appendix A (Cont.) Annealing and system changes are long term fixes. We will, of course, carefully consider the need for these drastic steps if it becomes evident that such measures may be recuired. At present, it appears that an adequate safety margin will exist throughout the Maine Yankee plant design lifetime, without changes to nperations strategy or plant design. Therefore, no schedule for changes is proposed beyond the activities discussed in the program description above for further demonstration of adequate vessel integrity margin and the low leakage core design and RWST heating. A-4

Appendix B PESPONSE TO "PEQUEST FOR ADDITIONAL INFORMATION" 0F ENCLOSURE TO PEF. 8 1. NRC PEOUEST: Geometry Geometrical description including design and as-built (when available) dimensions of the core, assemblies, shroud / baffle, thermal shield, downcomer, vessel, cavity, and surrounding shield and/or support structure. 2. NRC REQUEST: Material Description Pegion-wise material composition and material isotopic number densities (atoms / barn-cm) for the core, near-core regicos and RPV, suitable for neutron transport calculations. 3. NRC REQUEST: Neutron Source Present and expected EOL: a. Assembly-wise and core power history (EFPY). b. Pnd-wise and core power history (EFPY) for peripheral assemblies, c. Core average axial power histroy distribution. 4 MRC REQUFST: Vessel Fluence Description of available calculations of the vessel fluence including a. fluence values, locations, and corresponding power histories (EFPY), incuding 1/4T, 1/2T and 3/4T through the RPV. b. Description of available capsule-inferred vessel fluences including fluence values, locations, and corresponding power histories (EFPY). MAINE YANKEE RESPONSE: to Peauest 1-4: This information, assumed applicable to vessel flux, fluence, and materials properties calculation, has been provided previously (Ref. 7,11). In addition, we have been in communication with the NRC contracted analyst at Brookhaven National Laboratory, to whom we have provided information via telecon and also by informally providing a copy of the Westinghouse report on Maine Yankee fluence calculations. We consider the Westinghouse work to be the best available for Maine j Yankee at this point. i 1 We will continue to work with your analyst. i B-1

Appendix B (Cont.) J 5. NRC REQUEST: Surveillance Capsules a. Capsule materials, radial and axial dimensions and locations. h. Capsule fluence measurements, together with the accumulated power history (EFPY) and a description of the lead factors used to extrapolate the measurements to the peak wall fluence location. MAINE YANKEE RESPONSE: This information has been previuusly provided (Pef. 10, 12, 13, 14, and 15) and has been discussed further with the contract analyst at Brookhaven tutional Laboratory, i 6. NRC REQUEST: Vessel Welds Axial and azinuthal locations of vessel weld-seams with respect to the core. Overlay of current fluence map with weld locations. Identify the critical welds, vertical and circumferential, and give the weld wire heat i numbers. Give weld chemistry for the critical welds. For each weld wire heat number, report the estimated mean copper content, the range and the standard deviation, based on all the reported measurements for that weld wire heat. The welds may be surveillanca weldments for your vessel or others, nozzle dropouts that contain a weld, weld metal qualification data, or archive material. In the absence of any information, assume that copper content is at its upper limit (0.35 percent when using R.G. 1.99, Pev. 1) and that the nickel content is hich. MAINE YAtFEE PESPONSE: i Axial and Azimuthal Locations of Vessel Weld Seams With Respect to the a. Core - The vessel weld seams are shown in Figure B6-1 together with the initial Reference Temperature, RTNOT, for each weld. The core midplane and the extent of the active core are also indicated. This information is as submitted in CEN-189 ( Appendix C, Section C.6, Ref. 3). b. Overlav of Current Fluence Map With Weld Locations - 1 The neutron flux profile is shown relative to the vessel weld map in Fiqure B6-2. Values given in the figure are normalized iso-flux values. Peak flux values (normalized flux eaual to 1.0) occur at the 80 azimuthal locations (e.g., 80o, 170o, 260o, and 3500). c. Chemistry of Critical Welds - i B-2 i

Appendix B (Cont.) d The nickel, copper, and phosphorus content of the critical welds and the surveillance weld are given in Table B6-1. The wire heat and flux lot numbers are also indicated in the table. The basis for these values is discussed in CEN-189 (Section 6.3 and Appendix C, Section C.6, Re f. 3). No appropriate data are available to perform a statistical analysis of the copper content for the submerged arc welds. For the coated electrodes, the 0.07 w/o copper value reported i in the Table represents the upper bound value from 44 lots of similar coated electrodes as described in CEN-189, Section 6.3; the mean value for these 44. lots is 0.0275 w/o Cu with a range of 0.02 to 0.07 w/o and a standard deviation of 0.0236 w/o. j 7. NRC REQUEST: Systems Analysis a. Provide a list of transients or accidents by class (for example: excessive feedwater, operating transients which result from multiple failures including control system failures and/or operator error, steam line break and small break LOCA) which could lead to inside vessel fluid temperatures of 3000F or lower. Provide any Failure Modes and Effects Analyses (FMEAs) of control systems currently available or reference any such analyses already submitted. Provide the analysis of the most limiting transient or accident with regard to vessel thermal shock considerations. Estimate the freauency of occurrence of this event and provide the basis for this estimate. Discuss the assumptions made regarding reactor operator actions. b. Identify the computer programs used to calculate the limiting transient or accident. Indicate the degree to which the computer programs used have been verified and any other additional verification required to demonstrate that the computer program models adequately treat the identified important physical mocals (i.e., ECC mixing, heat transfer, and repressurization). MAINE YANKEE RESPONSE: CEN-189 (Ref. 3) and Appendix E discuss the transients and accidents for which evaluations have been performed for the l Maine Yankee vessel. Cooldown below 3000F does not appear credible for Anticipated operational occurrences (A00's) and Small Break Loss of J Coolant Accidents (SBLOCA's) with loss of feedwater. Cooldown below 3000F can occur for large Main Steam Line Breaks (MSLB's). The likelihood of events leading to fluid temperatures lower than 3000F is very low. An accident which by itself could produce such cooling would require a limiting failure in a catastrophic manner. Accidents or other events which by themselves do not lead to this degree of cooling but can do so when compounded by additional failures or operator errors can be hypothesized. Because their probahility is determined by the likelihood of the initiating event and the likelihood of the additional error (s) or failure (s), the likelihood of such an event cooling the plant to 3000F or helow is very low, although plant specific numerical estimates of probability are not available. B-3

Appendix B (Cont.) In the work reported thus far, operator action has been assumed to be less beneficial than called for in plant procedures. Sections 4.5 of Ref. 3, and Table ] of Appendix E describe operator actions assumed in evaluating a SPLOCA w/LOFW and MSLB's and A00's respectively. Instructions contained in plant procedures are discussed in Ref. 5. Operator instructions in plant procedures limit the repressurization in the event of cooldown to a pressure corresp'rding to 500F subcooling. This pressure control is not credited in the analyses in which it is assumed that the plant repressurizes to the PORV relief level (i.e. no operator action to limit pressure credited). The operator's instructions, i.e., control pressure at 50oF subcooling level, are familiar to the operations staff because of intensive training in inadeouate core cooling post-TMI. In addition, the 500F subcooling line is superimposed on the plant's MPT curve, with which the operators are ooite familiar, and the main control board mounted subcooling margin monitor gives a continuous readout of the degree of subcoolilng. Thus the operating strategy is relatively simple to carry out, extensive training las taken place, and it provices a large margin in RCS pressure. Finally, the repressurization assumed in the analysis is equivalent to a physical limit because (a) the PORV relief setpoint (equal to repressurization assumed), and (b) the HPSI pumps shutoff head is only sljahtly ahove the PORV setpoint, thus limiting repressurization by HPSI. (No positive displacement pumps are used at Maine Yankee for charging or HPSI). Methods utilized in PTS studies of Maine Yankee are described in Ref. 3 and Appendix E. B-4

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TABLE B(:,-/ MAINE YANKEE REACTOR VESSEL WELD MATERIALS Flux Lot Chemical Composition (w/o) Weld Seam Wire Heat ~ hickel Copper Phosphorus 1-203 A,B,1C L0EHa o,gg o,07b 0.009 2-203 A,B,&C 51989 3687 0.99c 0.35d 0.012d 3-203 A,B,&C 12008 Tandem 13253 Arc 3833 0.99c 0.22 0.015 EAGHa 0.97 0.07b 0.013 8-203 20291 3833 0.74e 0.35d 0.01 2d HADHa 0.94 0.07b o,01 2 f f 9-203 IP3571 3958 0.78f 0.36 0.015 f 33A277 3922 0.18 0.30 0.01 3 CBIJa 0.97 0.02 0.01 0 EODJa 1.04 0.02 0.009 3 CAFJ 1.01 0.03 0.008 a Coated electrode b Upper bound for coated electrode (see text and CEN-189) c Estimated nickel content (high nickel type wire or weld process; see text and CEN-189) d Regulatory Guide 1.99 upper bound prediction limit e Estimated nickel deposit content (based on wire analysis) f Surveillance weld analysis I

Appendix C PESPONSE TO " EVALUATION... AND PEQUEST FOR ADDITIONAL INFORMATION" (Ref. 4). 1. NRC REQUEST: RTtOT Values" Your response of November 2,1981, provided an initial RTNDT value of -3OoF for the weld metal, which we understand was for the surveillance weld and which matched the circumferential beltline weld material. For the longitudinal beltline welds, which are more critical for PTS events, your letter dated October 27, 1981, provided an initial RTNDT value of +100F as estimated per branch technical position MTEB 5.2. This is the value we intend to use unless you can support a lower value based on tests of archieval material, previously unreported data from vessel vendor's records, or a sound generic study of representative welds (in that order of prefe[ence. We cannot determine if the vessel ID fluence of 5.4 x 1028n/cm as of Septenber 30, 1981, is the fluence for the critical longitudinal welds. Please verify or provide the peak fluence at the critical longitudinal welds. When the above is provided we will then be able to verify your current RTNDT values or determine another value which we will use in our independent assessments. MAINE YANKFE PFSPONSE: The initial RTNOT values for the longitudinal beltline welds, as given in Figure 1, were submitted in Section C.6 of Appendix C to CEN-189. They are irrproved estimates of the Branch Technical Position MTEB 5-2 determined values as discussed in the Appendix and Section 6.4 of CEN-189. The +100F RTNOT represents an extremely conservative RTNDT value since it is 3.7 standard deviations above the mean for the generic weld population. Since the RTNDT values presented in Figure 1 were based on a combination of benchmark data from the surveillance weld, specific weld qualification test results, and a statistical analysis of representative welds, they should be used as conservative estimates of RTNOT for evaluation of the integrity of the Maine Yankee reactor vessel. The vessel ID fluence of 5.4 x 1018n/cm2 as of September 30, 1981, provided in the November 2, 1981 letter was a conservative estimate which was subsequently revised. The peak fluence at the ID of the Maine Yankee reactor vessel as of Decenher 31, 1981, (5.904 effective full power years 18n/cm, as described in Section C.5 of 2 at 2630 tnt) is 5.00 x 10 Appendix C to CEN-189, and Appendix D. Information requested concerning the neutron fluence values for the critical longitudinal seam welds is indicated by the iso-flux contours in Figure B6-2. The maximum relative flux and the estimated fluence at the inside surface of the vessel for each of the critical weld seams are as follows: K'ximum Relative Estimated Neutron Wold Seam Azimuthal Location Neutron Flux Fluence 12/31/81 2-203 30 0.432 0.216 x 1019n/cm2 2-203 150o 0.496 0.248 x 1019n/cm2 2-203 27no 0.828 0.414 x 1019n/cm2 3-203 9:o 0.828 0.414 x 1019n/cm2 3-203 2130 0.432 0.216 x 1019n/cm2 3-203 330o 0.496 0.248 x 1019n/cm2 C-1

Appendix C (Cont.) 2. NRC REQUEST: Rate of Increasino RTNDT Before we can verify your end of life RTNDT values we must have the increase in fluence per EFPY at the critical longitudinal welds. This is particularly necessary if you contemplate changing core configurations. Also we request the copper and nickel content of the critical longitudinal ] welds. MAINE YANKEE PESPONSE: The increase in fluence per EFPY at the critical j longitudinal welds may conservatively be assumed eoual to the historical i fluence rate used in the present analysis, representing cumulative fluence to December 31, 1981. Low leakage fuel management is planned for Cycle 7 l and subsecuent cycles with expected decreases in both the vessel average I and peak location fluence rates. As such, the historical fluence rate is i conservative for establishing vessel embrittlement. i The copper and nickel content of the critical longitudinal welds is contained in Table C6-1 of CEN-189, Appendix C.

3. & 4.

NRC REQUEST: RTNDT Limit and Basis for the Limit Since the "60 day" response stated that you do not consider a limit on PTNDT to be an appropriate basis for continued operation, the staff needs to develop a quantitative criterion for continued operation that, if implemanted, would assure maintenance of an acceptable low risk of vessel failure from PTS events for the near-term, pending longer term results of more detailed analysis or research. We will be developing this criterion considering recommendations that you may provide in your "150 day" response. MAINE YANKEE RESPONSF: Maine Yankee endorses the concept of a definitive criterion and appropriate basis for continued operation which assures low risk attributable to reactor pressure vessel failure. 2 Some concepts which could be considered are: a. Limit on maximum RTNDT (staff's proposal). 1 h. Limit on maximum fluence, c. Limit on extent of cooldown for events selected as appropriate to PTS considerations. d. Limit on crack initiation for events selected as appropriate to PTS considerations. e. Limit relating to crack arrest for events selected as appropriate to PTS consideration. f. Limit on risk from PTS events. We helieve approach (e), a criterion relating to ability to survive a PTS event without vessel failure, would most directly deal with PTS provided agreement can be reached on analytical methods to be employed and challenges to be withstood. C-2

Appendix C (Cont.) In the short term, we believe, at least for Maine Yankee, alternatives (a) and (b) would be acceptable and would be simple to administer because RTNDT and peak fluence are relatively straightforward to gauge. We are concerned, however, that these alternatives, if not properly prescribed, could become unnecessarily limiting due to imposition of excessive conservatism. This would have the effect, perhaps, of forcing " fixes" which would in reality not be needed and might degrade safety from other perspectives. Our acceptance of these alternatives would also be predicated on the availability of staff resources for review and approval of less restrictive, more sophisticated limits if and when it becomes necessary to apply them.' 5. NPC PEQUEST: Doeratnr Actions The extent to whiuh the overall concern of thermal shock which is being emphasized at Maine Yankee has been the subject of discussion between staff personnel (Project Mananer and Resident Inspector). From these discussions we recognize that PTS has received some emphasis in training and procedures and operators at Maine Yankee are sensitive to thermal shock considerations. However, we cannot determine from your "60 day" response to our letter of AJ0ust 21,1991 the degree of emphasis which is currently placed on the need for changes in procedures, training and manaaement involvement. We reouest that ynu expand your response to provide us a more detailed disucssion of what steps have been taken to ensure that your operators have a firm grasp of the issue and can be expected to cope with the events which serve to initiate PTS. MAINE YANKEE PESPONSE: The training given on the topics of PTS is as follows: Segment 1. MPT curve lesson plan a. approximately 3 hour b. Incorporated PTS concerns. 1. More fluence seen than expected by design calculations. 2. Weld material makeup. a. How this affects the brittle fracture limits. 3. Capsules taken from vessel. a. Results. b. Testing methods. 4. Methods considered to avoid PTS. a. Injection temperature. h. Injection pressure control. c. Maintain RCS pressure near the 500F min. AT line of the MPT curve. 5. Method used to avoid PTS. a. See 4 (c) above. c. RTNOT 1. Old curves vs. new curves. 2. Heatup and cooldown curves. f a. Calculation of and why they shift. C-3

Appendix C (Cont.) Segment 2 Casualty Procedure Instruction a. Includes discussion of changes to procedures for LOCA, steam generator tube rupture, and steam line break. 1. Operate HPSI system to maintain the RCS pressure near the 500F min AT line of the MPT curve to prevent overpressurizing a locally cooled reactor vessel. Segment 3 Simulator a. Includes discussion of casualty procedures for LOCA, steam generator tube rupture, and steam line break. Current Part'cipation in PTS Trainino i Segment 1 1. Requalification program will offer class to operating crews by June, 1982. 2. Current SRO program has received class. 3. Current RO program will receive class prior to license exam in May. Segment 2 1. Recualification program will offer casualty procedure instruction to operating crews by June, 1982. 2. Current SRO and PO programs will receive this instruction prior to licensing (SRO-Feb., CR)-May). 3. Upon receipt of new casualty procedures a training program will be initiated for requalification training. Segment 3 1. Operating crews have completed this program. 2. SRO program has completed this program. 3. RO program will complete prior to licensing in May. C-4

Appendix 0 RESULTS OF PLANT SPECIFIC FLUEfCE CALCULATIONS AND RTNDT MAPPING contains the results of discrete ordinates transport theory (DOT) fluence calculations performed by Westinghouse Ele:;tric Company under contract to Paine Yankee to determine the fast neutron fluence (E 71.0 MEV) accumulated by the Maine Yankee reactor pressure vessel. These results were utilized by Conbustion Engineering Company in determining the RTNDT Of vessel and weld material at various locations and points in plant lifetime used in the vessel integrity analyses reported in Pef. 3 (SELOCA w/ loss of feedwater, II.K.2.13 scope) and in investigations of Maine Yankee pressure vessel intenrity under A00 and MSLB conflitions shown in Appendix E. See Appendix C of CEN-189 (Pef. 3) for further discussion of the use of this information. contains RTNDT maps prepared by Combustion Engineering showing initial and post-irradiation toughness properties at critical locations in the Maine Yankee vessel, thte: Many figures and tables in the enclosures to appendix D are reproduced in Ref. 3 (CEN-189) also. They are included here for completeness. D-1

MAME - HAllHEE 'AWMICPOWER COMPM * ,s,,wonees,enno,o ENGINEEMNG OFFICE FR AMINGH AM. MASSACHUSETTS 01703 S 617 872 8100 October 13, 1981 DCC-MY-CE-81-3 NED-81-662 RECEIVED MY K.1.3.5 I0d ~' Mr. R. C. Jacques Project Manager ., _, _.. [ Tf,C :' ~ ' ~ " " N Combustion Engir.eering, Inc. 1000 Prospect Hill Road Windsor, Ct 06095

References:

(a) Letter G. T. McDonough to J. Atkinson dated July 16, 1981. (b) Letter, CE-TMI-218 from J. H. Hutton to CE Owners' Group dated August 7, 1981. (c) Letter, DCC-MY-CE-81-2 from C. M. Solan to J. H. Hutton, dated September 23, 1981. (d) Letter, FSD-RSA-81/387 from S. L. Anderson - W to G. M. Solan dated October 6,1981.

Subject:

Fluence Data for Reactor Vessel Materials Properties De a r Pa y, This letter transmits fluence analysis results for the Maine Yankee pressure vessel. These results should be used as input for determination of reactor vessel materials properties as described in Reference (a). This data transmittal fulfills the data request provided in Section C of Reference (b) as acknowledged in Reference (c). Please note that these results are identical to that telecopied to R. C. Jacques and J. Cavanaugh the week of October 5, ao clarified in subsequent conversations. The fluence data requested in Reference (b) can be constructed from data provided in Reference (d), Attachment A. Included in Reference (d) are calculated azimuthal, radial and axial flux distributions. The relative radial and axial distributions in Reference (d) are plotted in Figure 1, normalized to the pressure vessel inner wall location. Also included is a normalization factor (1.51) which corrects the calculated fluence level of surveillance capsule 263 to the measured value. The distribution of fluxes should be employed along with the normalization factor to construct the three dimensional fluxes in the Maine Yankee pressure vessel. The resultant fluxes are characteristic of the projected average power distribution for operation of Maine Yankee to December 31, 1981. This renresents 5.904 Effective Full Power Years (EFPY) of operations at the present rated power level, 2630 Mw(t). Table 1 provides a detailed operating history. The pressure vessel fluence distribution as of December 31, 1981 can be obtained by multiplying the three dimensional flux distribution obtait.ed by the prescription above by 1.863 x 108 seconds (5.904 EFPY).

Mr. R. C. Jscquac October 13, 1981 Page 2 The normalization f actor provided (1.51) is acknowledged tc produce a conservative fluence distribution. Maine Yankee and Westinghouse are investigating the discrepancy between the measured and calculated flux results and anticipate justification of a more appropriate normalization factor at a future date. Any assistance in this area, based on your experiences, is appreciated. 'O. ,.Please note that the azimuthal distributions in Reference (d) are referenced to a line drawn from the core center perpendicular to the flat portion of the core shroud. Figure 2 provides the orientation of the core abr{ud to the core inlet and outlet nozzles. Notice the lack of symmetry chdracteristic of a three loop plant. lL 3,If 3dditional information or clarification is required, please contact us. L -}h, Sincerely yours, MAINE YANKEE ATOMIC POWER COMPANY . M. Solan, Senior Engineer Reactor Physics Group GS/bjp Enclosure cc: J. H. Ca rrity J. H. Hutton J. Cavanaugh

. = _ - -. _. -. - -,b ,x s-t j l n~,.' 3, v v f 0c't ob er-13, 1981 , DCC-MY-CE-81-3 ' -^ a. '. ~ l Table 1 1 s Maine Yankee Cycles 1, l A, 2-6 4 Effective Full Power Years of Operation ^ to December 31, 1981 i i a Cycle Cycle Cycle Cumulative Exposure Loa ding Energy EFPY ~. Cycle (MWD /MTU) (MTU) (MWD) at 2630 MWt' 1 10367 81.540 845,325 0.880 1A 4500 83.110 373,995 1.269 2 17395 80.885 1,406,995 1 3 11075 83.065 919,945 ' 2.734 3.692 4 10496 81.843 859,024 4.586** [ 5 10796 83.034 896,435 5.519

c. -

6 4500* 82.248 370,116 5.904 i i Total 5,671,835 t t

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s October 13,1981 DCC-MLCE-81-3 E OUTLET 180 Figure 2 N0ZZLE (-- -m Maine Yankee 8 j ( Core Shroud and Capsule 263 p Orientation Relative to Core Vessel INLET INLET N0ZZLE e N07.ZLE f CORE 9 t [ \\ g g , SHROUD ~ t ( \\ Perpendicular 3 to Core Shroud 1 0 260 ,,o ,l

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~ g ~ 2700 9 l~~ Angular Location of ~ 1 H Capsule 263 3 I / Capsule 263 - g i Distance from core center to ) \\ g a center -line of the vessel \\ I 8 f wall assembly tube (as-built) \\ OUTLET D / OUTLET N0ZZLE Top of tube 85-29/64 in N0ZZLE Bottom of tube 85-3/8 in k[ h L_._ _ _i k? INLET 0 N0ZZLE I

l $$h#CE-j FSD-RSA-81/387 ~ i A T T A C 11 M E N T A Westinghouk Water Reactor N11 ear Techncicgy Otv:sen Electric Corporation Olvisions g,333 Pittsburgh Pennsytvania 15230 (412) 373-5165 October 8, 1981 Mr. George M. Solon Yankee Atomic Electric Company 1671 Worcester Road Framingham, Massachusetts 01701

Subject:

Reactor Vessel Fluence - Maine Yankee Enclosed for your information and use are the results of discrete ordinates transport calculations which were performed to determine the fast neutron (E>1.0 Mev) flux distributions within the Maine Yankee reactor pressure vessel. The data are presented in tabular form such that a three dimensional flux dis-tribution may be constructed using the following relationship, 4(R,Z,0) = 4(e)F(R)F(Z) where 4(e) represents the peak neutron flux at the pressure vessel-cladding interface obtained from Table 1 while F(R) and F(Z) are the relative radial and axial flux distributions given in Tables 2 and 3, respectively. The corresponding calculated fast neutron flux level at the center of sur-veillance capsule 263 is 3.25 x 1010 n/cm2-sec. Thus, the calculated lead factor for that capsule is 1.83. It should be noted that BMI riginally reported a measured flux value for cap-sule 263 of 4.76 x 1010 n/cm -sec (BCL-585-21) and subsequently revised the result to 4.91 1010 n/cm2-sec. (Letter, R. S. Denning to E. C. Biemiller, March 2,1931. ) In view of the large discrepancy between the reported measure-ment and the calculated value I believe that at this time, it would be prudent to normalize the calculated flux distributions to this maximum measured value. Thus, for fracture mechanics analysis, I would recommend multiplying the data in Table 1 by the ratio 4.91 x 1010 3.25 x 101U 1 believe that using this approach will produce conservative results for the Maine Yankee pressure vessel. Prior to issuance of the final report on these --l

Reactor Vessel Fluence Maine Yankee FSD-RSA-81/387 calculations the differences between calculations and measurement will be re-viewed and comments will be offered regarding potential reasons for the dis-crepancy. If you have any questions or require additional information, please feel free to contact me. Sincerely, WESTINGHOUSE ELECTRIC CORPORATION S. L. Anderson, Engineer Radiation and Systems Analysis i SLA/js i 4 4 s. i I l l ( I i I ) i "1 ~

i AZIMUTHAL DISTRIBUTION OF FAST NEUTRON FLUX f (E>1.0 Mev) AT THE INNER RADIUS OF THE MAINE YANKEE REACTOR PRESSURE VESSEL l 1 Azimuthal Neutron Flux Angle (Des.) (n/cm' Sec) 10 0.0 1.78 x 1010 i 1.0 1.74 x 10 10 l 2.0 1.69 x 10 l 10 3.0 1.67 x 10 10 l 4.0 1.72 x 10 ) 5.0 1.73 x 1010 10 6.0 1.72 x 10 10 l 9.0 1.56 x 10 10 12.0 1.30 x 10 10 15.0 1.05 x 10 lI 18.0 8.96 x 109 9 21.0 8.88 x 10 9 24.0 9.33 x 10 9 27.0 9.74 x 10 9 30.0 9.86 x 10 I 33.0 9.50 x 109 9 I 36.0 8.53 x 10 9 39.0 7.90 x 10 9 42.0 7.48 x 10 9 I 45.0 7.37 x 10 l Note: 1) 0.0 is referenced to the perpendicular to the core shroud. i

2) Calculations were based on burnup averaged core power distributions for Cycles 1 through 6.

j i i ._ _._.~.

RELATIVE RADIAL DISTRIBUTION OF FAST NEUTRON FLUX (E>l.0 Mev. WITHIN THE MAINE YANKEE REACTOR PRESSURE VESSEL Depth Into Vessel (cm) Relative Flux 0.00 1.000 1.00 0.934 2.00 0.850 3.00 0.754 4.00 0.668 5.00 0.587 6.00 0.513 7.00 0.448 8.00 0.389 i, 9.00 0.337 10.00 0.291 11.00 0.253 12.00 0.220 13.00 0.190 14.00 0.164 15.00 0.142 16.00 0.123 17.00 0.106 18.00 0.0908 19.00 0.0770 20.00 O.0650 1 21.00 0.0541 21.97 0.0455 4 88

t RELATIVE AXIAL DISTRIBUTION OF FAST NEUTRON FLUX (E> 1.0 Mev) ~ i WITHIN THE MAINE YANKEE REACTOR PRESSURE VESSEL j Height Above Relative Flux ( Midplane (cm) Depth = 0.0 Depth = 6.0 Depth = 15.0 Depth = 21.0 i r j 0.0 1.000 1.000 1.000 1.000 10.0 1.000 1.000 1.000 1.000 20.0 1.000 1.000 1.000 1.000 30.0 1.000 1.000 1.000 1.000 40.0 1.000 1.000 1.000 1.000 50.0 0.999 0.999 0.999 0.999 60.0 0.997 0.997 0.997 0.997 70.0 0.994 0.994 0.994 0.994 i 80.0 0.985 0.985 0.985 0.985 i ] 90.0 0.975 0.975 0.975 0.975 [ f 100.0 0.926 0.926 0.926 0.926 110.0 0.863 0.863 0.863 0.863 t i 120.0 0.792 0.792 0.792 0.792 130.0 0.704 0.704 0.704 0.704 140.0 0.600 0.600 0.600 0.600 j t- ( 150.0 0.463 0.463 0.463 0.463 ( 160.0 0.324 0.324 0.332 0.354 I f 170.0 0.210 0.210 0.223 0.261 175.0 0.166 0.166 0.178 0.221 l 180.0 0.127 0.127 0.140 0.189 j 185.0 0.0960 0.0960 0.109 0.160 l f 190.0 0.0700 0.0700 0.0831 0.136 195.0 0.0506 0.0510 0.0628 0.114 200.0 0.0350 0.0371 0.0468 0.0955 [ 205.0 0.0243 0.0269 0.0346 0.0820 t 210.0 0.0158 0.0181 0.0251 0.0705 i t 1 215.0 0.0100 0.0121 0.0184 0.0615 l 220.0 0.00615 0.00765 0.0132 0.0536 l i 225.0 - 0.00368 0.00475 0.00960 0.0471 230.0 0.00221 0.00288 0.00695 0.0423 235.0 0.00129 0.00188 0.00561 0.0387 l -4 240.0 7.55 x 10 0.00122 0.00461 0.0343 -4 -4 245.0 4.58 x 10 8.11 x 10 0.00380 0.0308 l 250.0 2.86 x 10 5.46 x 10 0.00317 0.0274 f -4 -4 -4 255.0 1.86 x 10 3.72 x l'0-4 0.00266 0.0247 ~4 -4 260.0 1.27 x 10 2.67 x 10 0.00226 0.0221 265.0 8.80 x 10 " 1.98 x 10 U.00193 U.0200

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C E P:w;r Sy:t:ms Tel 203/688-1911 ~ Comouston Eng1reennq tc % 9')297 1000 Prosp ct kil Road W,n e ar Ccr' rec' cut 06095 H POWER MYC-81 -109 SYSTEMS December 2,1981 Mr. J. H. Garrity CErlTRAL MAINE POWER COMPAflY 83 Edison Drive RECEIVED A*:gusta, Maine 04336 Subj ec t: Reactor Vessel Materials Property Data

Enclosure:

(1) RV Materials Property Map, Initial RTflDT,FigurmIN -$.g (2) RV Materials Property Map, Adjusted RTflDT Corresponding to December 31, 1981, Figure 2 (3) Tabulation of Pressure Vessel Beltline Materials for Maine Yankee

Dear Mr. Garrity:

Analysis of the flaine Yankee reactor vessel materials has been completed as part of the C-E Owners Group effort on the II.K.2.13 report. Enclosed are vessel " maps" showing the initial and post-irradiation toughness properties at critical locations in the Maine Yankee reactor vessel. Figure 1 presents values of initial reference temperature (RTrlDT) for each of the plates and welds, flote that for the vertical weld seams (1-203, 2-203 and 3-203), a single value of RTNDT is given in each shell course. That value represents all three of the vertical seams in a given shell course. Figure 2 is a map of the adjusted RTNDT (initial reference temperature plus the transition temperature shift) at the inner surface of the Maine Yankee reactor vessel corresponding to December 31, 1981 (5.904 effective full ower years). The peak fluence at the vessel inside surface, 4.99 x 1018 n/cm, was adjusted for the axial and azimuthal variations in neutron flux to obtain a fluence for each location being calculated. Transition temperature shifts were predicted using regulatory guide 1.99, revision 1, using available data on the copper and phosphorus content of the plates and welds. Where chemistry data were not available, upper bound values (0.35i Cu, 0.012 w/o P) were used. Material properties and chemistry data used in the analysis are sumnarized in the attached table. Very truly yours, C/ r< w R. C. Jacques RCJ/smb Project Manager cc: C. D. Frizzle H. F. Jones J. B. Randazza E. C. Wood

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APPENDIX E PESULTS OF MAINE YANVEE PLANT SPECIFIC EVALUATION OF VESSEL INTEGRITY UNDER MSLB AND A00 CONDITIONS Enclosed is a report prepared by Combustion Engineering describing the behavior of the Maine Yankee reactor coolant system under certain main steam i line break (MSLB) and anticipated operation 1 occurence (A00) conditions. The report concludes that under the conditions evaluated, crack initation would not occur at any poir>t in plant lifetime. Thus, reactor vessel integrity is assured with respect to pressurized thermal shock within the bounds of the analyses performed. Maine Yankee considers this report to be preliminary and will probably wish to supplement it at a later time following completion of our review. i 3 E-1

i l { Enclosure A Thermal-Hydraulic Evaluation 1. Introduction i I This enclosure presents the results of a thermal-hydraulic evaluation of the main steam line break (SLB) accident and of overcooling anticipated operational l occurrences (A00s). performed for input to the pressurized thermal shock (PTS) stress analyses for the Maine Yankee plant In order to illustrate how the PTS concern arises, a qualitative discussion 1 i follows of a representative SLB transient. Assume that a large break occurs in l the main steam piping upstream of the main steam isolation valve (MSIV) associated with one steam generator, to be referred to as the "affected" steam generator. The break increases steam flow from the steam generators, steam i generator pressures and temperatures decrease, and heat removal from the reactor coolant system (RCS) increases. Low steam generator pressure causes both a reactor trip signal and a main steam isolation signal (MSIS). Reactor trip terminates, or prevents, fission power generation; MSIS terminates blowdown of the unaffected steam generators by closing the MSIVs ard terminates j feedwater flow to all steam generators. l A low steam generator water level signal' in the affected steam generator will 4 actuate auxiliary feedwater (AFW). Since the Maine Yankee AFW system includes 3 automatic AFW isolation based on steam generator pressure, AFW flow will not be i initiated to the affected (depressurized) steam generator. Following AFW 1 isolation, the affected steam generator will dry out and RCS cooldown will terminate. During the RCS cooldown transient, pressurizer pressure decreases to the safety injection actuation signal (SIAS) setpoint. SI AS starts two high pressure safety injection (HPSI) pumps and three charging pumps.M In addition, following i SIAS on low pressurizer pressure the opera r will trip all reactor coolant I pumps (RCPs). The HPSI and charging pumps _ ill repressurize the RCS, Conditions identified in the emergency procedures for termination of emergency core cooling flow will be reached and charging and HPSI pump flow will be reduced in order to terminate RCS repressurization. i The PTS concern arises due to the rapid decrease of reactor coolant temperature in the reactor vessel downcomer. The temperature decrease will be largest in j the section of the downcomer associated with the affected steam generator. PTS effects are increased by the repressurization of the RCS by the charging and HPSI pumps. ] 2. SLB Evaluation l Results of a thermal-hydraulic evaluation of the SLB accident are pre"ented in this section. ] In order to bound the PTS effects of SLB, the evaluation is performed for an i SLB occurring during hot zero power (HZP) operation. This mode of operation maximizes RCS cooldown because steam generator water inventory is largest at HZP and because core decay heat is lowest at HZP. ? bLtbte Yankee Note: Chatging ptunps.teatign to petform HPSI fwtetion. Upon SIAS 2 pwnps only opetate to teptessorize RCS. Cortec.t flow delivery has been sitiLized in analysis. _,,...m -ym,

b To further bound PTS effects, a guillotine rupture of the main steam line is i i postulated, with the assumption of no moisture carryover during the blowdown transient. The assumption of no moisture carryover maximizes total energy removal from the affected steam generator and, therefore, maximizes integral l RCS heat removal. The assumption of no moisture carryover during a guillotine rupture maximizes, in addition, the rate of RCS cooldown. [ s l A complete list of assumptions and plant parameters used for the SLB thermal-t' hydraulic evaluation is provided in Table 1. While not all of these assumptions and parameters have been chosen to maximize PTS effects, results are expected to provide an upper bounc' on the rate and magnitude of RCS cooldown which can occur during SLB, primarily due to the following combination j of assumptions: 1) HZP operating mode, 2) guillotine break with no moisture carryover, and 3) zero decay heat, [ i Results of the SLB thermal-hydraulic evaluation are provided in Figures 1 l and 2. Figure 1 provides the water temperature versus time in the section of l l the reactor vessel downcomer associated with the affected steam generator. ) Water temperatures in the affected steam generator loop are substantially lower i than in the unaffected steam generator loops. The downcomer water temperature i was obtained assuming complete mixing of the cold leg flow with HPSI and charging pump flow. i 3 Figure 2 provides the downcomer pressure versus time. A rapid repressurization to the PORV setpoint pressure (2400 psia) is shown. Operator action to i terminate charginq and HPSI pump flo# prior to reaching the PORV setpoint has I not been credited. These pumps are assumed to be shut off at 30 minutes. g i 3. Overcooling A00 Evaluation j j Results of a thermal-hydraulic evaluation of overcooling A00s are presented in this section. ^ The overcooling A00s consist of events which cause increased heat removal via one or more steam generators. Potential causes of increased heat removal include: a lI 1) Decrease in main feedwater enthalpy due to loss of a main feedwater

heater, F

4 2) Increase in main feedwater flow due to a main feedwater control valve i malfunction or due to a main feedwater control system malfunction, t 3) Increase in main steam flow due to a turbine control valve malfunction r or due to a turbine control system malfunction, j I 1 4) Increase in mair. steam flow due to a turbine bypass control valve l } malfunction or due to a turbine bypass control system malfunction, and j t 5) Increase in main steam flow due to an atmospheric dump valve malfunction. If uncontrolled RCS cooldown results from any of the above malfunctions, the cooldown will be accompanied by a decrease in steam generator pressure. When RCS cold leg temperatures decrease to about 440 F, MSIS will be caused by low SLthic Yankee Note: Cha,1ghtg pumps and HPSI pumps ate the same. Ope,tato.1 teould tetmbutte.teptessutization by stoppbig HPSI pumps. __.,__.-.-_m-

k* l steam generator pressure (400 psia). Following MSIS, MSIV closure and main i feedwater isolation will occur, thus terminating RCS cooldown for t(he first I four malfunctions listed above. Since tne atmospneric oump valves ADVs) are j upstream of the MSIVs, malfunction (5) can cause RCS cooldown to continue after MSIS. Consequently, a thermal-hydraulic evaluation of the atmospheric dump l valve malfunction is provided. } The Maine Yankee plant has a total ADV capacity of no more than 5 percent of j full power main steam flow rate at 900 psia steam pressure. The thermal-hydraulic evaluation assumes that all ADVs open, and remain open, resulting in i uncontrolled RCS cooldown for 90 minutes. In order to maximize RCS cooldown, the ADV malfunction is assumed to occur in the HZP operating mode with no decay 1 I heat. i Operator actions assumed for the A00 are the same as listeu in Table 1 for SLB, however, the time frame will differ due to the less rapid ecoldown for the A00 transient. The resulting downcomer water temperature and pressure transients are provided in Figure 3. I i ) l i 1 I 1 l f i l l i l i i

r Table 1. Assumptions and Plant Parameters Used for SLB Thermal-Hydraulic Evaluation Parameter Value Steam Flow Area 2 a) Affected Steam Generator 4.6 ft b) Unaffected Steam Generators 2 i) Before MSIV Closure 2.4 ft

11) After MSIV Closure 0.0 Blowdown Quality 1.0 Initial Power Level 0.0 Decay Heat 0.0 MSIS Setpoint 400 psia SIAS Setpoint 1600 psia HPSI Flow Shutoff Pressure above 2400 psia AFW Flow 0.0 Operator Actions SLB A00 a) Trip RCPs after SIAS on Low Pressurizer Pressure 30 sec 10 min b)

Terminate Charging and HPSI Pump Flow 1800 sec 90 min

( FIGURE 1 I SLB SCOPIf4G CALCULATION, LARGE BREAK AT ZER0 POWER, MAIN FEED ISOLATION AFTER MSIS, AUTOMATIC ISOLATION OF AUXILIARY FEED ON LOW STEAM GENERATOR PRESSURE MAINE YANKEE 600 i i i i 500 i ? 6 w 400 5 E t 5 e# 300 x 5 W i x8 200 d $s 5 C 100 = 0 i 0 500 1000 1500 TIf1E, SEC

r FIGURE 2 SLB SCOPING CALCULATION, LARGE BREAK AT ZERO POWER, MAIN FEED ISOLATION AFTER MSIS, AUTOMATIC ISOLATION OF AUXILIARY FEED ON LOW STEAM GENERATOR PRESSURE MAINE YANKEE 2500 i g 2000 5 2 1500.. u: 5$u o. 1000... htbic Vm hee Note: Asswnes opeutot fa.its to folecte procedu.tal bittiuctions te mabthtbi 50*F subcoolbig by thwttling HPSI fle:e. SOC _ I 0 0 500 1000 1500 TIME, SEC

l FIGURE 3 l l A00 SCOPIrlG CALCULAT;0ft ADV FAILED OPEf4 ( 5 % CAPACITY ) AT ZERO POWER t%If4E YAf4KEE 3000 5 N E 2000 y Ma.hte Yankee Hote: ASMuncs ope % tot falls g .to foi.loto ptocedural .insttuetiotts to w h mabttabt 50 F.subcoolbig by tJttet.tLbig HPSI fEcx. a-1000 0 t i i i 0 20 40 60 80 90 TIME, MIt4UTES 600 5 E* 400 g Eu =8 .a ' w$J 200 "E Sw i ES WW 0 0 20 40 60 80 90 TIME, MIf4UTES

1. Results of Fracture Mechanics Analysis for Maine Yankee Steam Line Break Transient The stress analysis and fracture mechanics analysis were performed using j the methods outlined in CEN-189. Plant specific material properties for the controlling longitudinal weld in the Maine Yankee vessel were used in the analysis as follows: PCT Ni .99 = PCT Cu' .35 = PCT P .012 = 0 Initial RT -50 F = NDT This corresponds to the intermediate shell axial weld.at an azimuthal angle of 270 degrees. The fluence factor at the mid-core level at this location is 86% of the peak fluence in the vessel. At the 12/31/81 level of 5.9 EFPY and peak fluence of.507 x 10 n/cm2 (E >l MeV), this corresponds 19 19 2 to a point fluence of.437 x 10 n/cm and an adjusted surface RT value NDT 0 of 168 F. The plot of K; vs time for this case is shown in Figure 1 for various assumed crack depths. These stress intensities result from the stresses due to the pressure and temperature transient given in Figures 1 & 2 of Enclosure A. The applied stress intensity values were used in determining the critical crack depth diagram as shown in Figure 2 for the Maine Yankee vessel at an additional 26 EFPY of operation. A small region is indicated where K; exceeds the arrest toughness, however, no crack initiation region is present. These results indicate that no crack initiation would occur throughout the Maine Yankee plant life for this particular Steam Line Break transient. 2. Results of Fracture Mechanics Analysis for Maine Yankee A00 Transient The pressure and temperature transient for this A00 case is presented in Figure 3 of Enclosure A. The resulting plot of K; vs time is given in Figure 3 for various assumed crack depths. The projected end-of-life critical crack depth diagram is given in Figure 4 which corresponds to an additional 26 EFPY of operation. The obvious lack of data in this figure indicates that no crack initiation would occur throughout the Maine Yankee plant life for this A00 transient.

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