ML19209A917

Response in Opposition to Jf Doherty 790724 Amends to Contentions 26,23 & 24.Amend to 26 Raises Untimely New Concern,Amend to 23 Provides No Specific Basis & Amend to 24 Is Based on Wrong Info.W/Certificate of Svc

ML19209A917
Person / Time
Site:	Allens Creek File:Houston Lighting and Power Company icon.png
Issue date:	08/10/1979
From:	Biddle C, Newman J BAKER & BOTTS, HOUSTON LIGHTING & POWER CO., LOWENSTEIN, NEWMAN, REIS, AXELRAD & TOLL
To:	Doherty J DOHERTY, J.F.
References
NUDOCS 7910050623
Download: ML19209A917 (32)

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Vt .- s- x , UNITED STATES OF AMERICA \ .'A NUCLEAR REGULATORY COMMISSION < - BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of 5 5 HOUSTON LIGHTING AND POWER COMPANY 5 Docket No. 50-466 5 (Allons Creek Nuclear Generating 5 Station, Unit 1) 5 APPLICANT'S RESPONSE TO " JOHN F. DOHERTY'S AMENDMENTS TO CONTENTIONS NUMBERED: 26, 23 & 24 Houston Lighting & Power Company (Applicant) files this response to the supplemental pleading filed in this proceeding by Mr. Doherty on July 24, 1979. Amended Contention No. 26 Although styled as an "anendment", Mr. Doherty has in fact raised a new contention concerning stud bolt integrity. The original contention 26 requested visual inspection of the stud bolts for reasons left unexplained. The " amendment" for the first time now raises a concern about the relationship of applied stress to yield ctrength for the reactor head stud bolts. As noted in a previous filing,$/ Applicant believes that the Board's June 25th order allowing amendments

 */    " Applicant's Response To TexPirg's Amended Contentions" dated July 10, 1979.

1104 344 C

is applicable only to the refinement of contentions filed on or before May 11, 1979. The Commission's regulations do not afford Mr. Doherty any further opportunity to file ldditional contentions without justification under the factors enumerated in 10 C.F.R. 5 2.714(a). He has not attempted such a justifica-tion and, consequently this untimely additional contention should be dismissed. Aside'from the fatal untimeliness of this pleading, Mr. Doherty has also failed to provide any " reasonably specific" basis for the contention. 10.C.F.R. 2.714(b). The pleading mentions, without benefit of explanation, a perceived difference between (1) the magnitude of applied stresses (including or limited to " tensile," "ATWS" and " strain energy") and (2) the ultimate yield strength. Apparently, Mr. Doherty believes that the second compares unfavorably to the first. Simply stated, then, Mr. Doherty maintains that the stud bolts are weak; what is missing is a " reasonably specific" statement of why he believes the stud bolts do not have the strength to sustain any and all credible design loads. Absent this explanation, the contention fails to meet the requirements of the Commission's regulations. 10 C.F.R. 2.714(b).

                                              >  I104 345 Applicant complies fully with the commission's regulation in 10 C.F.R. 550.55a which incorporates ASME Boiler and Pressure Vessel Code Section III pertaining to vessels and closure bolts which are a part of the reactor coolant pressure boundary. Mr. Doherty offers nothing to suggest that this is not true.      And, clearly, he does not show the special circumstances required to challenge a Commission's regulation. His untimely contention must be rejected for this additional reason.

Amended Contention No. 23 In this snendment, Mr. Doherty has attempted to detail a hypothetical pressure-surge induced LOCA. In so doing, he has again omitted any description of the initiating events and probable consequences. More importantly, however, Mr. Doherty still refuses to explain why the scenario he postulates--including the unidentified " power distribution shapes and peaki.rg factors" he is concerned with--is any different from the pipe break or open steam relief valve accidents fully analyzed and designed for. See "NRC Staff Response to John F. Doherty's Additional Contentions," at 10 (June 27, 1979). Without this minimal explanation there is no basis for the contention set forth with " reasonable specificity." 10 C.F.R. 2.714(b). According, the contention should be dismissed.

                                                   > 1104 346 Amended Contention No. 24 This amendment condenses to a single disputed fact: will the' peak energy of the fuel exceed the limit of 280 cal /gm if a droppped rod is worth 1.4% or more?    Mr.

Doherty alleges that it will. The. sole basis for this assertion are the " calculations in-the PSAR of the Montague Nuclear Plant." Attached to this response is a full summary of the results of the control rod drop accident analysis contained in the Montacue PSAR (Table 15.1.38-7). This table clearly shows, using Mr. Doherty's own evidence, that a rod worth in excess of 1.0% will result in a peak enthalpy well below the limit. The additional pages ceferenced by Mr. Doherty (also enclosed) show as well that the pivotal statements. which form the basis of this contention are wrong in every detail. This contention, as " amended" should be dismissed. Respectfully submitted, c . 7 6 u & A ,q ' J. Gregory Copeland OF COUNSEL: C. Thomas Biddle, Jr. Charles G. Thrash, Jr. BAKER & BOTTS 3000 One Shell Plaza 3000 One Shell Plaza Houston, Texas 77002 Houston, Texas 77002 Jack R. Newman LOWENSTEIN, NEWMAN, REIS, Robert H. Culp AXELRAD & TOLL 1025 Connecticut Ave., N.W. 1025 Connecticut Ave., N.W. Washington, D.C. 20036 Washington, D.C. 20036 ATTORNEYS FOR APPLICANT HOUSTON LIGHTING & POWER COMPANY TB:02:H 4 1104 347

UNITED STATES OF AME3ICA NUCLEAR REGULATORY COE4ISSION BEFORE THE ATOMIC SAFETY AND 'f CENSING BOARD In the Matter of S S HOUSTON LIGHTING & POWER COMPnNY S Docket No. 50466 S (Allens Creek Nuclear Generating S Station, Unit 1) S CERTIFICATE OF SERVICE I hereby certify that copies of the foregoing Applicant's Response to " John F. Doherty's Amendments to Contentions Numbered: 26, 23 & 24 in the above-captioned proceeding were served on the following by deposit in the United States mail, postage prepaid, or by hand delivery this Af6. day of A iA f- , 1979.

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Sheldon J. Wolfe, Esq., Chairman Richard Lowerre, Esq. Atomic Safety and Licensing Assistant Attorney General Board Panel for the State of Texas U.S. Nuclear Regulatory Commission P. O. Box 12548 Washington, D. C. 20555 Capitol Station Austin, Texas 78711 Dr. E. Leonard Cheatum Route 3, Box 350A Hon. Charles J. Dusek Watkinsville, Georgia 30677 Mayor, City of Wallis P. O. Box 312 Mr. Gustave A. Linenberger Wallis, Texas 77485 Atomic Safety and Licensing Board Panel Hon. Leroy H. Grebe U.S. Nuclear Regulatory Commission County Judge, Austin Ccsnty Washington, D. C. 20555 P. O. Box 99 Bellville, Texas 77418 Chase R. Stephens Docketing and Service Section Atomic Safety and Licensing Office of the Secretary of the Appeal Board Commission U.S. Nuclear Regulatory U.S. Nuclear Regulatory Commission Commission Washington , D. C. 20555 Washington, D. C. 20555 R. Gordon Gooch, Esq. Atomic Safety and Licensing Baker & Botts Board Panel 1701 Pennsylvania Avenue, N. W. U.S. Nuclear Regulatory Washington, D. C. 20006 Commission Washington, D. C. 20555 1104 348

Steve Schinki, Esq. Staff Counsel U. S. Nuclear Regulatory Commission Washington, D. C. 20555 John F. Doherty 4438 1/2 Leeland Houston, Texas 77023 Madeline Bass Framson 4822 Waynesboro Drive Houston, Texas 77035 Robert S. Framson 4822 Waynesboro Drive Houston, Texas 77035 Carro Hinderstein 8739 Link Terrace Houston, Texas 77025 - D. Marrack 420 Mulberry Lane Bellaire, Texas 77401 Crenda McCorkle 6110 Darnell Houston, Texas 77074 F. H. Potthoff, III 1814 Pine Village Houston, Texas 77080 Wayne E. Rentfro P. O. Box 1335 Rosenberg, Texas 77471 James M. Scott, Jr. 8302 Albacore Houston, Texas 77074 (- Z[ t C. Thomas Biddle, Jr. /' 1104 349

Table 15.1.38-7 s RESULTS OF DESIGN BASIS CONTROL ROD DROP ACCIDENT Number o l' Core Rod Core Clobal Power Peak Failed Rod Exposure Drop Scram Average Peak Peaking Enthalpy Fuci Worth MWD /T Velocity Time Enthalpy Enthalpy Factor (cal /gm) Rods

     .01435 Beginning 5 ft/sec             A      28.7608    202.9354 1.102 221.9564 444 of          5 ft/see        B     28.5543    200               218.772 Life         5 ft/see        C     28.1874    196               214.314                -

5 ft/sec D 27.1343 184 201.090 2.79 ft/see A 27.2850 186 203.294 2.79 ft/sec

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B 27.0308 183 199.998 2.79 ft/sec C 26.6027 178 194.478 2.79 ft/sec D 25.4293 163 177.948

     .01142 3500            5 ft/sec        A     25.359     153.0929  1.087' 164.981            0 5 ft/see         B    25.304     152               163.793 5 ft/sec         C    25.018     148               159.445
's                         .5 ft/see         D    24.324     140.5             151.292 2.79 ft/see      A    24.262     140               150.749 2.79 ft/see      B    24.147     139               149.662 2.79 ft/see      C    23.817     134.5             144.770 2.79 ft/sec      D    23.055     125               134.444
     .01130 End of          5 ft/see         A    29.7848    182.7158 1.079    195.851      260 Life      i   5 ft/see         a    29.434     178               190.762 b

5 ft/see C 29.242 176 188.604 5 ft/sec D 28.151 166 177.814 2.79 ft/sec A 28.635 170.5 182.670 2.79 ft/sec B 28.193 166.3 178.138 2.79 ft/sec C 27.968 164 175.656 2.79 ft/sec D 26.755 152 162,708 A = 5 sec. tech spec. B = 4 sec. tech spec. Refe_rence 1(b) C = 4 sec. aver. age 1104 350 ~.,

   } D = =easured average          ,

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lo 3-4 10 10 to DISTANCE (M ETER S) FIG.15.1. 3 8 -l CONTROL ROD DROP ACCIDENT NRC-DRL ASSUMPTIONS INH AL ATICN(THYROID) ('j DOSE VERSUS DISTANCE MONTAGUE NUCLE AR POWER STATICN UNITS I AND 2 PRELIMIN ARY SAFETY AN A LYSIS REPORT SU P P t. E M F. N T 7 12/12/75

1- P00F0MNil to the actual performance of operating reactors. Specific comparisons have been made for the Oyster Creek and Dresden 2 plants. The results of these comparisons show the calculated and actual results agree within experimental and manufacturing tolerances. The design methods have been shown to be able to compute local powers to within 2 3% fuel assembly segment powers to within

     +10%. Pu/U ratios vs exposure to within 23%, and core reactivities and cold shutdenn margin to within 0.5% .aK.

Experimental tests have also been used to verify the analytical calcu-lations of both reactivity and isotopic composition. These tests give resulta nearly identical to the comparisons with the operating plants. Reference 25 contains a complete discussion of errors, uncertainties, and ,o calculations / data comparisons pertaining to the analytical methods used in the

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design of BWR cores. 4.3.3.2 Nuclear Evaluations 4.3.3.2.1 Ceneral The analyses presented in Subsection 4.3.2 show th__ the safety design . basis is satisfied in conjunction with the nuclear design requirements of Subsection 4.3.1. Adequate protection is provided for the cladding and nuclear system process barrier. The nuclear requirements for reactivity control system.: 2 and the jettings of the reactor protection system are primarily associated with . limitations on the levels and rates of change of reactivity, power, and tempere-ture. Normal plant operation is conducted at rates and values of these para-meters, such that reactor transients are readily observable and controllable by plant personnel. 4.3.3.2.2 Reactor Protection The reactor. protection system responds to some abnormal operational tran-siento by initiating a scram. The reactor protection system and the CRD system act quickly enough to prevent the initiating disturbance from causing fuel damage. The scram reactivity curves used in the reactivity excursion analysen at hot startup at various exposures during the first fuel cycle are included as Figures 4.3-4, 4.3-Sa and 4.3-Sb. The scram reactivity for termination of these abnormal 6peratfonaT eransfents is shown in Figure 15.1.1-1. Abnormal operational transients are evaluated in Chapter 15. No fuel damage results from any abnormal operational transient. The full scram power decay curve is presented in Figure 4.3-12. This curve displays the relative core heat flux, including decay heat contribution, as a function of time after rod motion begins. The following conditions apply: Supplement 2 11/8/74

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083074

Full pc,us'c scram End of cy.cle conditions Constant pressure Constant flow 4.3.3.2.3 Control Rod Movement and Patterns The specified rod withdrawal sequences and the Rod Pattern Control System maintain rod worth at acceptably low valt. 1 to minimize the consequence of a rod reactivity accident. At any specified reactor state, peak enthalpies for rod removal accidents vary proporcionately with rod worths. Peak enthalpy provf n.:s the best index for determining the consequences of a reactivity accident when correlated to experimental measure =ents. Analyses and a survey of pertinent experl= ental data (Reference 15) indicate that prompt dispersal of finely frag =ented fuel into the coolant with subsequent 1.rge pressure rise rates does not occur at excursion energy densities below 425 cal /g. Energy densities above thin icvel can cause pressure surges that may endanger the reactor coolant prenmure boundary. To provide a margin below the 425 cal /g, a design limit on peak fuel enthalpy of 280 cal /g is selected. This fuel enthalpy limit is supported by -w a careful study of all available SPERT, TREAT, KIWI, and PULSTAR tests (Ref- i V erence 16). In addition, more than a thousand transient tests have been perfor=ed at the Caps :le Drive Core (CDC) facility at SPERT during the past several years. A large majority of these tests were in the enthalpy range of 280 cal /g. To enu=erate all the tests is not practical. However, typical tests are discussed in Reference 17. This is a report on fuel pin transient tests that were con-ducted in the CDC facility. The nuclear-to-mechanical energy conversion ratio was found to be essentially zero even for tests that resulted in peak fuel enthalpies as high as 338 cal /g. The enthalpy of Uo2 , as a function of fuel temperature, was obtained from experirsental data taken by Hein and Flagella (Reference 18). These data indicated the ft el melting range to be 270 to 337 cal /g. In addition, more recent data support the UO 2 celting range-t be- 2 70 -to 337-cal /g-(References-19 20, and 21) . ~~ ~ Hence, no phase change in UO 2 is expected entil enthalpies in the range of 270 to 337 cal /g are achieved.

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There are no experimental data to date that indicate a possibility of pro =pt fuel failure in the fuel enthalpy range discussed. Therefore, the peak fuel enthalpy and design li=1t of 280 cal /g is considered justifiable and conservative. 4.3-28 Supplement 2 11/8/74 082374 1104 353

9 Specified control rod withdrawal sequences are designed to limit rod worth o that the drop of any control rod from the fully inserted position to the ponition of its drive results in a peak fuel enthalpy of not more than 280 cal /g. A velocity limiter limits the average measured rod velocity plus 3 standard devia-tions to less than 2.79 ft./sec. Control rod removal excursion analysis from the shutdown flux level (Reference 4) for a typical BWR using axial gadolinia indi-cates that peak fuel enthalpies of 280 cal /g result from rod worths of 0.0145 a k (cold, critical) or 0.0145 Ak (hot, critical, zero voids) and removal rates of 2.33 and 2.79 ft/see respectively. These analyses also show that for excursions initiated from flux levels corresponding to 20lll power, the maximum possible con-trol rc,d worth, 0.020 ak, is insufficient to cause peak enthalpies of 280 cal /g. Preplanned rod patternr enforced to restrict incremental rod worth to approxi=ately 0.01 Ak, although larger values are acceptable within the 280 cal /g limit. Therefore, no rod would have a worth high enough to produce peak enthalpy of 280 cal /g even if the rod were removed at 2.79 ft/sec. , 4.3.3.2.4 Xenon Tennnients. The maximum xenon reactivity buildup on shutdown . from full power and the rate of xenon reactivity burnout on return to full power when the maximum shutdown :.enon buildup occurs, are calculated for both , te beginning-of-life and the end-of-cycle reactor conditions. The maximum rate of reactivity change is obtained by assuming an instantaneous return to full power. The results of these calculations are shown in Figure 4.3-13 for the beginning-of-life condition. From this analysis it was determined that the maximum reactivity addition caused by burnup of xenon was +0.00010 (ak/k)/ minute. Assuming a control rod worth of 0.001 ok/k with an insertion rate of 3 in./sec, the reactivity addition by the control rod insertion is -0.00125 (A/kA)/ minute. Thereforb, a very weak control rod can easily compensate for a xenon-burnup reactivity addition. The standby liquid control system, used for emergency shutdown only, is more than adequate to compensate for the reactivity added by xenon decay. With a boron injection rate of 6 ppm / minute, the reactivity insertion of the liquid control system is -0.0013 (ak/k)/ minute. The design injection rate of 6 to 25 ppm / minute. Boiling water reactors do not -have instability problems due tcr xenon. --- - - -- -- This has been demonstrated by operating BWR's for which xenon instabilities have never been observed (such instabilities would readily be detected by the l IPRM's), by special tests which have been conducted on operating BWR's in an m. 4.3-29 Ii04 354 " 573

attempt to force the reactor into xenon instability, and by calculations. All C of these indicators have proven that xenon transients are highly damped in a BWR Ge to the large negative power coefficient. The most recett analysis and experi=ents conducted in this area are reported in Reference 22. Yields of I-135 and Xe-135 for the various fissionable isotopes are pro-vided in Table 4.3.12. { 4.3.3.2.5 Variation of Nuclear Parameters. The BWR nuclear fuel design incor-porates sufficient. conservatism to allow for minor variations in the nuclear parameters such as excess reactivity, reactivity coefficients and reactivity insertion rates. These parameters are individually analyzed for each fuel design, with small reliance being placed on previous designs or prior practice. Excessive reactivity, for example, is a matter of primary. design effort. Extensive analyses are perfor=ed to establish suitable amounts and locations of the burnable poison (Gd 230 ) such that the specified reactivity margins are obtained. These margins are bounded by shutdown limits on one end, and performance considerations on the other. The reactivity coefficients, as already described, provide the negative feedback necessary for normal and accident reactor control. Small variations in the magnitude of these coefficients are not significant due to realtive abundance of negative reactivity feedback in the BWR. Large unexpected variations are precluded by the extensive design evaluation perfor=ed for each fuel design. Reasonable variations in the controlled reactivity insertion rates have very small effect in he t BWR. Control rods are operated one at a time resulting in very low reactivity addition rates which are well below the limiting criteria. Considerable variations could be tolerated without major concern. Finally, the in-core instrumentation system provides an important safeguard against the effects of unexpected or unusual nuclear parameters. Together wi:h the process computer, the instru=entation system provides prompt and reliable data which can be' used to identify, analyze and formulate appropriate action as needed and mitigate the effects of undesirable variations in the nuclear param-eters. For a further discussion of dn-core instrumentation, see Subsection 4.3.6. 4.3.3.2.6 Scram Function Curves. Both the total scram reactivity worth and shape function are strongly dependent on the fuel design and loading pattern (e.g., reload fuel, axial gadolinium distribution, etc.), and for this reason r 1104 355 4.3-30 121473

15.1.38 Control Rod Drop Accidenc ,, 15.1.38.1 Identification of Causes. There a're many ways of inserting reactivity into a boiling water reactor. However, most of them result in a relatively slow rate of reactivity insertion and therefore pose no threat to the system. It is possible, however, that a rapid removal of a high worth control rod could result in a potentially significant excursion. Therefore, the accident which has been chosen to encompass the consequences of a reactivity excursion is the control rod drop accident. 15.1.38.2 Starting Conditions and Assumotions. Before the control rod drop accident is possible, the following sequence of events must occur (1) The complete rupture, breakage or disconnection of a fully inserted control rod drive from its cruciform control blade at or near the coupling. (2) The sticking of the blade in the fully inserted position as the rod drive is withdrawn. (3) The falling of the control rod to the red drive position. This unlikely set of circumstances makes possible the rapid-removal of a control rod. The dropping of the rod results in a high local k. in a small region of the core. For large, loosely couples cores, this would result in a highly peaked power distribution and subsequent shutdown mechanisms. Significant shifts " ~ in the spacial power generation would occur during the course of the excursion. Therefore, the method of analysis must be capable of accounting for any possible effects of the power distribution shifts. The Rod Pattern Control System limits the worth of the rod which could be dropped. This system prevents the movement of an out of sequence rod or rod gang in the 100% to 50% rod density range and from the 50% rod density point to the preset power level, the RPCS will only allow group notch mode or gang rod with-drawal or insertion. The 50" rod density configuration corresponds to the con-d! tion in which 50% of the rods are fully inserted in the core and 50% are fully withdrawn. With the stipulation that no out of sequence rod may nr. moved prior to 50% rod density, the postulated rod drop accident cannet result in peak enthalpics in excess of 280 cal /gm for any possible plant operation or coro exposure. --- -- --__-- - - - - - - - - - - - - - - _ _ _ _ _ _ - - - - O 1104 356 15.1-153 083074 Supplement 3 12/12/74

The envelope of the maximum worth control rods with the RPCS operational is given by Figures 4.3-2a and 4.3-2b. The analysis performed for the control rod drop accident assumes that the maximum worth control rod which exists at the 50% control rod density patcern has its drive fully withdrawn and drops from the core. It shall be emphasized that the RPCS would prevent this from occurring; however, this is a convenient analytical procedure for establishing an upper branch on the rod worth which would result in a peak fuel enthalpy which approaches the design limit of 280 cal /gm. Using this approach, it is demonstrated that a rod drop accident involving an in-sequence rod enforced by the RPCS will result in in peak enthalpies less than 280 cal /gm. The reactivity function for the design basis control rod drop accident at various first cycle exposures are given in Figures 15.1.38-3 through 15.1.38-5. The corresponding scram reactivities used for these analyses are given in Figures 4.3-4, 4.3-5a, and 4.3-5b. 15.1.38.3 Accident Description. The accident is defined as: (1) The RPCS is functioning. . (2) The highest worth rod that can be developed at any time in core life , under any operating conditions drops from fully inserted position to the control rod drive position. (3) The rod drops. (4) The scram is that defined in the technical specifications. The detailed analysis of this accident is discussed in Reference 1, la and lb. A continuing effort is being =ade in the area of analytical =echods to assure that nuclear excursion calculations eflect the latest " state-of-the art." The soquence f events and the approxi= ate ti=es of occurrence are as follows: Approximate Event Elapsed Time (1) Reactor is operating at 50% control rod density pattern. (2) Maximum worth control blade beco=es decoupled. (3) Operator selects and withdraws the control rod d_ rive of the decoupled =ax1=u= worth rod along_..ith..the_other_ con _ __ ____ _ . _ _ _ __ _ trol rods assigned to that RPCS notch group or gang to the fully vichdrawn position. (4) Blade sticks in the fully inserted position. 1104 357 O 15.1-154 083074 Supple =ent 3 12/12/74

Approximate , Event Elapsed Time (5) Blade becomes unstuck and drops at the nominal measured velocity plus 3 standard deviations. 0 (6) Reactor goes prompt critical and initial power burst is terminated by the Doppler Reactivity Feedback. <1 sec (7) APRM 120% power signal scrams reactor. (8) Scram terminates accident. <5 see 15.1.38.4 Identification of Operator Actions. The termination of this excur-sion is accomplished by automatic safety features or inherent shutdown mechanisms. Therefore, no operator action during the excursion is required. 15.1.38.5 Analysis of Effects and Consecuences 15.1.38.5.1 Realistic Evaluation Methods. The analytical methods and associated assumptions which are used in evaluating this accident are considered to pro-vide a realistic, yet conservative assessment of the consequences. 15.1.38.5.1.1 Methods, Assumotions and Conditions. The methods, assu=ptions, and conditions for evaluating the excursion aspects of the control rod drop accident are described in detail in References 1, la and lb. 15.1.38.5.1.2 Results and Consequences 15.1.38.5.1.2.1 Fuel Damage. The fuel damage thresholds are based on both { experimental and theoretical data. This information was discussed previously in Section 4.2 of this document, with additional detailed .information presented in Section 5 of Reference 2 The results of the rod drop accident analysis are presented in Table 15.1.28.8. The peak enthalpy results of the design basis control rod drop accident (Ref-erence Ib) are less than the 280 cal /gm design limit for all exposures. The number of failed' fuel rods due to the design basis control rod drop accident is less than 770 rods for all plant operating conditions and exposures. However, the radiological exposure calculations have been performed on the assumed failure of 770 fuel rods. 15.1.38.5.1.2.2 Fission Product Release f rom Fuel. The following ersumptions are used in calculating fission product activity release from the fuel: (1) The reactor has been operating at 3758 M*n't until 30 min prior to the accident. When translated into actual plant operation, this assumption means that the reactor war. shut down from-design power, taken critical, and brought to . the initial temperature condit ions within 30 min of the departure f rom design power. The 30-min time represents a conservative estimate of the shortest period in which the required plant changes could be accomplished and defines the decay time to be applied to the fission product inventory calculations. 1104 358 45.,-,55 092774 Supplement 1 2/21/75

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POOROHINL (2) An average of 1.8*. of the noble gas activity and 0.32~. of the halogen activity in a 'perf urated fuel rod is assumed to be released. These percentages are consistent with actual measurements made during defective fuel experiments (Reference 3). (3) For this plant, the following fission product activities are contained in the core, at the time the accident occurs: Noble gases 4.8 x 108 Ci Iodine 8.7 x 100 C1 (4) The fraction of solid fission product activity available for release from the fuel is negligible. (5) The fission products produced during the nuclear excursion are neglected. The excursion is of such short duration that the fission products generated are not;11g ib le to compartnen with the fission products already preuent in the fuel. Using the above assumptions, the following amounts of fission product activity are released from the failed fuel rods to the reactor coolant: - Noble gases (Xe, Kr} 5 1.3 x 10 Ci Iodine 4.6 x 10' Ci C 1104 359 C 15.1-156 083074 Supplement 3 12/12/71

MONTAGUE 152 PSAR O -

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O 15.1.38.5.1.2.3 Condenser Activity The following assumptions are used in calculating the amount of fission product activity transported frcxn the reactor vessel to the main condenser:

a. The recirculation flow rate is 25 percent of rated, and the steam flow to the condenser is 5 percent of rated.

The 25 percent recirculation flow and 5 percent steam flow are the maximum flow rates compatible with the maximum fuel damage. The 5 percent steam flow rate is greater than that which would be in effect at the reactor power level assumed in the initial conditions for the accident. This assumption is conservative

                  - because - it -results     in--the'- transport of more fission products through the steam lines than would be expected.

Because of the relatively long fuel-to-coolant h"at transf er time constant, steam flow is not significantly affected by the increased. core heat generation within Q 1104 560 15.1-156a Supplement 3 12/12/74

MONTAGUE 1 & 2 PSAR C THIS PAGE INTENTIONALLY LEFT BLANK C

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MONTAGUE 162 PSAR the time required for the main steam line isolation valves to achieve full closure.

b. The mala steam line isolation valves are assumed to receive an automatic closure signal 0.5 sec after detection of high radiation in the main steam lines and to be fully closed at 5 sec from the receipt of the closure signal. The automatic closure signal oriqinates

     ,                 f rom the main steam line radiation monitors. The total time    required to isolate the main steam lines (5.5 sec) combined with the assumptions in (1) , dictates the total amount    of fission product activity transported to the condenser before the steam lines are isolated.

c. All of the noble gas activity is assumed to be released to the steam space of the reactor vessel.

d. The mass ratio of the halogen concentration in steam to that of the water is assumed to be 2 percent.

e. Fission product plate-out is neglected in the reactor vessel, main steam lines, and condenser.

Of those fission products released from the fuel and transferred to the condenser, it is assumed tha't 100 percent of the noble

gases are airborne in the condenser. The iodine activity airborne in the condenser is a function of the partition factor.

The partition factor assumed applicable is 100. Based on the above condition , the activity airborne in the condenser is presented in Table 15.1.38-1. 15.1.38.5.1.2.4 Activity Released to Environment

                                                         ~

The fission product activity released to the environment is a function of the total amount of activity airborne in the condenser and the condenser leak rate. For the purpose of this analysis it is assumed that:

a. 100 percent of the noble gas activity transferred to the condenser is airborne and available for release to the environment.

b. The iodine activity airborne is in proportion to the wa terborne act s.vity , the partition factor, and the volumes -o f air and wa te r . -- -- - - --- - - - - - - - - - -

c. The condenser leak rate is 0.5 percent of the combined condenser and turbine-free volume per day.

d. The activity released from the condenser becomes airborne in the turbine building and is released to the environment at a rate of 8.4 air changes per day.

15.1-157

                            ""===                P00RORL81NAL Based on tne above assumptions, the fission product release rate to the environment is presented in Table 15.1.38-2.

( 15.1.38.5.1.2.5 Radiolooical Effects Based on the release rates presented in Table 15.1.3 8 -2, the resultant radiological exposures are presented in Table 15.1.38-

3. It should be noted that all of the exposures are orders of magnitude below the guidelines set forth in 10CFR100.

15.1.38.5.1.3 consideration of Uncertainties Consideration of uncertainties with regards to the core physics calculations have been reported previously in Ref. 2. In addition, Ref. 1 presents a sensitivity analysis of the rod drop accident with regards to rod drop velocities, scram-insertion rates, and control rod worth for a wide spectrum of operating conditions. This approach has been taken to demonstrate the ' comparison between a realistic and a worst case condition. 15.1.38.5.2 Conservative (AEC) Licensing Basis Evaluati'on Methods 15.1.38.5.2.1 Methods, Assumotions and Conditions While the AEC has not published an official guide for the control rod drop accident, the assumptions, methods, and conditions used in this report are typical of those used by the AEC in past licensing efforts. 15.1.38.5.2.2 Results and Consequences 15.1.38.5.2.2.1 Fuel Damage As noted in the previous sections, the exact extent of fuel damage has not begn established for this accident. However, for the purpose of providing a relative dose effect, it is assumed that 770 fuel rods experience cladding damage. 15.1.38.5.2.2.2 Fission Product Release from Puel It is assumed that 50 percent of the halogens and 100 percent of the noble gases contained in those rods which experience cladding damage are released from the fuel. Those rods which experience cl.ulding damage are annumed_to have -a ---pea king - f a ctor- - o r -- 1. 5. - -- Theretore, the activity released from these rods is 75 percent and 150 percent, respectively, of the halogen and noble gas activity contained in the average fuel rod. Of those fission products released from- the fuel, 90 percent of the halogens and 0 percent of the noble gases are absorbed by the reactor water. The remaining activity is released to the condenser prior to isolation valve closure. 15.1-158 1105 002

MONTAGUE 162

                                               '"""           i 3

u n 3RN!g 15.1.38.5.2.2.3 Condenser Activity Based on the failure and transport mechanisms defined above, and assuming a plate-out factor of 2 in the condenser for iodines , the activity airborne in the condenser is presented in Table 15 .1.38-4 . 15.1.38.5.2.2.4 Activity Released to Environment The fission product activity released to the environment is dependent upon the activity airborne in the condenser, the condenser leak rate, and the turbine building leak rate. For the purpose of this analysis it is assumed that the condenser leak rate is 0.5 percent per day and the turbine building leak rate is in finite . Based on the airborne activity presented in the previous subsections and the above leakage rates, the noble gas and iodine release rates to the environment are presented in Table 15.1.38-5. 15.1.38.5.2.2.5 Radiolocical Etf ects '

a. Offsite. On-site meteorology for a ground level .

release, is assumed f or this event.

• Consideration of the fission product release rates in Table 15.1.38-5 and the above meteorology results in the radiological

-O exposures presented in Figures 15.1.38-1 and 15.1.38-2.

It should be noted that these exposures are well below the guidelines set forth in 10CFR100.

b. Control Room. Based on the control room assumptionn listed in Section 15.1.39, the whole body dose in the control room due to this accident is 26 mrem.

15.1.38.5.3 Comoarison of Realistic and Conservative Parameters As mentioned previously, the basis for the conservative calculation for this accident is past AEC practice in licensing BWRs. The comparison of realistic and conservative parameters is made in Table 15.1.38-6. 15 .1.38.6 References

1. R.C. Stirn et al., " Rod Drop Accident Analysis for Large Boiling Water _ Reactors," NEDO-10527, . March.1972. - - --

la. NEDO-10527, Supplement No. 1, Aug 1972. 1b. NEDO-10527, Supplement No. 2, March 1973,

2. J.E. Boyden et al. , " Summary Memorandum of Fxcursion Analysis i Uncertainties," Dresden Nuclear Power Station Unit 3 Amend- 3 ment No. 3.

15.1-159 1\05 DD3 Supplement 6 7/3/75

y l , PSAR

3. N. R. -Horton, W. A. Williams, J. W. Holt = claw, " Analytical C

Methods for Dialuating the Radiological Aspects of General 5 Electric Boiling Water Reactors," APED-5756, March 1969. m C b 1105 W C 15.1-160 Supplement 5 5/2/75

O O O- - MONTAGUE 162 PSAR . a TABLE 15.1.38-1 t t CONTPOL ROD DROP ACCIDENT REACTIVITY AIRBORNE IN TIIE CONDENSER , f (REALISTIC ANALYSIS) i r Airborne Activity [Ci) vs Decay Time i i Isotope 1 min 1 hr 2 hr 8 hr 1 day 4 days 30 days I-131 2.540 2.S*0 2.S+0 2.4+0 2.3+0 1.7+0 1.6-1 I-132 3.0-1 3.8-1 3.7-1 3.5-1 3.1-1 1.6-1 S.4-4 I-133 1.3+0 1.340 1.2+0 1.0*0 S.9-1 S.4-2 0 I-134 3.0-1 1.4-1 6.2-2 S.3-4 1.7-9 0 0 6.6-1 S.9-1 3.2-1 6.1-2 3.5-5 0 1-135 7.3-1 Tbtal Iodine S.2+0 S.040 4.7+0 4.1+0 ' . 3+0 1.940 1.6-1 Kr 83m S.1+1 3.S+1 2.54 1 2.840 8.2-3 0 0 Kr l 8Sm 2.7+2 2.3+2 2.0+2 7.6+1 6.1+0 7.1-5 0 ~~ Kr485 4.7+2 4.7+2 4.7+2 4.747 4.742 4.6+2 4.0+2

      $     Krj87       2.0+2     1.242      6.94*         2.840     S.6-4         0            0 Kr-88       4.S+2     3.S*2      2. 7 + :      6.2+1     1.2+0     2.1-8            0 m

Kr-89 1.2-1 3.4-7 7.7- 0 0 0 0 ca 8 CD $ S.S+0 LJT " Xe 131m 3.641 3.6*1 3.6+1 3.S+1 3.4+1 2.8+1 f Xe-133m 1.S+2 1.S*2 1.S+2 1.442 1.1+2 4.441 1.5-2 , Xe-133 7.4+3 7.4+3 7.3+3 7.1t3 6.543 4.3+3 1.3+2 Xe-135m 1.8+1 1.S*0 2.8-1 9.9-2 1.9-2 1.1-5 0 Xe-135 1.8*3 1.7+3 1.5-3 9.8+2 2.9+2 1.2+0 0 Xe 137 4.4-1 1.4-5 3.8-10 0 0 0 0 Xe-138 6.S+1 S.9+0 S.1-1 2.1-7 0 0 0 , Total NG 1.1+4 1.0+4 1.0+4 8.9+3 7.1+3 4.8+3 S.4+2 , r;ote: 2.Se0 = 2.Sx100 i

MONTAGUE IS2 PSAR TABLE 15.1.38-2 CONTROL ROD DROP ACCIDENT FISSION PRODUCT RELEASE RATE TO ENVIRONMEtTI' (REALISTIC ANALYSIS) , Release Rate (Ci/sec) vs Time 30 days Isotope! 1 min 1 hr 2 hr 8 hr 1 day 4 days I-131 8.4-10 4.3-8 7.2-8 1.3-7 1.3-7 1.0-7 9.4-9 l 9.2-9 3.1-11 I-132 1.3-10 6.4-9 1.1-0 1.9-8 1.8-8 I-133 4.4-13 2.1-8 3.5-8 S.4-8 3,4-8 3.1-9 0 I-134 1.0-10 2.3-9 1.8-9 2.9-11 0 0 0 2.5-10 1.1-8 1.7-8 1.7-8 3.5-9 2.0-12 0 I-135 i Tbtal , Iodine . 1.8-9 8.4-8 1.4-7 2.2-7 1.9-7 1.1-7 9 . 4 -9 i Kr-83m 1 1.7-8 6.1-7 7.2-7 1.5-7 4.8-10 0 0 Kr-85m ' 9.1-6 3.9-6 S.7-6 4.2-6 3.5-7 4.1-12 0 Kr-85 l 1.6-7 8.0-6 1.4-5 2.6-5 2.7-5 2.7-5 2.3-5 6.7-R 2.0-6 2.0-6 1.5-7 3.2-11 0 0 [ Kr-87 I Kr-88 I 1.5 ^i 6.0-6 8.0-6 3.4-6 6.8-8 1.2-15 0 0 0 0 0 { m Kr-89 l I 4.1-11 S.8-15 0 1.2-8 6.1-7 1.0-6 1.9-6 2.0-6 1.6-6 3.2-7 to Xe-131mi 8.8-10 S.1-8 2.5-6 4.3-6 7.4-6 6.4-6 2.5-6 Xe-133m l' 2.5-6 Xe-133 1.3-4 2.1-4 3.9-4 3.8-4 2.5-4 7.5-6 Xe- 13 5m ,' 6.1-9 2.6-8 8.0-9 S.4-9 1.1-9 6.4-13 0 Xe-135 , 6.1-7 2.9-S 4.5-S 5.3-5 , 1.7-5 6.7-9 0 Xe-137 ! 1.5-10 2.4-13 0 0 0 0 0 [[ 1.2-14 0 0 0 2.2-8 1.0-7 1.5-8 c3 Xe-138 l tr, Total NG 3.7-6 1.8-4 2.9-4 4.9-4 4.3-4 2.8-4 3.1-5 CD , CD . CN Note: 8.4-10 = 8.4x10-80 . I

                                                             +

MOtr1* AGUE 152 PSAR On - TABLE 15.1.38-3 cot 4TPOL FOD DPOP ACCIDQrr r (REALISTIC ANALYSIS) Radiological Effect Dose frem) Location Dose Duration Thyroid Beta GanT ta Exclusion Area 2 hr 6.6x10-* 3.4x10-* 3.2x*0-* (815 m) 6 Low Population Zone 24 hr 5.4x10-* 1.6x10-5 9.6x10-* (4,023 m) G

                                                                                                           \\ob 001 15.1-163                                  Supplement 6 7/3/75

MONTAGUE 152 PSAR TABLE 15.1.38-4 COhrrROL ROD DROP ACCIDENT ACTTVTTY AIRBORNE IN THE CONDENSER CONSERVATIVE (NHC) ANALYSIS Airborne Activity (Ci1 vg Decay Time Isotope 1 min 1 hr 2 hr 8 hr 1 day I-131 6.9+4 6.9+4 6.9+4 6.7+4 6.3+4 I-132 1.0+5 7.4+4 5.5+4 8.9+3 6.9+1 I-133 1.1+5 1.1+5 1.0+5 8.5+4 5.0+4 I-134 1.3+5 5.7+4 2.6+4 2.2+2 7.0 -4 I-135 1.1+5 9.7+4 8.7+4 4.7+4 9.0+3 Total Iodine 5.2+5 4.1+5 3.4+5 2.1+5 1.2+5 Kr-83m 2.7+5 1.9+5 1.3+5 1.5+4 4.4+1 Kr-85m 5.5+5 4.7+5 4.1+5 1.6+5 1.3+4 Kr-85 2.8+4 2.8+4 2.8+4 2.8+4 2.7+4 Kr-87 1.7+6 9.8+5 5.8+5 2.4+4 4.7+0 g Kr-88 2.6+6 2.1+6 1.6+6 3.7+5 6.9+3 Kr-89 2.2+3 6.2-3 1.4-8 1.9-42 0.0 Xe-131m 2.0+4 2.0+4 2.0+4 2.0+4 2.0+4 Xe-133m 1.9+5 1.9+5 1.9+5 1.8+5 1.6+5 Xe-133 5.3+6 5.3+6 5.2+6 6.2+6 5.0+6 Xe-135m 1.u+6 1.2+6 1.1+6 6.0+5 1.1+5 Xe-135 5.0+6 4.9+6 4.8+6 4.0+6 1.8+6 Xe -13 7 1.6+4 5.3-1 Xe-138 1.0+6 1.5 5.9-33 0.0 9.0+4 7.8+3 3.3-3 3.2-20 Total NG 1.8+7 1.5+7 1.4+7 1.1+7 7.1+6 Notes b.9+4 = 6.9x10* b

          -                                       ~            ~       -

1105 008 0 15.1-164 Supple:: tent 6 7/3/75

   .                                                              MONTAGUE 162 PSAR TABLE 15.1.38-5 CONTROL ROD DROP ACf'TDDrt_F_Tj'ETON PRODt!CT RELF.ASE l< ATM' m t r#VI: nit:Mhg CONSEl<VATIVE (NFC) ANALYSIS Release Rate (Ci/sec) vs Time Isotope             1 min                   _1 hr            2 hr                                      8 hr        1 day I-131              4.0-3                      4.0-3          4.0-3                                    3.9-3        3.6-3 I-132              5.8-3                      4.3-3          3.2-3                                     5.1-4       4.0-6 I-133              6.4-3                      6.2-3          6.0-3                                     4.9-3       2.9-3 I-134              7.2-3                      3.3-3          1.5-3                                     1.3-5       4.0-11 I-135               6.2-3                     5.6-3          5.1-3                                     2.7-3       5.2-4 Total Iodine             2.8-2                      2.3-2          2.0-2                                     1.2-2       7.0-3 Kr-83m              1.6 -2                    1.1-2          7.6-3                                     8.6-4       2.5-6 Kr-85m              3.2-2                     2.7-2          2.3-2                                     9.1-3       7.3-4 Kr-85               1.6-3                     1.6-3          1.6-3                                     1.6-3       1.6-3 Kr-87               9.6-2                     5.7-2          3.3-2                                     1.4 -3      2.7            Kr-88               1.5-1                     1.2-1          9.3-2                                     2.1-2       4.0-4 Kr-89               1.3-4                     3.6-10         8.1-16                                          0       0,     g Xe-13tn             1.2-3                     1.2-3          1.2-3                                     1.2-3       1.2-3                 .

Xe-133m 1.1-2 1.1-2 1.1-2 1.0-2 9.4-3 Xn-13 3 3.0-1 3.0-1 3.0-1 3.0-1 2.9-1 Xu- 135m 8.0-2 7.1-2 6.4-2 3.5-2 6.6-3 Xe-135 2.9-1 2.8-1 2.8-1 2.3-1 1.0-1 Xe- 137 9.5-4 3.1-8 8.4-13 3 0 Xe-138 5.8-2 5.2-3 4.5-4 1.9 10 1.8-27 Total NG 1.0+1 8.9-1 8.2-2 6.1-1 4.1-1 Note 4.0-3 = 4.0x10-8 i, IiOS 009 J 15.1-165 Supplement 6 7/3/75

MONTAGUE IS2 PSAR TABLE 15.1.38-6 CONTROL ROD DROP ACCIDENT - PARAME"'ERS TO BE TABULATED FOR POS"'ULATUD ACCIDENT ANALYSTS Conservative (NRC) Realistic Assicotions Ass' motion s I. Data and assumptions used to estimate , radioactive source from postulated accidents A. Power level 3,758 MWt 3,758 ! Wt 13 . Fuel d.amaqcd , 770 Rod:s 770 Rodo C. knicaun at . activity by nuclide Ta521e 15.1.38.5 Tablo 15.1.38.2 D. Iodine fractions

1. Organic 0 0

2. Elemental 1 1

3. Particulate 0 0 E. Reactor coolant activity before the accident 15.1.39.5.1.2 15.1.39.5.1.2 II. Data and assumptions used to estimate activity released A. Condenser leak rate (%/ day) 0.5 0.5 _

B. Turbine building leak rate (%/ day) = 840 III. Dispersion Data A. Boundary and LP: distances (m) 815/a,023 815/4,023 D. */Qs for time intervals of:

1. 0-2 hr - SB 6.87x10-* 6. d7x 10 -* 5 b

2. 0-8 hr - LP: 2.29x10-* 2.29x10-*

3. 8-24 hr - LPt 7.85x10-* 7.85x10-*

IV. Dose Data A. Method of dose calculation Regulatory Regulatory Guide 1.3 Guide 1.3 B. Dose conversion assumptions Regulatory Regulatory

                                                               - -- Guid e 3 - - --- - Guide 1.3 - -

C. Peak activity concentrations Table 15.1.38.4 Table 15.1.38.1 in containment

                                                   ~~        -

D. Dooes Fig. 15.1.38-1 , Table 15.1.38.3 Ii05 010 . 15.1.38-166 Supplement 6 7/3/75

102, O  : 10' _ I _ i

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__ FIG.15. l. 38- 2 CONTROL RCD DROP ACCIDENTNRC-DRL WHOLE BODY OOSE VERSUS DISTANCE MONTAGUE NUCLEAR POWER STAT IO N i 10b. 01i UNITS I AND 2 PRELIMINARY SAFETY ANALYSIS REPORT SUPPLEM ENT 7 12 /12 / 7 5

a . . 0.016 0.015 -

                                                                            ^

0.014 0.0 ' 3 - 0.012 - 0.011 - 0.010 - 0.009 -

                                                                                     -                      1 0.008   -

0.007 - O.006 - 0.005 - 0.004 - e o,',, - I' 0.002 - 0.CCI - _ ,_ _, _ _, _ y ___ _ .,._ __ _ ,_ _ ___, _ _ _ , _ . _ . , _ . _ _ , .- 0 1 2 3 4 5 6 7 8 9 to 11 12 feet CENTER ACO IS OUT Figure 15.1.38- 1 Design Basis Rod Drop Accident Shape Function (Beginning of Life) 110S, 012 15.1 '17Ca 083074 Supplenent 3 12/12/74

  .e O.016 0.015     =

0.014 - 0.013 0.012 -

                                                                                                                                  -s 0.011     -

0.010 - 0.000 = 0.008 - 0.007 - O 0.006 - t 0.005 - 0.004 - b 0.003 - 0.002 - 0.001 -

g. . . _ .

_g. -_g _ . p _ . _ g . _ _ - g - _ . _7 - _ - p . g i 1 2 3 4 5 6 7 8 9 10 11 12 im CENTER ROD IS CUT Figure 15.1.38-4. Design Basis Rod Drop Accident Shape Function (3.5 C'a'd/ ) 15.1 170b 083074 3"99 *=ent 3 ,2/,2/74 0S 013

                                                                                                .~    . . ~

0.016 0.015 - 0.014 - 0.013 - 0.012 - O ) 0.011 - 0.010 - O.009 - 0.008 - 0.007 - 0.006 - 0.005 - 0.004 - 0.003 - o 0.002 1 0.001 - I t I t I . _ _ . . t_ _ _ . I --_i-. - I -l 6 1 2 3 4 5 6 7 8 9 ,10 11 12 feet CENTE R AOO IS CUT 110S 014 Figure 15.1.38- 5 Design Basis Rod Drop Accident Shape Function (~ .4 Cd/t) 15.1-170c 083074 Supple =ent 3 12/12/7/.}}