ML19263C488

Proposes Amend to License DRR-65.Proposed Changes Include Core Performance Characteristics & Safety Analyses Re Operation at 2,700 Mwt

ML19263C488
Person / Time
Site:	Millstone
Issue date:	02/12/1979
From:	Counsil W NORTHEAST UTILITIES
To:	Reid R Office of Nuclear Reactor Regulation
References
TAC-46174, NUDOCS 7902260284
Download: ML19263C488 (400)

Text

{{#Wiki_filter:NOH'I'HEAS'I' NORTHEAST UTILITIES SERVICE COMPANY U'I'IGl'I'IGS UOh5o"SouNecTl cut 06101 203-666-6911 February 12, 1979 Docket No. 50-336 Director of Nuclear Reactor Regulation Attn: Mr. R. Reid, Chief Operating Reactors Branch #4 U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Gentlemen:

References:

(1) W. G. Counsil letter to R. Reid dated November 1,1978. (2) W. G. Counsil letter to R. Reid dated December 15, 1978. (3) W. G. Counsil letter to R. Reid dated January 17, 1979. Millstone Nuclear Pcwer Station, Unit No. 2 Proposed License Amendment, Power Uprating In Reference (1), Northeast Nuclear Energy Company (NNECO) identified its intent to increase licensed core thermal power to 2700 MWt for Cycle 3 operation, and proposed a schedule of submittals associated with the uprating program. This topic was the subject of a meeting in your Bethesda offices on November 21, 1978. This letter is being docketed to fulfill commitments made in Reference (1) and to address various topics discussed durirg the above mentioned meeting. Pursuant to 10CFR50.90, NNEC0 hereby proposes to amend its operating license, DPR-65, by incorporating the attached proposed revisions into the Millstone Unit No. 2 Technical Specifications. The proposed changes are proviaed in Section 8 of the attached report, and fulfill the commitments made in Reference (2) by identifying the changes necessary in DPR-65 and in Appendix A to support operation at 2700 MWt. Sections 1 through 7 provide a summary of the Cycle 3 relcad, including core per-formance characteristics and the appropriate safety analyses. Section 8 contains the revised pages of the license, Appendix A, and also a revised page in Appendix B to reflect the proposed increase in thermal power. The principal tests in the startup test program are provided in Section 9, as well as a discussion of accep-tance criteria. 79022602f 6 TW CONNECTICUT UGHT AND POWER CCYPANY YoE PARTrC9D E CCTR'C UGHT COVFANY M 51tDN UngsarHUEE TTS EL FCT AIC CCMPANY

• N YCkt W ATE R PGM H CC'i'ANY NOATHE AST UTIUT:ES SI'R it CE CCVPANY

In Reference (2), NNECO provided an evaluation of the radiological consequences of five inciden:,s, and indicated that the remaining five would be evaluated in this letter. Lvo of these, the Steam Generator Tube Rupture and the Control Room Dose to Operatres Post-LOCA, are addressed in Section 10 of the attached report. The remaining three incidents, the Main Steam Line Rupture, the Control Rod Ejec-tion, and the Seized Rotor, are not addressed as certain inputs are not yet available from our NSSS vendor, Combustion Engineering. The radiological conse-quences of these incidents are expected to be well within 10CFR100 limitations; however, cycle-specific verification cannot be provided at this time. An evaluation of the ranaining three incidents will be submitted as soon as the information becomes available. During the lovember 21, 1978 meeting, several questions raised by the Staff were not completely addressed during the meeting. The following infonnation is pro-vided to address those questions.

1. Concerning the small break LOCA ana}ysis, it is anticipated tgat the limiting break size will be 0.1 f t. rather than the 0.05 ft. break.

This is due primarily to two NRC approved changes to the Combustion Engineering CEFLASH-4AS model . The first involves a more sophisticated phase separation model for more detailed axial power distribution. The second concerns changes to the steam generator model to more accurately treat steam condensation on the primary side. The net effect of these changes is a shift in the limiting break in the direction of larger breaks.

2. The four reduced flow CEA guide tube assemblies will all be located under Bank 7 CEA's. Bank 7 CEA's are used for power shaping during normal operation. Concerns of flow redistribution due to utilization of the four reduced flow assemblies is not justified. The change in RCS bypass flow by the introduction of four reduced flow assemblies is less than 0.05%. This change is readily accommodated by the assumed uncertainties in relevant analyses.

3. Short TORC as opposed to long TORC was utilized for Millstone Unit No.

2 - specific calculations. The short version yields conservative re-sults when compared to the long version. The above proposed changes have been reviewed pursuant to 10CFR50.59 and one item has been found to constitute an unreviewed safety question. The results of the Seized Rotor Event presented in Section 7.3.4 indicate that the calculated failure fuel fraction is 1.0%. Because Cycle 2 analyses indicate that no fuel failures are predicted for this event, this item is judged to be an increase in consequences and as such constitutes an unreviewed safety question. Nonetheless, Section 7.3.4 indicates that this condition is acceptable. The radiological consequences of this incident will be submitted as soon as the information becomes available. NNECO cannot render a final safety evaluation of the Cycle 3 reload in that the radiological consequences of three incidents have not been calculated, and both large and small break ECCS performance calculations have not been completed. Pre-liminary information indicates that these analyses will confirm that Millstone

                                          .3 -
                                                       *f Unit No. 2 can be safely operated at 2700 MWt during Cycle 3 with the Technical Specifications proposed in Section 8. With the exception noted above, NNEC0 has concluded that the remaining Technical Specification changes do not constitute an unreviewed safety question.

The Millstone Unit No. 2 Nuclear Review Board has reviewed and approved the above proposed changes, and concurred in the above determinations. NNEC0 has reviewed the above proposed changes pursuant to the requirements of 10CFR170, and has determined that no additional fee is required. The basis for this determination is provided in Reference (3). The resources of NNEC0 and NUSCO remain available to expedite Staff review and approval of the attached material. Very truly yours, NORTHEAST NUCLEAR ENERGY COMPANY

                                                   /        ~

b4 ' W. G. Counsil Vice President Attachments

STATE OF CONNECTICUT )  %

                      ) es. Berlin COUNTY OF liARTFORD    )                                     '# # !

Then personally appeared before me W. G. Counsil, who being duly sworn, did state that he is Vice President of Northeast Nuclear Energy Company, a Licensee herein, that he is authorized to execute and file the foregoing information in the name and on behalf of the Licensees herein and that the statements contained in said information are true and correct to the best of his knowledge and belief. Notary Public

                                                 'M. l My 00:T:miss!:a Eg.rcs l,brd 31, ][31

1. INTRODUCTION AND

SUMMARY

This report provides an evaluation of the design and performance of Millstone Point Unit II during its third fuel cycle, its first at a full rated power of 2700 MWT; Cycles 1 and 2 were run at a full power rating of 2560 MWT. All planned operating conditions, with the exception of the stretch power rating and its concurrent operating characteristics, remain the same as for Cycle 2. The core will consist of presently operating Batch B, C and D assemblies in addition to fresn Batch E fuel. The data presented herein assumes a nominal Cycle 2 endpoint of 8700 MWD /T. However, system requirements have created a need for a flexible cycle length ranging from 8100 MWD /T to 9300 MUD /T. In performing analyses of postulated accidents, determining limiting safety settings and establishing limiting conditions for operation, limiting values of key parameters were chosen to assure that expected Cycle 3 conditions are enveloped for the above burnup range of Cycle 2 termination points. During scheduled shutdown maintenance af ter Cycle 1 operation, it was observed that CEA fingers caused wear in the CEA guide tubes as reported in Reference 1. The mechanical integrity of the worn areas was restored by installing stainless steel sleeves in the guide tubes. Sleeving was performed on those assemblies which had sustained wear in Cycle 1 and 6n those assemblies which were installed in CEA locations in Cycle 2. For Cycle 3 operation, those assemblies will also be sleeved which were not placed under CEAs in cycles 1 and 2 but will be located under CEAs in Cycle 3. Sixty - eight Batch E assemblies will also incorporate stainless steel sleeves in the guide tubes. The remaining four Batch E assemblies will utilize the guide tube flow hole modification for mitigating guide tube wear as described for sixteen assemblies for Calvert Cliffs II Cycle 2 in Reference 2.

The evaluations of the reload core characteristics have been examined with respect to the Cycle 2 safety analysis described in Reference 3, hereafter referred to as the " reference cycle". In all cases, it has been concluded that either the reference cycle analyses envelope the new conditions or that the revised analyses presented here continue to show acceptable results. Where dictated by variations from the reference cycle, proposed modifications to the plant Technical Specifications are provided and are justified by the analyses reported herein.

2. OPERATING HISTORY Millstone Point Unit II is presently operating in its second fuel cycle utilizing Batch B, C and D fuel assemblies at a core power level of 2560 MWT. The startup experience was reported to the NRC in Reference 4. Operation has continued in the interim at or near licensed power with only brief reductions and outages as necessitated by routine maintenance, testing and minor repairs.

No unusual carryover effects are anticipated from the second to the third fuel cycle. It is presently estimated that Cycle 2 will terminate in March, 1979. However, flexibility in this date is necessary because of uncertainties in future station capacity factor and secondary side maintenance requirements. Depending upon these considerations, the Cycle 2 termination point was expected to be between 8100 and 9300 MWD /T. As of January 31, 1979, a Cycle 2 burnup of 8100 MWD /T had been reached. Operation of Cycle 3 is estimated to commence in mid-May,1979.

4

3. GENERAL DESCRIPTION The Cycle 3 core will consist of the types and quantities of fuel batches outlined in Table 3-1. The primary change for Cycle 3 is the removal of 72 Batch B assemblies and their replacement by 72 new Batch E assemblies.

One fuel management pattern (Figure 3-1) has been developed which will accomodate Cycle 2 endpoint flexibility between 8100 and' 9300 MWD /T. This Cycle 3 core loading pattern is 90 rotationally symmetric. That is, if one quadrant of the core were rotated 90* into its neighboring quadrant, each assembly would be aligned with a similar assembly. This similarity includes batch types, number of fuel rods, initial enrichment and beginning of cycle burnup. Figure 3-2 shows the beginning of Cycle 3 burnup distribution based on a Cycle 2 endpoint of 8700 MWD /T. The initial enrichment of each assembly is also shown.

TABLE 3-1 MILLSTONE F0 INT II CYCLE 3 CORE LOADING Initial Initial Average Number Shim Total Assembly Number of Enrichment Burnup* of Loading Total Fuel Designation Assemblies wt% U-235 MWD /MTU Shims wt% BuC Shims Rods B+ 5 2.33 25,400 12 2.7 60 820 40 2.82 19,700 0 --- 0 7,040 C C+ 16 2.82 24,800 12 .83 192 2,624 C. 12 2.82 24,900 12 .46 144 1,968 D 48 3.03 7,6 00 0 --- 0 8,448 D* 24 2.73 10,600 0 --- 0 4,224 E 48 3.24 0 0 --- 0 8,448 E* 24 2.73 0 0 --- 0 4,224 217 396 37,796 Notes

• Assumes a Cycle 2 length of 8,700 MWD /T.

                                         ~

E E E' E E D C E D D C+ C E* E D* C E* C+ D C. E D C C D D* E* C E D E* D B D* C* D* E C+ C+ D* D* C D C E D C D E* C. D E* C E C E* C. C D* C C B Millstone Point 11 cycle a LOADING PATTERN 3-1

X.XX INITIAL ENRICHMENT, WT% U-235 YYYY BOC 3 BURNUP MWDIT 3.24 3.24 0 0 3.24 3.24 3.24 3.03 2.82 0 0 0 7,900 20,300 3.24 3.03 3.03 2.82 2.82 2.73 0 6,300 9,600 24,000 19,800 0 3.24 2.73 2.82 2.73 2.82 3.03 2.82 0 10,700 18,300 0 25,600 7,000 25,100 3.24 3.03 2.82 2.82 3.03 2.73 2.73 2.82 0 6,300 18,300 17,700 8,400 10,000 0 20,300 3.24 3.03 2.73 3.03 2.33 2.73 2.82 2.73 0 9,600 0 8,400 25,400 10,700 24,600 10,900 3.24 2.82 2.82 2.73 2.73 2.82 3.03 2.82 3.24' 0 24,000 25,600 10,000 '10,800 21,100 6,~400 18,900 0 3.03_ 2.82 3.03 2.73 2.82 3.03 2.73 2.82 3.24 7,900 19,800 7,000 0 24,700 6,400 0 21,050 0 2.82 2.73 2.82 2.82 2.73 2.82 2.82 2.33 20,300 0 25,100 20,300 10,900 18,900 21,050 25,400 Millstone Point 11 ASSEMBLY AVERAGE BURNUP AND cycle a INITIAL ENRICHMENT DISTRIBUTION 3-2

4. FUEL DESIGN 4.1 MECHANICAL DESIGN The fuel assembly complement for Cycle 3 is given in Table 3-1. The mechanical design of the reload fuel assemblies, Batch E, is essentially identical to that introduced with the Millstone II Bi .ch D fuel with the exception of a slight reduction in fuel rod fill pressure. The major design changes introduced to reduce CEA guide tube wear are detailed in Section 4.2.

C-E has performed analytical predictions of cladding creep collapse time for all Hillstone II fuel batches that_ will be irradiated during Cycle 3 and has concluded that the collapse resistance of all fuel rods is suffi-cient to preclude collapse during their design lifetime. The analyses utilized the CEPAN computer code (Reference 5) and included as input conservative values of internal pressure, cladding dimensions, cladding tenperature and neutron flux. 4.2 HARDWARE MODIFICATIONS TO MITIGATE GUIDE TUBE WEAR The mechanical design of the fuel assemblies is as described in the preceding section except that (1) those fuel assemblies which sustained substantial r side tube wear in Cycle 1 have had stainless steel sleeves installed in the guide tubes as a means of improving the mechanical strength margins in the worn areas, (2) all Batch B, C and D fuel assemblies to be installed in CEA locations in Cycle 3 will have stainless steel sleeves installed in the guide tubes in order to prevent guide tube wear, (3) sixty-eight Batch E fuel assemblies will have stainless steel sleeves installed in the guide tubes in order to prevent guide tube wear, and (4) the remaining four Batch E fuel assemblies will have the guide tube flow hole modification which mitigates the degree of guide tube wear to a point where sleeves are not required. A detailed discussion of the design of the sleeves and its effects on reactor operation is contained in Reference 6. A detailed discussion of the Batch E design change is included in Reference 2.

4.3 THERMAL DESIGN

     'u>ing the FATES model (Reference 7 ), the thermal performance of the various types of fuel assemblies has been evaluated with respect to their Cycle 1 and Cycle 2 burnups, proposed burnups during Cycle 3, their respective fuel geometries and expected flux levels during Cycle 3. The twice burned assemblies have been determined to be the limiting fuel batch with respect to stored energy.

4.4 CHEMICAL DESIGN The metallurgical requirements of the fuel cladding and the fuel assembly structural members for the Batch E fuel have not been changed from the original Cycle 1 and Cycle 2 designs. Therefore, the chemical or metallurgical performance of the Batch E fuel will be unchanged ' rom that of the original core fuel and discussions in the FSAR, Reference 8, are still valid. 4.5 OPERATING EXPERIENCE Fuel assemblies incorporating the same design features as the Batch E fuel assemblies have had operating experiences at Calvert Cliffs 1, Fort Calhoun 1, St. Lucie I and Maine-Yankee. The operating experience has been successful except for the CEA guide tube wear problem which has been addressed in Section 4.2.

5. NUCLEAR DESIGN 5.1 PHY3ICS CHARACTERISTICS 5.1.1 Fuel Management Figure 3-1 and Table 3-1 summarize the core design for Cycle 3.

The fuel management is three-batch with a mixed central zone. The newly installed Batch E is comprised of two sets of assemblies each set having a unique enrichment in order to minimize radial power peaking. The projected end of Cycle 2 burnup is expected to be between 8100 and 9300 MWD / T . A loading pattern has been devised which will accommodate this potential range of Cycle 2 endpoints as noted in Section 3. With the replacement of 72 Type B assemblies by fresh Type E assemblies, the Cycle 3 burnup capacity for full power operation is expected to be between 10,500 and 11,300 MWD / T depending on the final Cycle 2 termination point. 5.1.2 Physics Characteristics Physics characteristics for Cycle 3 are listed in Table 5-1 along with the corresponding values for the reference cycle. Table 5-2 presents a summary of CEA shutdown worths for Cycle 3 with a comparison to reference cycle data. In this table, the reactivity allowances are also listed for both Cycle 3 and the reference cycle with the resulting shutdown margins. The power dependent insertion limit for Cycle 3 is shown in Figure 5-1. CEA group identification is shown in Figure 5-2. Table 5-3 shows the reactivity worths of various CEA groups calculated at full power conditions for Cycle 3 and the reference cycle.

5.1.3 Power Distribution Figures 5-3 through 5-5 illustrate the all rods out (ARO) planar radial power distributions at B0C 3, M0C 3 and E0C 3 that are characteristic of the high burnup end of the Cycle 2 shutdown window. These planar radial power peaks are characteristic of the major portion of the active core length between about 20 and 80 percent of the fuel height. Figure 5-6 illustrates the plann/ radial power distribution within the upper 20 percent of the core produced with the insertion of the first CEA regulating group, Bank 7. In this case, the power distribution shown is predicated on the low burnup end of the Cycle 2 shutdown window, providing an illustration of maximum power peaking expected for this configuration. Higher burnup Cycle 2 shutdown points tend to reduce power peaking in this upper region of the core with Bank 7 inserted. It is a characteristic of both ARO and Bank 7 inserted conditions that the Cycle 3 peaks are highest at BOC and decrease with cycle burnup. The radial power distributions described in this section are calculated data without uncertainties or other allowances. However, single rod power peaking values do include the increased peaking that is characteristic of fuel rods adjoining the water holes in the fuel assembly lattice. For both DNB and kw/ft safety and setpoint analyses in either rodded or unrodded configurations, the power peaking values actually used are higher than those expected to occur at any time during Cycle 3. These conservative values, which are specified in Sections 7 and 9 of this document, establish the allowable limits for power peaking to be observed during operation.

The range of allowable axial peaking is defined by the limiting conditions fer operation of the axial shape index (ASI). Within these ASI limits, the necessary DNBR and kw/ft margins are maintained for a wide range of possible axial shapes. The maximum three-dimensional or total peaking factor anticipated in Cycle 3 during normal base load, all rods out operation at full power is 1.87, not including uncertainty allowances and augmentation factors. 5.1.4 Reactivity Coefficients Expected values of the fuel and moderator temperature coefficients are presented in Table 5-1. The values of fuel and moderator temperature coefficients used in the safety analyses for Cycle 3 were chosen in such a manner as to conservatively bound these values for the cycle. 5.1.5 Safety Related Data 5.1.5.1 Ejected CEA The calculated reactivity worths and planar radial power peaks are shown in Table 5-4 for the limiting conditions during the cycle. These values encompass the worst conditions anticipated during Cycle 3. The pointwise Doppler feedback technique was not utilized in the PDQ calculations of ejected CEA worths and associated power peaks in order to be consistent with the Doppler treatment in the safety analysis itself (see Section 7.3.1). r.l.5.2 s Dropped CEA Table 5-5 contains data for dropped CEA configurations (for both extremes in core life) and covers the range of expected values. These data have been calculated with pointwise Doppler feedback as was done for the reference cycle (reference 3). This treatment is consistent with the

safety analysis since the time to minimum DNBR is on the order of one to two minutes allowing ample time for fuel temperature redistribution following the CEA drop. 5.1.6 Augmentation Factors Augmentation factors have been calculated for the Cycle 3 core using the calculational model described in Reference 7. The input information required for the calculation of augmentation ' factors that is specific to the core under consideration includes the fuel densification characteristics, the radial pin power distribution and the single gap peaking factors. Augmentation factors for the Cycle 3 core have been conservatively calculated by combining for input the largest single gap peaking factors (calculated near end of cycle) with the most conservative (flattest) radial pin power distribution. The calculations yield non-collapsed clad augmentation factors showing a maximum value of 1.054 at the top of the core. The augmentation factors for Cycle 3 are compared to the reference cycle values calculated with the same model in Table 5-6. 5.2 PHYSICS ANALYSIS METHODS 5.2.1 Analytical Input to In-core Measurements In-core detector measurement constants to be used in evaluating the reload cycle power distributions will be calculated in the manner described in Reference 9.

5.2.2 Uncertainties in Measured Power Distributions The power distribution measurement biases and uncertainties which are applied to Cycle 3 are 6.0% on Fr and 7.0% on qF . These values are to be used when the INCA model described in Reference 9 is used for monitorina power distribution parameters during operation. 5.2.3 Nuclear Design Methodology 5.2.3.1 Use of Coarse Mesh Neutronics Calculations The coarse mesh computer program ROCS (Reference 10) has been used along with the standard fine mesh design program PDQ (Reference 11) in the Cycle 3 safety analysis,

a. ROCS was used to survey a variety of core configurations to determine limiting conditions.

b. ROCS was used to obtain axial power shapes,to weight the relative importance of fine mesh PDQ planar power and burnup distributions in the detemination of three-dimensional effects and to detennine the impact of the three-dimensional gross power distributions on reactivity parameters.

c. ROCS was used to compute selected safety parameters. The calculation of those limiting parameters which require knowledge of 1-pin peaking factors continues to be based on the fine mesh PDQ program.

d. Two- and three-dimensional ROCS calculations were used in conjunction with two-dimensional PDQ calculations to obtain best estimate core parameters such as those shown in Table 5-3.

TABLE 5-1 MILLSTONE F0 INT II CYCLE 3 PHYSICS 'CHARACTERISTICS Reference Units Cycle Cycle 3 Dissolved Boron Dissolved boron content for criticality, CEAs withdrawn hot, full power, equilibrium PPM 660 830 xenon, BOC Boron Worth Full power BOC PPM /%Ap 88 93 Full Power EOC PPM /%Ap 77 82 Reactivity C6efficients (CEAs Withdrawn Moderator temperature coefficients, hot operating, hot, full power, equilibrium 10-4 ap/ F . f, .2 xenon, B0C hot, full power, E0C 10~4 Ap/ F -2.0 1.8 Doppler coefficient hot B0C zero power 10-5 3 joy _),44 ,),44 hot BOC full power 10-5 a/F -1.13 -1.13 hot E0C full power 10-5 a/F -1.22 -1.22 Total Delayed Neutron Fraction 8 eff BOC .00608 .00624 E0C .00526 .00524 Neutron Generation Time, t* . BOC 10-6 sec 28.0 27.2 E0C 10-6 sec 32.2 31.8

TABLE 5-2 MILLSTONE POINT II CYCLE 3 LIMITING VALUES OF CEA REACTIVITY WORTHS AND ALLOWANCES,

                                                       %ao BOC                              E0C Reference Cycle      Reload Cycle Reference Cycle   Reload Cycle Worth Available 9.0               9.7           10.0

• 11.0 Worth of all CEAs inserted Stuck CEA allowance 3.0 3.1 3.1 3.5 Worth of all CEAs less highest worth CEA 6.0 6.6 6.9 7.5 stuck out Worth Reouired (Allowancesi Power defect, HFP to HZP 1*7 1*8 2*4 2*2 (Doppler, Tavg, Redistribution) 0.0 0.0 0.1 0.1 Moderator voids CEA bite, boron deadband and maneuvering 0.4 0.6 0.4 0.6 band Shutdown margin and safeguards allowance 3.2 3.2 3.2 3.2 Total reactivity required 5.3 5.6 6.1 61 Available Worth less Allowances 1.4 Margin available 0.7 1.0 0.8

TABLE 5-3 MILLSTONE POINT II CYCLE 3 REACTIVITY WORTH OF CEA REGULATING GROUPS AT HOT FULL POWER %Ap* LOC E0C degulating CEAs Reference Cycle Cycle 3 Reference Cycle Cycle 3 Group 7 0.76 0.66 0.82 0.76 Group 6 0.48 0.36 0.49 0.44 Group 5 0.37 0.38 0.38 0.44 Group 4 1.16 0.92 1.26 1.05 Notes

• Values shown assume sequential group insertion.

TABLE 5-4 MILLSTONE POINT II CYCLE 3 CEA EJECTION DATA Limitina Value Maximum Radial Power Peak Reference Cycle Cycle 3 Full Power with Bank 7 Inserted; Worst CEA Ejected 4.04 3.90 Zero Power with Banks 7+6+5+4 Inserted;Uorst CEA Ejected 7.99 8.34 MaximumEjectedCEhWorth(%ao) Full Power, with Bank 7 Inserted; Worst CEA Ejected .31 .29 Zero Power with Banks 7+6+5+4 Inserted; Worst CEA Ejected .74 .65 Note: Appropriate uncertainties are included in the above data.

TABLE 5-5 MILLSTONE POINT II CYCLE 3 FULL LENGTH CEA DROP DATA AT FULL POWER Limiting Values Reference Cycle Cycle 3 Minimum Worth, %Ao .10 .08 17.0 16.0 Maximum Percent Increase in Pin Peak Notes

1. No uncertainties are included in above data.

2. CEAs are either fully withdrawn or fully inserted for radial calcelations.

                                                                           /

TABLE 5-6 MILLSTONE POINT II CYCLE 3 AUGMENTATION FACT 0ks AND GAP SIZES FOR CYCLE 3 AND REFERENCE CYCLE Reference Cycle Cycle 3 Noncollapsed Noncollapsed Core Core Clad Gap Clad Gap Height Height Augmentation Size Augmentation Size (percent) (inches) Factor (inches) Factor (inches) 98.5 134.7 1.045 1.606 1.054 1.743 86.8 118.6 1.041 1.417 1.049 1.538 77.9 106.5 1.038 1.280 1.045 1.384 66.2 90.5 1.033 1.087 1.040 1.179 54.4 74.4 1.028 1.898 1.034 0.974 45.6 62.3 1.025 0.756 1.029 0.820 33.8 46.2 1.020 0.567 1.023 0.61 5 22.1 30.2 1.014 0.378 1.016 0.410 13.2 18.1 1.010 0.236 1.011 0.256 1.5 2.0 1.002 0.047 1.002 0.0 51 Notes Values are based on approved model described in Reference 7 e 4.

25%

1. 00 '- (1.00,7 025) 1 0.00 -

l 5 I h 0.30 - I l (0.73, 7-046) k O.70 l E l (0.65,6-025)

               ~

(0.5G,6 050)

                             !                                           TRANSIENT INSERTION LIMIT b                               LONG TERM g 0.50 l STEADY STATE cr                         l lNSERTION Li,MIT o 0.40       -

l 8

 ,                                                 I h                                               SHORT TERM g 0.30      -

l STE ADY ST AT E I SERTION LIMil (0.20, 5-060)

u. 0.20 -

0.10 - 0 Banks i e i 7 i i , , , , 5 , , , , , , 3 , , , 0 20 40 60 80 100 0 20 40 60 80 100 0 20 40 60 80 100

                                          ,      i       i 6   ,      ,     ,     ,     ,      ,  4  ,  ,    ,

0 20 40 60 80 100 0 20 40 60 80 100

                                                                        % OF CEA INSERTION kiillstone Point il               CEA INSERTION LIMITS vs THERMAL POWER                       Figure cnie 3                      WITH FOUR REACTOR COOLANT PUMPS OPERATING                               5-1

NUMBER CEA CEA 0F GROUP TYPES CEAS 7 4C 9 6 1, 8 , 9W 4 5 2 4 . 4 12 8 3 6 8 2 7 8 1 5 8 8 10 8 A 11 16 i 6 11 12 4C N

                                                 \

9W 11 5 11 7 N N 11 8 12 7 10 2

                                                         '\
                                                            \

6 5 10 4C 1 2 X CEA TYPE lAillstone Point 11

0.79 1.00 0.74 1.04 1.20 1.16 0.94 0.72 1.06 1.16 0.87 0.94 1.28 0.72 0.83 0.91 1.28 0.86 1.19 0.92 0.74 1.06 0.91 0.90 1.12 1.07 1.27 0.93 1.05 1.16 1.29 1.12 0.72 0.93 0.84 0.94 1.22 0.88 0.87 1.07 0.94 0.85 1.13 0.88 0.81 1.18 0.95 1.19 1.26 0.84 1.13 1.22 0.85 1.02 0.98 1.30 0.92 0.92 0.94 0.89 0.85 0.67 X NOTE: X= MAXIMUM l-PIN PEAK =1.50 re Millstone Point 11 ASSEMBLY RELATIVE POWER DENSITY fIj"3 cycle 3 BOC, EQUILIBRIUM XENON

0.76 0.94 0.74 0.99 1.13 1.09 0.91 0.76 1.05 1.11 0.88 0.93 1.2d 0.76 0. 51 0.93 1.26 0.88 1.17 0.93 0.74 1.05 0,93 0.94 1.13 1.08 1.26 0.95 0.99 1.11 1.26 1.13 0.79 1.00 0.89 1.00 1.13 0.88 0.88 1.08 1.00 0.91 1.17 0.95 0.76 1.10 0.93 1.17 1.26 0.89 1.1d 1.26 0.91 0.94 0.93 1.24 0.92 0.94 1.00 0.95 0.91 0.76 I NOTE: X= MAXIMUM 1-PIN PEAK =1.45 Figure Millstone Point 11 ASSEMBLY RELATIVE POWER DENSITY 5-4 cycle a MOC, EQUILIBRIUM XENON

0.78 0.95 0.76 0.99 1.12 1.08 0.92 0.80 1.05 1.09 0.89 0.93 1.20 0.80 0.91 0.95 1.23 0.90 1.14 0.93

                                                  ~

0.76 1.05 0.95 0.95 1.12 1.06 1.22 0.95 0.99 1.09 1.23 1.12 0.83 1.00 0.91 1.00 1.12 0.89 0.90 1.06 1.01 0.93 1.15 0.95 0.78 1.08 0.93 1.14 1.21 0.91 1.15 1.22 0.92 X 0.95 0.93 1.20 0.92 0.94 1.00 0.96 0.92 0.80 NOTE: X MAXIMUM 1-PIN PEAK - 1.38 Figure Millstone Point 11 ASSEMBLY RELATIVE POWER DENSITY 5-$ cycle 3 EOC EQUILIBRIUM XENON

BANK 7 CEA INSERTIONS 0.66 0.80 0.65 0.93 1.06 0.98 0.76

                                                                            /

0.53 0.94 1.12 0.87 0.83 .

                                                                            // /

0.53

                                 '/ .48
                                 /  0
                                     / /'    O.84    1.30    0.97    1.22      0.92
                                  //b 0.65    0.94      0.84    0.96    1.27    1.25    1.42      1.10 0.93    1.12      1.29    1.27    0.91    1.16    1.07      1.17 1.06   0.87      0.97    1.24    1.16    1.08    1.40      1.09 0.66 0.98   0.82      1.20    1.41    1.07    1.41    1.40      0.97 0.81                                                              X
                          '////                                              '////

0.78 0 75 0.88 1.06 1.18 1.13 0.96 /.51 0 /

                          / .//
                          //                                                 ////

NOTE: X = MAXIMUM 1-PIN PEAK = 1.69 "S"" Millstone Point 11 ASSEMBLY RELATIVE POWER DENSITY WITH Cycle 3 CEA BANK 7 INSERTED AT HFP BOC 5-6

6. Thernal-Hydraulic Design 6.1 DNBR Analysis Steady state DNBR analyses of Cycle 3 at the rated power of 2700 tiWT have been performed using the TORC code (Reference 12) with the CE-1 DNB correla-tion (Reference 13). Table 6-1 contains a list of pertinent thermal-hydraulic design parameters used for both the safety analysis and the generation of reactor protection. system setpoint information.

The analyses were performed to include the following two considerations.

1. TORC was used in the generation of limiting conditions for operation on DNBR nargin in the Technical Specifications. TORC was also used for all A00's and postulated accidents which were analyzed for Cycle 3.

2. The engineering factor on local heat flux, the nuclear uncertainty on Fr and an additional factor designated as an augmentation factor for fuel rod bowing were combined statistically for these analyses. (The method by which these factors are combined is described in Section 4 of Refer-ence14.) Because the local heat flux factor was applied in this way, credit for its inclusion in the TORC deck was taken in determining the setpoints. This is identical to the procedure used in the revised analysis for Calvert Cliffs Unit 1 Cycle 3.

In the DNB limit analysis, the assumed uncertainties in various measured parameters are not combined in a single equation but are factored into func-tional relationships as biases at various points in the analysis. This biasing of functional relationships throughout the analysis is equivalent to adding the absolute power uncertainties equivalent to the uncertainties in the various measured parameters and applying the total power uncertainty to the best estimate calculation. The specific uncertainties along with their equivalent power uncertainties are given below.

ASI 0.06 ASIU 22.2% Pressure 22 PSI 108% Temperature 2*F 2.0.9% Flow 4% 2.5.0% Power 5% (LSSS) 135% 2% (LCO) 14% 1 In the Cycle 3 analysis, the equivalent sum of these uncertainties is 12.4% for LSSS,10.3% for LCO. Treating these measurement uncertainties as statistically independent, the proper method for combining them is Root Sum Square (RSS). The RSS combination yields 6.6% for LSSS, 5.8% for LCO, giving a net conservatism in the analysis of 5.8% for LSSS, 4.5% for LCO. For the Cycle 3 analysis, a partial credit of 3% has been taken for the LC0 and LSSS. Investigations have been made to ascertcin the effect of the CEA guide tube wear problem and the sleeving repair on DNBR margins as established by this type of analysis. The findings were reported to liRC in Reference 6 which concludes that the wear problem and the sleeving repair do not adversely affect DilBR margin. The four batch E assemblies that will utilize the same guide tube flow hole modification for mitigating guide tube waar as des-cribed for Calvert Cliffs II Cycle 2 (Reference 2) are also not expected to experience any adverse effect on DNBR margin.

TABLE 6-1

                 !!illstone-2 Themal-Hydraulic Parameters at Full Power Reference General Characteristics                                 Unit           Cycle 2   Cycle 3 Total Heat Output (core cnly)                        Milt           2560      2700 106 BTU /hr    8737      9215 Fraction of Heat Generated in Fuel Rod                              .975      .975 Prinary System Pressure Nominal                                          PSIA           2250      2250 tiinimun in Steady State                         PSIA           2200      2200 11aximum in Steady State                         PSIA           2300      2300 Design Inlet Temperature                               F            544       551 Total Reactor Coolant Flow                           GPit           370,000   370,000 (liininun Steady State)                         106Lb/hr       140.2*    138.8*

Coolant Flow Through Core 106 Lb/hr 135.0* 133.7* Hydraulic Diameter (flominal Channel) ft 0.044 .044 Average Itass Velocity 106 Lb/hr-ft2 2.53* 2.50* Pressure Drop Across Core (at core low =133x106 ) PSI 10.2 10.3 (!!ininum Steady State Flow irreversible as over entire fuel assembly) Total Pressure Drop Across Vessel (at vessel PSI 33.2 33.4 flow = 138x106) (Based on nomincl dimen-sions and minimum steady state flow) Core Average Heat Flux (accounts for above BTU /hr-ft2 177,700 183,000 fraction of heat generated in fuel rod and axial densification factor) Total Heat Transfer Area (accounts for ft2 47,940 49,100 axial densification factor) Film Coefficient at Average Conditions BTV/hr-ft 2,

                                                         *F             5820      5875 Maximum Clad Surface Temperature                      *F             657       657 Average Film Temperature Difference                   *F             31        31 Average Linear Heat Rate of Undensified' Fuel Rod (accounts for above fraction of heat generated in fuel rod)                    kw/ft         S.99      6.17 Average Core Enthalpy Rise                             BTU /lb       65.*      69.*

Calculated at design inlet temperature, nominal primary system pressure

TABLE 6-1 (Continued) Calculational Factors Engineering Heat Flux Factor 1.03 1.03 Engineering Factor on Hot Channel Heat Input 1.03 1.03 Flow Factors Inlet Plenum Nonuniform Distribution 1.05 Rod Pitch, Bowing and Clad Diameter 1.065 1.065 Fuel Densification Factor 1.01 1.01 Plenum factor not used in TORC; inlet flow distribution is input. a

6.2 EFFECTS OF FUEL R0D B0WIllG ON DNSR MARGIfi Fuel rod bowing effects on DNBR margin for Millstone II, Cycle 3 have been evaluated within the guidelines set forth in Reference 15. Within the range of Cycle 2 termination points and Cycle 3 lifetines identified in this document, no more than 73 assemblies will exceed the DNB reduction or penalty threshold burnup of 24,000 MWD /T. At EOC 3, the maximum burnup attained by any of these assemblies will be 37,100 MWD /T. From Reference 15,the corresponding DNB penalty for 37,100 MWD /T is 4.4 percent. An . examination of the power distributions show that the maximum radial peak at HFP in any of the assemblies that eventually exceed 24,000 MWD /T is at least fifteen percent less than the maximum radial peak in the entire core. Since the percent increase in DNBR has been confirmed to be never less than the percent decrease in radial peak, there exists at least fifteen percent DNB margin for assemblies exceeding 24,000 MWD /T relative to the DNB limits established by other assemblies in the core.

References (Sections 1 through 6)

1. CEN-79-P, " Reactor Operation With Guide Tube Wear". February 3,1978

2. CEN-101(B)-P, "Calvert Cliffs Two Cycle 2 Reload Submittal Update".

August 21, 1978

3. lii11 stone' Unit No. 2 Cycle 2 Proposed License Amendment Relating to Refueling Docket flo. 50-336, September,1977

4. Letter from W. G. Cotmsil to R. Reid, "liillstone Nuclear Power Station, Unit No. 2. Cycle 2 Startup Test Results". Docket No. 50-336, July 13,1978

5. CErlPD-187, "CEPAfl Method of Anu., zing Creep Collapse of Oval Cladding",

June,1975

6. CEN-83(N)-P, llillstone Unit 2 Reactor Operation With flodified CEA Guide Tubes", Februnry 8,1978

7. CEllPD-139, "C-E Fuel Evaluation Model Topical Report", July 1,1974

8. "liillstone flutlear Power Station Unit 2, Final Safety Analysis Report"

9. CEllPD-145-P, "I"CA fiethod of Analyzing In-core Detector Data in Power Reactors". April,1975

10. Letter from !!. R. Paradis to T. E. Short, " Omaha Public Power District Ft. Calhoun Station Unit tio.1", dated August 8,1970

11. W. R. Cadwell, "PDQ-7 Reference Manual", WAPD-TM-678, January 1978

12. CENPD-161-P, " TORC Code - A Computer Code for Determining the Themal liargin of a Reactor Core", July,1975

13. CENPD-162-P-A, "CE Critical Heat Flux", September,1976

14. CENPD-225, " Fuel and Poison Rod Bowing", October,1976

15. " Interim Safety Evaluation Report on Effects of Fuel Rod Bowing on Themal Hargin Calculations for Ligh? Water Reactors", Pages 21 and 26 (llRC Report) 4 g

7.0 TRAtlSIENT AriALYSIS The purpose of this section is to present the results of Northeast Utilities Millstone Point Unit 2, Cycle 3 safety analysis (other than LOCA) at 2700 MWt (hereaftar referred to as Stretch Power Operation). The Design Bases Events (DBEs) considered in the Stretch Power safety analyses are listed in Table 7-1. These events can be categorized in the following groups :

1. Anticipated Operational Occurrences (A00's) for which the Reactor Protective System (RPS) assures that no violation of the Specified Acceptable Fuel Design Limits (SAFDLs)will occur.

2. Anticipated Operational Occurrences for which initial steady state over-power margin must be maintained in order to assure no violation of the SAFDLs.

3. Postulated Accidents.

For all DBEs so indicated in Table 7-1, an explicit analysis was performed to determine the consequences of these events during Stretch Power operation. In addition, all transients were analyzed using a revised time to 90% insertion of the scram rods of 3.1 seconds to account for the new guide tube flow hole design on four Batch E assemblies to be located under CEA's. This is conservative in that the remaining 68 CEA insertion times are not expected to exceed the Cycle 2 Technical Specification value of 2.75 seconds. (See Section 1) A few cases were not reanalyzed (see Table 7-1). These cases are either eliminated by Technical Specification restrictions or it can be readily demonstrated that previous analyses at 2560 MWt remain conservative at 2700 tMT operation. These arguments are presented in the ensuing sections. Unless othenvise noted, the reference cycle in each case was the Cycle 2 analysis (Re f. 1 ) . In all cases, the NSSS code, CESEC, was used to simulate the NSSS transient response and TORC /CE-1was used to calculate the transient DNBR. The results of these analyses are provided in the following sections.

7.1 AtiTICIPATED OPERATIONAL OCCURRENCES FOR WHICH THE RPS ASSURES N0 VIOLATION OF SAFDLs The events in this category were analyzed for Stretch Power operation of Northeast Utilities' Millstone Point Unit 2, Cycle 3 to demonstrate that the SAFDLs and the primary coolant system pressure limit will not be exceeded. The results of the analyses are presented in ti 2 following sections. Protection against exceeding the SAFDLs will c'.ntinue to be assured by the Reactor Protection System (RPS) setpoints. The setpoints will be modified (as necessary) to include changes necessitated by the results of the Stretch Power analyses of these events. The methodology used in generation of the RPS trip setpoints are in accordance with those described in CENPD-199-P, "C-E Setpoint Methodology," (Reference 2) except where specifically noted otherwise.

TABLE 7-1 NORTHEAST UTILITIES MILLSTONE POINT UNIT 2, CYCLE 3 DESIGN BASIS EVENTS (DBEs) CONSIDERED IN STRETCH POWER ANALYSIS 7.1 Anticipated Operational Occurrences for which the RPS Assures no Analysis Violation of SAFDLs: Status 7.1.1 Control Element Assembly Withdrawal Reanalyzt.J 7.1. 2 Boron Dilution Reanalyzed 7.1.3 Startup of an Inactive Reactor Coolant Pump Not Analyzed 7.1.4 Excess Load Not Analyzed 7.1.5 Loss of Load Reanalyzed 7.1.6 Loss of Feedwater Flow Reanalyzed 7.1.7 Excess Heat Removal due to Feedwater Malfunction Not Reanalyzed 7.1.8 Reactor Coolant System Depressurization Reanalyzed 7.1.9 Loss of Coolant Flowl Reanalyzed 7.2 Anticipated Operational Occurrences which are Dependent on Initial Overpower Margin for Protection Against Violation of SAFDLs: 7.2.1 Loss of Coolant Flow Reanalyzed 7.2.2 Full Length CEA Drop Reanalyzed 7.2.3 Part Length CEA Drop Not Analyzed 7.2.4 Part length CEA Malpositioning Not Analyzed 7.2.5 Transients Resulting from Malfunction of One Steam Generator Reanalyzed 7.3 Postulated Accidents 7.3.1 CEA Ejection Reanalyzed 7.3.2 Steam Line Rupture Reanalyzed 7.3.3 Steam Generator Tube Rupture Reanalyzed 7.3.4 Seized Rotor Reanalyzed I Incident is discussed under Section 7.2

TABLE 7-2 MILLST0:iE POINT UNIT 2 CORE PARAMETERS ASSU:1ED IN THE SAFETY ANALYSES FOR STRETCH POWER Ref. Cycle 2 Cycle 3 Physics Parameters Units Values Values Total Planar Radial Peaking Factors For DNB Margin Analyses (Fr) Unrodded Region 1.440 1.598 Bank 7 Inserted 1 .5 50 1.806 For kw/ft Limit Analyses (F ) Unrodded Region 1 .540 1.584 Bank 7 Inserted 1 .6 60 1.822 Peak Augmentation Factor 'l.045 1.054 Moderator Temperature Coefficient 10~4 Ap/ F -2.5 +.5 -2.5* +.5 Shutdown Margin (Value Used in Zero %Ap -3.2 -3.2 Power SLB) Safety Parameters MWT 2611 2754 Power Level Maximum Steady State Core Inlet Temp F 544 551 Minimum Steady State RCS Pressure psia 2200 2200 6 133.7 Minimum Reactor Coolant Core Flow x10 lb/hr 134.9** (2200 psia, 551 F) ASI DNBR LCO Limit at Maximum I .21 .16 P Allowed Power Level

                                        % Insertion of                 25        25 Maximum CEA Insertion at Maximum Allowed Power (2700 MWt)                Group 7 Maximum Allowed Initial Peak                  kw/ft                16.0        16.0 Linear Heat Rate (DBEs Other
'chan LOCA)

Steady. State Linear Heat kw/ft 21.0 21.0 Rate to Fuel Centerline Melt

                                                                    -4 ap/ F

• Except Steam Lige Rupture event which .used an MTC of -2.2x10

** (2200 psia, 544 F)

7.1.1 CEA WITHDRAWAL EVENT The CEA withdrawal event was reanalyzed for MP-2 due to operation at stretch power for Cycle 3. As stated in CENPD-1991P (Reference 2), the CEA Withdrawal event initiated at rated themal power is one of the DBEs analyzed to detemine a bias factor used in establishing the TM/LP setpoints. This bias factor, along with conservative temperature, pressure, and power readings assures that the TM/LP trip prevents the DNBR from dropping below the SAFDL limits (DNBR = 1.19 based on CE-1 correlation) for a CEA Withdrawal event. Hence, this event was analyzed for Cycle 3 to generate the bias' term input to the TM/LP trip. The CEA Withdrawal transient may require protection against exceeding both the DNBR and fuel centerline melt (kw/ft) SAFD'.s. Depending on the initial conditions and the reactivity insertion rate cssociated with the CEA withdrawal, either the Variable High Power Level Themal Margin / Low Pressure (TM/LP) trip reacts to prevent exceeding the ONBR SAFDL. An approach to the kw/ft limit is terminated by either the Variable High Power Level trip or the Local Power Density trip. The zero power case was analyzed to demonstrate that SAFDLs are not exceeded. For the zero power case, a reactor trip, initiated by the variable high power trip at 25% (15% + 10% uncertainty) of rated thermal power, was assumed in the analysis. The key parameters for the cases analyzed are reactivity insertion rate due to rod motion and moderator temperature feedback effects, and initial axial power distribution. The Resistance Temperature Detector (RTD) response time is also important in detemining the pressure bias factor. The range of reactivity insertion rates considered in the analysis is given 'in Table 7.1.1-1,along with the values of other key parameters used in the analysis of this event. These parameters were chosen to produce the most severe rate of change of DNBR at the time a trip is encountered thereby producing the limiting case in tems of SAFDL protection requirements. The initial axial power shape and the corresponding scram worth versus insertion used in the analysis of both cases is a bottom peaked shape. This power distribution maximizes the time required to teminate the decrease in DNBR following a trip. The CEA Withdrawal transient initiated at 102% of 2700 MWt results' in a maximum pressure bias factor of 45.0 psia. This bias factor accounts for measurement systen processing delays during the CEA Withdrawal event. The pressure bias factor for this cycle has increased from the reference cycle due to the increase in the RTD time constant and the increase in the CEA drop time to 90% insertion. This pressure bias factor is used in generating TM/LP trip setpoint and will prevent the SAFDLs from being exceeded during a CEA Withdrawal event.

The zero power case initiated at the limiting conditions of operation results in a minimum CE-l DNBR of 1.58. Also, the analysis shows that the fuel-centerline temperatures are well below those corresponding to the fuel centerline melt SAFDL. The sequence of events for the zero power case is presented in Table 7.1.1-2 Figures 7.1.1-1 to 7.1.1-4 oresent the transient behavior of core power, core average heat flux, the coolant temperatures, and the RCS pressure. The analysis of the CEA Withdrawal event presented herein shows that the DNB and fuel centerline melt SAFDLs will not be exceeded during a CEA Withdrawal transient for Cycle 3.

TABLE 7.1.1-1 KEY PARAMETERS ASSUMED IN THE CEA WITHDRAWAL ANALYSIS Units fvele 2 , ' Cycle 3 Parameter Initial Core Power Level HWt 0-102% of 2560 0-102% of 2700 Core Inlet Coolant Temperaure F 532-544 532-551 Reactor Coolant System Pressure psia 2200 2200

                                          -4 Ap/ F              +5             +.5 Moderator Temperature Coefficient       10
                                                                  .85             85 Doppler Coefficient Multiplier CEA Worth at Trip - FP                  10~2 Ap                 -4.6           -4.32 CEA Worth at Trip - ZP                  10-2 Ap                 -3.2           -3.1
                                            -4 Ap/sec       0 to 2.0       0 to 1.62 Reactivity Insertion Rate               X10 Holding Coil Delay Time                 sec                      0.5            0.5 sec                      2.75           3.1 CEA Time to 90 Percent Insertion (Including Holding Coil Delay)

Resistance Temperature Detector sec 5.0* 8.0 Response Time

• Tech Spec change during Cycle 2 changed this value to 10 seconds t

TABLE 7.1.1-2 SEQUENCE OF EVENTS FOR CEA WITHDRAWAL FROM ZERO POWER Time (sec) Event Setpoint or Value 0.0 CEA Withdrawal Causes Uncontrolled ---- Reactivity Insertion 28.1 High Power Trip signal Generated 25% of 2700 MWt 28.5 Reactor Trip Breakers Open ---- 29.0 CEAs Begin to Drop Into Core ---- 29.3 Maximum Power Reached 141% of 2700 MWt 30.6 Maximum Heat Flux Reached 61.5% of 2700 MWt 30.6 Minimum CE-1 DNBR Occurs 1.58 32.7 Maximum Pressurizer Pressure 2358 psia Reached

7.1.2 BORON DILUTION EVENT The Boron Dilution event was reanalyzed for MP-2 due to operation at stretch power for Cycle 3. This event, as described in the FSAR, is analyzed to detemine that there is sufficient time available for an operator to terminate an approach to criticality and to establish correspond-ing shutdown margin requirements as prescribed by the Technical' Specifications. The Boron Dilution event, which is similar to slow CEA withdrawal, produces power and temperature increases which cause an approach to both the DNBR and kw/ft SAFDLs. Since the TM/LP trip system monitors the transient behavior of core power level and cene inlet temperature, the TM/LP trip assures that the DNBR SAFDL is not exceeded for power increases within th setting of the Variable High Fower Level trip. For power excursions in excess of the Variable High Power Level trip, a reactor trip is actuated. The approach to the kw/f t SAFDL is terminated by either the Local Power Density trip, Variable High Power Level trip, or the DNBR related trip discussed above. For a boron dilution initiated from hot zero power, critical, the pcuer transient resulting from the slow reactivity insertion rate is terminated by the Variable High Power Level trip prior to approaching the SAFDLs. Table 7.1.2-1 lists the key transient parameters assumed in each mode of operation in comparison to the reference cycle values. The conservative input data chosen consists of high critical boron concentrations and low inverse boron worths. Both of these choices produce the most adverse effects since they reduce the time to criticality. The time to criticality was determined by using the following e:.pression: C Initial ot crit =T BD C Crit where at crit = Time interval to dilute to critical T = Time constant BD C = Critical boron concentration (ppm) Crit C Initial Boron concentration (ppm) Initial The Boron Dilution event initiated at the maximum allowed power level is similar to a slow CEA Withdrawal event. At the maximum allowed power level,jheBoron Dilution event causes.a positive reactivity change of 4 fi.7x10 Ap/sec compared to the~ maximum CEA withdrawal rate of 1.25x10 Ap/sec. Hence, the CEA llithdrawal event is more limiting than the Boron Dilution event initiated ac the maximum allowed power level. Table 7.1.2-2 compares the results of the analysis for Cycle 3 with that for Cycle 2. The key results are the minimum times required to lose prescribed shutdown margin in each operational mode. As seen from the table, sufficient

time exists for tiie operator to initiate appropriate action to mitigate the consequences of this event.

TABLE 7.1.2-1 KEY PARAMETERS ASSUME 0 IN THE BORON DILUTION ANALYSIS Parameter Cycle 2 Cycle 3 Critical Boron Concentration, PPM (All Rods Out, Zero Xenon) Power Operation 1200 1300 Startup 1092 1400 Hot Standuy 1300 1400 Hot Shutdown 1300 1400 Cold Shutdown 1300 1400 Refueling 1200 1300 Inverse Boron Worth, PPM /% ap Power Operation 70 70 startup 65 65 Hot Standby 55 55 Hot Shutdown 55 55 Cold Shutdown 55 55 Refueling 55 55 Minimum Shutdown Margin Assumed, % ao Power Operation -- Startup -3.2 -3.2 Hot Standby -3.2 -3.2 Hot Shutdown -3.2 -3.2 Cold Shutdown -1.0 -1.0 Refueling -

                                                   -10.0            -10.0

TABLE 7.1.2-2 RESULTS OF BORON DILUTION EVENT Criterion-for Minimum Time to Lose Time to Lose Prescribed Shutdown Prescribed Shutdown Margin (Min) Margin (Min) Mode Cycle 2 Cycle 3 83 15 Startup 94 61 15 Hot Standby 68 68 61 15 Hot Shutdown 22 20 15 Cold Shutdown 56 44 30 Refueling 1

7.1.3 STARTUP OF AN INACTIVE REACTOR COOLANT PUMP EVENT The Startup of an Inactive Reactor Coolant Pump event was not analyzed for Cycle 3 because the Technical Specifications do not pemit operation with less than 4 Reactor Coolant pumps operating. i, e I

7.1.4 EXCESS LOAD EVEhT For the excess load event, operation at stretch power will reduce the power mismatch between core power and steam generator load demand for the full power case. Therefore, the excess load' incident analysis presented h the FSAR for 2560 Mwt operation conservatively bounds Cycle 3 operations at 2700 MWt. For this event, the Thermal Margin / Low Pressure trip, Variable High Power Trip and Local Power Density trips prevent exceeding the specified acceptable fuel design limits.

7.1.5 LOSS OF LOAD EVENT , The Loss of Load event was reanalyzed to demonstrate that the DNBR SAFDL and the RCS pressure upset limit of 110~ of 2500 psia are not exceeded during operation at stretch power for Cycle 3. The assumptions used to maximize RCS pressure.and minimize DNBR during the , transient are: a) The event is assumed to result from the sudden closure of the turbine stop valves without a simultaneous reactor trip. This set of assumptions causes the greatest reduction in the rate of heat removal frem the reactor coolant system and thus results in the most rapid approach to the DNBR SAFDL. b) The steam dump and bypass system, the pressurizer spray system, and the power operated pressurizer relief valves are assumed not to be operable. This maximizes the pressurizer pressure reached during the transient. The Loss of Load event was initiated at the conditions shown in Table 7.1.5-1 The combination of parameters,shown in Table 7.1.5-1 maximizes peak RCS pressure and causes the greatest decrease in the transient DNBR. As can be inferred from the table, the key parameters for this event are the initial RCS pressure, the moderator and fuel temperature coefficients of reactivity, and the initial axial power distribution. The methods used to analyze this event were identical to those described in the FSAR except TORC /CE-1, rather than COSM0/W-3, was used to calculate the DNBR. The initial RCS pressure is assumed to be at 2200 psia, the lowest operating pressure allowed by Technical Specifications. The low initial RCS pressure leads to the maximum value of the pressurizer pressure because it provides the greatest rate of increase of power, heat flux, and coolant temperatures. These, in turn, produce the maximum rate of change of pressurizer pressure at the time the high pressurizer pressure trip setpoint is reached. Consequently, this causes the maximum pressure and heat flux overshoot after trip. The initial core average axial power distribution for this analysis was assumed to be a botton peaked shape. This distribution is assumed because it er'aimizes the negative reactivity inserted during the initial portion of the scram following a reactor trip and maximizes the time required to mitigate the pressure and heat flux transient. The Moderator Temperature . Coefficient (MTC) of +.5x10-4 ap/0F was assumed in this analysis. This MTC in conjunction with the increasing coolant temperatures, enhances the rate of change of heat flux and the pressure at the time of reactor trip. .A Fuel Temperature Coefficient (FTC) corresponding to beginning of life conditicns was used in the analysis; since this FTC causes the least amount of negative reactivity change for mitigating the transient increases in the core heat flux and the pressure. The uncertainty on the FTC used in the analyses is shown in Table 7.1.5-1.

The Loss of Load event, initiated from the conditions given in Table 7.1.5-1, results in a high pressurizer pressure trip signal at 10.8 seconds. At 14.3 seconds, the primary pressure reaches its maximum value of 2552 psia. This compares to an FSAR value of 2520 psia. The increase in secondary pressure is limited by the opening of the main steam safety valves, which open at 6.8 seconds. The secondary pressure reaches its maximum value of 1055 psia at 18.6 seconds after initiation of the event. The minimum DNBR during the transient of 1.33 occurs at 13.3 seconds. Table 7.1.5-2 presents the sequence of events for this event. Figures 7.1.5-1 to 7.1.5-4 show the transient behavior of power, heat flux, RCS coolant temperatures, and the RCS pressure. The results of this analysis demonstrate that the consequences of the Loss of load event do not exceed the DNBR SAFDL or the upset pressure limit. i t

TABLE 7.1.5-1 KEY PARAMETERS ASSUMED IN THE LOSS OF LOA 0 ANALYSIS FSAR Cycle 3 Units Parameter 2611 2754 MWt Initial Core Power Level O 544 551 F Core Inlet Coolant Temperature 133.7 x106 lbm/hr 134.9 Core Coolant Flow 2200 psia 2250 Reactor Coolant System Pressure x10

                                            ~4 ap/ F    +.5          +.5 Moderator Temperature Coefficient 85            85 Doppler Coefficient Multiplier                                     -3.2
                                         %ep             -2.4 CEA Worth at Trip sec              3.0           3.1 Time to 90% insertion of Scram Rods                                                        Manual Reactor Regulating System           Operating Mode Manual Operating Mode Inoperative   Inoperative Steam Dump and Bypass System G

TABLE 7.1.5-2 Setpoint or Value Time (Sec) Event 0.0 Loss of Secondary Load Steam Generator Safety Valves 1000 psia 6.8 Open 2422 psia 10.8 High Pressurizer Pressure Trip Signal 11.2 CEAs Begin to Drop in o Core Pressurizer Safety Valves Open 2500~ psia 12.3 1.33 13.3 Minimum DNBR Occurs Maximum RCS Presst-re Reached 2555 psia 14.3 16.5 Pressurizer Safety Valves --- are Fully Closed Maximum Steam Generator 1055 psia 18.6 Pressure Reached . G e F

7.1.6 LOSS OF FEEDUATER FLOW EVENT The Loss of Feedwater Flow event was reanalyzed to ensure that the DNBR SAFDL and the RCS pressure do not exceed their respective limits during operation at stretch power for Cycle 3. Further, the event was analyzed to ensure that the water inventory remaining in the steam generators following trip is sufficient to provide at least ten minutes for the operator to initiate auxiliary feedwater. The analysis was performed assuming a reduction in feedwater flow to the steam generator without a corresponding reduction in steam flow. The result of the mismatch causes a reduction in the water inventory in the steam generators. The methodology used to analyze this event is the same as that described in the FSAR, except for the use of TORC /CE-1 in place of COSM0/W-3 to assess thermal (DNB) margin. The event is assumed to be initiated from the conditions shown in Table 7.1.6-1. The combinations of these parameters maximizes the RCS pressure and causes the greatest decrease in DNBR for the reasons given in the loss of Load Section (see Section 7.1.5). The initial conditions listed in Table 7.1.6-1 were used to analyze the event and to ensure that at least ten minutes exists before the operator must initiate auxiliary feedwater. The steam bypass and dump valves, the pressurizer spray system and the pressurizer relief valves were assumed to be in operation since this maximizes the steam flow from the steam generators and thus decreases the water inventory in the steam generator. The pressurizer relief vnives and the pressurizer spray system were assumed to be in normal operation such that a reactor trip on low steam generator water level is initiated prior to when a trip on high pressurizer pressure would be initiated. The event initiated from the conditions listed in Table 7.1.5-1 results in a high pressurizer pressure trip at 31.0 seconds. The pressurizer pressure reaches its maximum value of 2476 psia at 35.3 seconds. The minimum transient CE-1 DNBR during this event is equal to 1.33. The sequence of events for the limit-ing pressure and DNBR cases are given in Table 7.1.6-1. The transient behavior of core power, core average heat flux, the coolant temperatures, and the RCS pressure are presented in Figures 7.1.6-1 to 7.1.6-4. The Loss of Feedwater Flow event, analyzed to ensure that at least 10 minutes exists before the operator needs to initiate auxiliary feedwater, shows that a reactor trip on low steam generator level occurs at 16.5 seconds. This corresponds to a water level which is E0 inches below normal operating level. The analysis shows that the water inventory remaining in the steam generator following trip is sufficient to provide at least 15 minutes for the operator to initiate auxiliary feedwater. Figure 7.1.6-5 shows the water inventory

in the steam generator as a function of time.

p 9

The analysis of the complete Loss of Feedwater Flow reported herein demonstrates that DNBR SAFDLs and the RCS pressure upset limit will not be exceeded. Further, it demonstrates that the operator will have at least 10 minutes after initiation of the event to restore emergency feeddater flow by the actuation of the auxillary feedwater system. Controls, annunciators, and flow meters are available at the main control board to pennit the operator to take the necessaryaction. t 5 0

                                 ' TABLE 7.1.6-1 KEY PARAMETERS ASSUMED IN TiiE LOSS OF FEEDWATER FLOW ANALYSIS Parameter                          Units            FSAR          Cycle 3 2611          .2754 Initial Core Power Level                      MWt Inlet Coolant Temperature                       F           544             551 0         134.9          133.7 Core Mass Flow Rate                       x10 lbm/hr Reactor Coolant System Pressure           psia             2200           2200 psia             844             860 Steam Generator Pressure Moderator Temperature Coefficient          x10~4 Ap/ F     +0.5           +0.5 Doppler Coefficient Multiplier             ----             0.85            0.85 Reactor Regulating System                  Operating Mode   Manual        Manual Steam Dump and Bypass System               Operating Mode    Inoperative   Inoperative Feedwater Regulating Sys tem               Operating Mode    Incperative   Inoperative Auxiliary Feedwater Systems                Operating Mode   Manual        Manual Pressurizer Reliad Valves                 Operating Mode    Inoperative   Inoperative e

e a

TABLE 7.1.6-2 - SEQUENCE OF EVENTS FOR LOSS OF FEEDWATER FLOW Time (Sec) Event ~ Setpoint or Value 0.0 Loss of Normal Feedwater ---- 31.0 High Pressurizer Pressure Trip 2422 psia Signal Generated 32.4 CEAs Begin to Drop into Core ---- 33.9 Minimum PNBR Occurs 1.33 35.3 Maximum RCS Pressure 2476 psia 35.7 ' Steam Generator Safety 1000 psia Valves Begin to Open 40.0 Maximum Steam Generator Pressure 1064 psia 6 1

7.1.7 EXCESS HEAT REMOVAL DUE TO FEEDWATER MALFUNCTION EVENT For the excess heat removal due to feedwater malfunction event, operating at stretch power will reduce the mismatch between steam demand and the ' maximum possible feedwater supply for the full power case due to the ~ initially higher feedwater flow at Stretch Power. Consequently, the statement in the FSAR that the Excess Load Event is more adverse is still valid for Cycle 3 at 2700 MWt. For this event, the Thermal Margin / Low Pressure Trip, Variable High Power Trip and the Local Power Density trip prevent exceeding-the Specified Acceptable Fuel Design Limits. 0

7.1-8 RCS DEPRESSURIZATION EVENT The RCS Depressurization event was reanalyzed for MP-2 due to operation at stretch power for Cycle 3. As stated in CENPD-199-P (Ref. 2) this event is one of the DBEs analyzed to detennine a bias tenn input to the TM/LP trip. Hence, this event was analyzed for Cycle 3 to obtain a pressure bias factor. This bias fa: tor accounts for measurement system processing delays during this event. The trip setpoints which incorporate a bias factor at least this large will provide adequate protection to prevent the DNBR SAFDL from being exceeded during this event. The assumptions used to maximize the rate of pressure decrease and consequently the fastest approach to the DNBR SAFDLs are: 1) The event is assumed to occur due to an inadvertant opening of both This pressurizer relief values while operating at rated themal power. results in a rapid drop in the RCS pressure and consequently a rapid decrease in DNBR.

2) A bottom peaked initial axial power shape and corresponding scram worth versus insertion are used in the analysis. This power distribution maxi-mizes the time required to tenninate the decrease in DNDR following a trip,

3) The charging pumps, the pressurizer heaters and the pressurizer backup heaters are assumed to be inoperable. This maximizes the rate of pressure decrease and consequently maximizes the rate of approach to DNBR SAFDL.

The analysis of this event shows that a pressure bias factor of 35 psia is adequate. This is less than the bias required by the CEA Withdrawal event. Hence, the use of the pressure bias factor determined by the CEA Withdrawal event will prevent exceeding the SAFDLs during an RCS Depressurization event.

7.2 ANTICIPATED OPERATIONAL OCCURRENCES WHICH ARE DEPENDENT ON INITIAL OVERPOWER MARGIN FOR PROTECTION AGAINST VIOLATION OF SAFDLs The events in this category were analyzed for 2700 MWt operation of Northeast Utilities' Millstone Point Unit 2, Cycle 3, to determine the initial margins that must be available during operation within the Tech Spec LC0 limits such that SAFDLs and upset pressure limits will not be exceeded during any of these events. The initial overpower margin required to prevent the SAFDLs from being exceeded for any of these events was determined by analyzing these events for the initial conditions specified in Table 7-2. These conditions we e chosen to assure that enough initial overpower margin is available to prevent exceeding the SAFDLs for the most limiting A00 in this category. The method of generating Limiting Conditions for Opera-tion (LCO) is discussed in Reference 2. 4 6 0 e

7.2.1 LOSS OF COOLANT FLOW EVENT The Loss of Coolant Flow was reanalyzed for Millstone Point 2 to detennine the impact of operating at a higher power level and of changing the low flow trip setpoint system for Cycle 3 in comparison to Cycle 2. For Cycle 3, the core power will be increased from 2560 MWt to 2700 MWt,gthemaxigum allowable core inlet temperature will be increased from 544 F to 551 F. Also, a speed sensing system for determining coolaric flow rate will be applicable for the 4-Pump Loss of Flow event. As stated in the FSAR, Section 14.6.1, the design basis loss of flow events are a complete loss of flow to all four operating reactor coolant pumps, a loss of power to two coolant pumps, and a loss of power to one coolant pump. Because the loss of power to one co61 ant pump is covered by the more limiting loss of power to two coolant pumps event, the loss of power to one coolant pump was not analyzed separately. A reactor trip for the 4-pump loss of coolant flow design basis event is initiated by a low coolant flow rate signal as determined by the reactor coolant pump speed sensing system. (Reference 16) This signal is compared with the 4-pump flow speed sensing underflow fraction setpoint. Operating at 2700 MWt conditions, a reactor trip is assured at a flow rate greater than or equal to 91.5% of full 4-pump flow with a trip signal delay time for the speed sensing sys tem of 0.45 seconds. For a partial loss of flow (i.e., 2 pump), m reactor trip is initiated by a low coolant flow rate as detennined by a reduction in the square root of the sum of the steam generator hot to cold leg pressure drops. This signal is compared with the 4-pump / AP setpoint. A partial loss of flow initiated at full power initial conditions leads to a reactor trip with a flow rate greater than or equal to 89% of full flow with a trip signal delay time for this system of 0.65 seconds. Since the speed sensing system is only applicable to a 4-pump loss of flow and not a partial loss of flow, an explicit analysis was performed to determine which Loss of Flow Design Bases Event is limiting (i.e., 4-pump or 2-pump) in the following steps: A. The time-dependent core and individual loop flows and steam generator pressure drops are determined by using the COAST program (described in CENPD-98, Reference 3), which solves the conservation equations for mass flow and momentum. The general forcing functions for the fluid momentum equations consist of the pump torque values from the manufacturer's four quadrant curves, wherein the torque is related to the pump angular velocity and discharge rate. The flow coastdown tests for Millstone II have demonstrated that the C0AST code predictions are conservative.

B. The resultant flows are used as input to CESEC (described in CENPD-107, Reference 4), a digital computer code which simulates the NSSS response. This calculation has been perfomed to demonstrate that the Reactor Coolant System (RCS) pressure during the transient does not exceed the Upset Pressure Limit of 2750 psia (110% of design). C. Limiting axial power distributions are determined from a large sample (4 ,000) of possible distributions which are calculated using the QUIX code (described in Reference 2) as a function of axial shape index, core burnup, and CEA configuration. The limiting axial power shapes are those distributions that produce the lowest initial steady state power to a 1.19 CE-1 DNBR, at a given axial shape index, as predicted by the TORC code (described in Reference 5). After the limiting axial power distributions are determined, a consistent set of scram reactivity curves, corresponding to each shape index, is selected. These scram curves are produced by quasi-static OUIX calculations. Calculations have shown that the use of quasi-static reactivity calculations produce results which are conservative relative to space-time (i.e., scra' reactivity characteristics where delayed neutron effects are present). A conservative total available shutdown worth, including a stuck CEA worth allowance, and all uncertainties is also used in the analysis. D. The RCS flow coastdown, axial power distributions, and corresponding scram curves are input into STRIKIN-II (described in CENPD-135, Reference 6) to detemine the time dependent hot channel and core average heat flux distributions during the transient. The STRIKIN-II digital computer program solves the cne-dimensional radial cylindrical heat conduction equation for each axial node along the rod. The conduction model explicitly represents the clad, gas gap, and seven radial fuel nodes for each of 20 axial nodes. The core average power is obtained by solving the point kinetics equation using Doppler and moderator feedback and scram CEA reactivity versus time. The hot channel power is calculated by multiplying the core average power by normalized radial and axial peaking factors. For conservatism, the axial power distribu-tion is held constant in STRIKIN-II during the transient. In addition, for conservatism, credit for the heat flux decay is taken only when the initial minimum CE-1 DNBR is located in an axial region of the core where the scram rods have passed the axial node of minimum DNBR before the time of minimum DNBR is reached. The use of STRIKIN-II to calculate the absolute core average and hot channel heat flux distributions as a function of time is consistent-with the methodology utilized and 7.rproved by the NRC on other C-E plants operating at an increased coser level (e.g., Calvert Cliffs Unit 1 Reference 7).

E. TORC was used to calculate the minimum DNBR during the transient. This is consistent with the methods currently used by C-E to calculate the available and required margins. The TORC code and its application to these analyses were reviewed and epproved by NRC in Reference 7. The Loss of Coolant Flow produces a rapid approach to the DNBR SAFDL due to the rap'd decrease in the core coolant flow. Protection against exceed-ing the DNSR SAFDL for this transient is provided by the initial steady state thermal margin which is assured by maintaining the technical specifi-cations LCOs on DNBR margin and by the response of the RPS as mentioned above. The key transient parameters for the Loss c.f Flow events are shown in Table 7.2-1. The most important of these parameters are- the low flo,i trip setpoint, trip signal delay time, time to 93% insertion and flow coastdown. The characteristic flow coastdown for the 4 pump case is shown in Figure 7.2.1-1. The results of the analysis show that the 4-pump Loss of Flow is more limiting than the 2-pump case. This occurs because by the time the CEAs start to drop into the core, the flow is lower for the 4-pump loss of flow event than the 2-pump loss of flow event. Consequently, only the 4-pung loss of flow event will be presented. , Table 7.2.1-2 presents the NSSS and RPS responses during a four pump loss of flow initiated at the most negative shape index (Ip) allowed by the LCOs. The low flow trip setpoint is reached at 1.0 seconds and the scram rods start dropping into the core .95 seconds later. A minimum DNBR of 1.19 (CE-1) is reached at 2.3 seconds. Figures 7.2.1-2 to 7.2.1-5 present the core power, heat flux, RCS pressure, and core coolant temperatures as a function of time. Figure 7.2.1-6 presents a trace of hot channel DNBR vs time for the limiting case that is characterized by an Ip = .16. The operation of the RPS in conjunction with the Initial Overpower Margin maintianed by the LCOs in the Technical Specifications assure that the minimum DNBR will be greater than or equal to 1.19 for the Loss of Coolant Flow event. I m

                      .~m.- - - - -

TABLE 7.2.1-1 KEY PARAMETERS ASSUMED IN THE LOSS OF COOLANT FLOW AtlALYSIS Units Cycle 2 Cycle 3 Parameter 2611 2754 Initial Core Power Level (MWT) 544 551 Core Inlet Coolant Temperature ( F) 6 134.9 133.7 Core Mass Flow Rate (101tm/hr) (psia) 2209 2200 Reactor Coolant Sys tem Pressure 857 Steam Generator Pressure (psia) 861

                                                 ~4                              +.5 Moderator Temperature Coefficient       (10      Ap/F)   +.5 1.15                 1.00 Ocppler Coefficient Multiplier          -----
                                           % of 370,000 GPM 89.0                91.5 LFT Setpoint -4 pump LOF 89.0               89.0
                   -2 pump LOF sec                  .65                 .45 LFT Response Time - 4 pump LOF
                        - 2 pump LOF        sec                 .65                 .65 sec                0.4                0.5 CEA Holding Coil Delay sec                2.75               3.10 CEA Time to 90% insertion (Illcluding Holding Coil Delay)

(10'2 Ap) -4.98 -5.00 CEA Worth at Trip Total Radial Peaking ~1.428 1.630 Factor, F T, AR0 r Figure 7.2.1-1 Figure 4-Pump RCS Flow Coastdown 7.2.1-1 b e

TABLE 7.2.1-2 SEQUEtlCE OF EVEtlTS FOR LOSS OF FLOW Setpoint or Value Time (Sec) Event 0.0 Loss of Power to all Four Reactor Coolant Pumps 91.5% of 4-Pump 1.00 Low Flow Trip 1.45 Trip Breakers Open Shutdown, CEAs Begin to 1.95 Drop into Core 1.19 2.30 Minimum CE-1 DNBR Maximum RCS Pressure, psia 2301 6.20 O h

7.2. 2 FULL LENGTH CEA DROP EVENT The Full Length CEA Drop event was reanalyzed to determine the initial margins required to prevent the DNBR and the fuel centerline melt SAFDLs from being exceeded during operation at stretch power for Cycle 3. The methods used to analyze this event are identical to those discussed in Section 7.2 of Reference 1, except TORC /CE-1 was used instead of COSM0/W-3 to calculate the DNBRs. Table 7.2.3-1 lists the input parameters used for Cycle 3 and com to the reference cycle.

1. The most negative moderator and fuel temperature coefficients of reactivity. These reactivity coefficients are conservative because they maximize the power rise following the negative reactivity inser-tion. This produces the maximum power at the time of minimum DNBR.

2. Charging pump and proportional heater systems are This maximizes theassumed pressureto be drop inoperable during the transient.

during the event.

3. All other systems are assumed to be in manual mode.

The The event is initiated by insercing .08%Ap over a period of 1.0 second. maximum increase in planar radial peaking factors assumed for the dropped CEA is 16% in both the unrodded planes and rodded planes containing Bank 7.(see Table 7.2.3-3 shows the maximum post drop radial peaks for Table 7.2.3-3). stretch power operation. The axial power shape in the hot channel is assumed to remain unchanged and hence the fractional increase in the 3-D peak .for the maximum power is taken to be equal to the maximum increase in the Bank 7 olanar radial peaking factor of 1.16. Since there is no trip assumed, the peaks will stabilize at these asymptotic values after a few minutes as the secondary side continues to demand 102% power. Table 7.2.3-2 presents the sequence of events for the Full LengthThe CEA Drop transient event initiated at the conditions described in Table 7.2.3-1. behavior of key NSSS parameters are presented in Figures 7.2.3-1 to 7.2.3-4. The transient initiated at the most negative Ip (Ip = .16) at the maximum power level allowed by the LCOs results in a minimum CE-1 DNBR .of 1.21. 'A maximum allowable initial linear heat generation rate of 17.8 kw/ft could exist as an initial condition without exceeding the Specified Acceptable This amount of Fuel Design Limit (SAFDL) of 21 kw/ft during this transient. margin is automatically assured by setting the Linear Heat Rate LCO based on the allowable linear heat rate for LOCA. The Limiting Conditions for Operation assure that the SAFDLs will not be exceeded during a dropped CEA event.

TABLE 7.2. 2-1 KEY PARAMETERS ASSUMED IN THE FULL LENGTH CEA DROP ANALYSIS Parameter Units Cycle 2 Cycle 3 Initial Core Power Level MWt 2611 2754 8 544 551 Core Inlet Temperature F Reactor Coolant System Pressure psia 2200 2200 Core Mass Flow Rate x100 lbm/hr 134.9 133.7 Moderator Temperature Coefficient x10~4~Ap/ F -2.5 -2.5 Doppler Coefficient Multiplier -- 1.15 1.15

                                         % Insertion of Bank 7     25          25 CEA Insertion at Maximum Allowed Power Dropped CEA Worth                         %Ap                         .10       08 Maximum Allowed Power Axial Shape         Ip                          .21         .16 Index Radial Peaking Distortion Factor         Unrodded Region            1.17        1.16 Bank 7 Inserted Region     1.17        1.16

TABLE 7.2.2-2 SEQUENCE OF EVENTS FOR FULL LENGTH CEA DROP Time (Secl Event Setpoint or Value 0.0 CEA Begins to Drop into Core -- 1.0 CEA Reaches Full Inserted Position 100% Inserted 2.5 Core Power Level Reaches Minimum and 89% of 2700 Mwt Begins to Return to Power due to Reactivity Feedbacks 140.0 Reactor Coolant System Pressure Reaches 2161 psia a Mininum Value 140.0 Minimum DNBR is Reached 1.21 150.0 Core Power Returns to its Maximum Value 102% of 2700 Mwt O O . . _ - . . _ . . ~ . I

TABLE 7.2.2-3 FULL LENGTH CEA DROP RADIAL PEAK RESULTS Cycle 3 Post-Drop Radial Peaking Factors

• Value DNBR Limit (Fr )

Unrodded Region 1.85 2.09 Bank 7 Region Local Power Density Limit (Fzy) 1.84 Unrodded Bank 7 Region 2.11

    *Without uncertainties.

7.2.3 PART LENGTH CEA DROP _ The Part length CEAs have been removed; hence, this event was not analyzed. 7.2.4 PART LENGTH CEA MALPOSITIONING The Part length CEAs h u e been removed; hence, this event was not analyzed. 9 O

7.2.5 A00'S RESULTING FROM THE MALFUNCTION OF ONE STEAM GENERATOR The transients resulting from malfunction of one steam generator were analyzed for stretch power operation to determine the margin requirements The to prevent DNBR and fuel centerline melt SAFDLs from being exceeded. methods used to analyze these events are identical to those reported in Section 7.2.3 of Reference 1, expect TORC /CE-1 was need instead of COSM0/ W-3 to calculate the DNBR. The four events which affect a single steam generator are identified below:

1. Loss of Load to One Steam Generator

2. Excess Load to One Steam Generator

3. Loss of Feedwater to One Steam Generator

4. Excess Feedwater to One Steam Generator Of the four events described above, it has been determined that the Loss Hence, of Load to One Steam Generator transient is the limiting asymmetric event.

cnly the results of this transient are reported. The event is initiated by the inadvertent closure of a single main steam isolation valve. Upon the loss of load to the single steam generator, its pressure and temperature begin to increase and continue in this manner until the secondary safety valves open. The intact steam generator " picks up" the lost load , which causes its temperature and pressure to decrease. An asymmetry in the reactor inlet temperature occurs which is dominated by the contribution from the intact steam generator and causes the average reactor inlet temperature to decrease. In the presence of a negative moderator temperature coefficient,Thus, this prompts a positive reactivity change which induces a power increase. the most negative value of this coefficient is used in the analysis. The transient caused by this assumed sequence of events results in the greatest asymmetry in core inlet temperature distribution and the limiting DNBR case. The transient was initiated at the initial conditions given in Table 7.2.6-1. During this event, a reactor trip on low steam generator water level in the intact steam generator is initiated at 10.0 seconds. The transient initiated at the most negative Ip (Ip = .16) at maximum power level allowed by the LCOs results in a minimum CE-1 DNBR of 1.24 and the fuel centerline melt limit is not approached. The required thermal margin for this event is less thanTable that required for the more limiting Loss of Flow event (See section 7.2.1). 7.2.6-2 presents the sequence of events for this event. Based on the transient analysis, none of the transients which result from the malfunction of one steam generator will exceed the DNBR and fuel centerline melt SAFDLs. m

TABLE 7 2. 5-1 KEY PARAMETE.ss ASSUMED IN THE ANALYSIS OF LOSS OF LOAD TO ONE STEAM GENERATOR Cycle 2 Cycle 3 Units Parameter 2611 2754 MWt Initial Core Power O 544 551 F Initial Core Inlet Temperature 2200 2200 Initial Reactor Coolant System psia Pressure -4 U -2.5x10-4 -2.5x10 Moderator Temperature Coefficient Ap/ F 0.85 0.85 Doppler Coefficient Multiplier m

TABLE 7.2.5-2 SEQUENCE OF EVENTS FOR LOSS OF LOAD TO ONE STEAM GENERATOR Event Setpoint or Value Time (Sec) 0.0 Spurious Closure of a Single Main Steam -- Isolation Valve (Left Side) 0.0 Steam Flow from Unaffected Steam Generator Increases to Maintain Turbine Power (Right Side) 3.6 Safety Valves Open on Isolated Steam Generator 1010 psia (LeftSide) 10.0 Low Steam Generator Level Trip Signal -50 inches Generated (Left Side) (below normal water level) 11.0 CEAs Begin to Drop into Core -- 12.2 Maximum Steam Generator Pressure Occurs (Left 1055 psia Side) 12.3 Minimum Steam Generator Pressure Occurs 664 psia (RightSide) 12.5 Minimum DNBR Occurs 1.24 m e 9 O O

7.3 POSTULATED ACCIDENTS The events in this category were analyzed for stretch power operation of Unit 2, Cycle 3 to ensure acceptable consequences. , The following sections present the results of the analyses performed for stretch power operation. O e 9 4

7.3.1 CEA EJECTION EVENT The CEA Ejection event was analyzed at 2700 MWt operation for Cycle 3 to demonstrate that the criterion for clad damage is not exceeded during stretch power operation for Cycle 3. The reactivity-forced power transient was simulated by a digital computer program, CHIC-KIN (Reference 9), which simultaneously solves the one group neutron point kinetics equations together with the time and space dependent thermal and hydraulics equations for heat generation and transport within a single channel. The kinetics model incorporates the standard six-delay group representation along with explicit reactivity. contributions from:' (a) CEA motion, (b) Doppler effect, and (c) moderator density variations. By simulating the core average channel, the CHIC-KIN code computes the core average integrated energy output during the course of the transient. In the CEA Ejection event, the principal reactivity feedback mechanism affecting the power transient is the Doppler feedback. In the point kinetics approach, utilized in CHIC-KIN, a spatial Doppler weighting factor (K) accounts for the fact that tha Doppler feedback effect is a function of the spatial flux distribution. In order to represent the radial Doppler effect in a conservative manner, a space-time analysis was performed in which point kinetics calculations for various radial slices were compared with time-dependent, two-dimensional diffusion theory results obtained with a C-E modified version of the TWIGL code (Reference 10). The results of the space-time analysis have demonstrated that the use of the static (non-Doppler flattened) radial fuel rod peaking factor, as obtained from two-dimensional diffusion theory ca'culations, in conjunction with the average hot spot energy releases, yield energy increases that are conservatively large. Radial Doppler weighting factors obtained as a function of the ejected CEA worth are defined such that CHIC-KIN and TWIGL results give the same total core energy release. The average energy rise in the hottest fuel pellet is obtained from the follouing relttionship: AEg=[(P/A)H x AE Ave x K] - EHT (7.3-1)

                                                        ~

Where AE is the average core energy rise obtained from CHIC-KIN; (P/A)H Ave (the three-dimensional fuel rod peaking factor) is the ratio of the hot spot power density to the core average power density obtained from static, non-Doppler flattened diffusion theory ca-lculations; K is the reduction factor defined above. For the zero power case, it is conservatively assumed that EHT, which accounts for heat transferred out of the fuel rod during the transient, is zero. The average energy in the hottest fuel pellet at the beginning of the transient is added to the net average energy rise in the hottest fuel

pellet as obtained from Equation (7.3-1) to determine the total average enthalpy in the hottest fuel spot in the core. A similar procedure is used to compute the total centerline enthalpy in the hottest spot. The initial energy is obtained by correlating the initial local fuel temperature.with an empirical temperature-enthalpy relationship (see Reference 11). The spatial variation of the core local-to-average power ratio results from the convolution of the axial power distribution with radial pin power census distributions for the post-ejection condition, which are based on static core physics calculations. Combining these reruits with the tota.1 average and centerline enthalpies in the hottest fuel spot yields the fractional number of fuel rods with specific total average and centerline enthalpies. The calculated enthalpy values are compared to threshold enthalpy values to detemine the amount of fuel experiencing the various degrees of fuel damage. These threshold enthalpy values are (References 12, 13, and 14). Clad Damage Threshold: Total Average Enthalpy = 200 cal /gm Incipient Centerline Melting Threshold: Total Centerline Enthalpy = 250 cal /gm Fully Molten Centerline Threshold: Total Centerline Enthalpy = 310 cal /gm The criterion for detennining the fraction of fuel rods that will release their radioactive fission products during a CEA ejection is the same as the one quoted above for determining clad damage. Thus, it is assumed that any fuel rod that exceeds a total average enthalpy of 200 cal /gm releases all of its gap activity. The gap activity corresponding to the hottest fuel rod during the core cale is conservatively assumed for each rod that suffers clad damage. - To bound the most adverse conditions during the cycle, the most limiting of either the Beginning of Cycle (BOC) or End of Cycle (E0C) value was used in the analysis. A B0C Doppler defect was used since it produces the least amount of negative reactivity feedback to rgitigate the transient. A BOC moderator temperature coefficient of +0.5x10- Ap/ F was used which results in positive reactivity feedback with increasing coolant temperatures. An EOC Beta Fraction was used in the analysis to produce the highest power rise during the event. The zero power CEA ejection event was analyzed assuming the core is ~ initially operating at 1 Mut. At zera power, a Variable Overpower trip - is conservatively assumed to initiate at 25% (15% + 10% uncertainty) of 2700 MWt and terminates the event. - The full and zero power cases were analyzed, assuming a value of 0.05 seconds i for the total ejection time, which is consistent erith the FSAR. i Table 7.3.1-1 lists all the key parameters used in this analysis. 9

The power transient produced by a CEA ejection initiated at the maximum allowed power is shown in Figure 7.3.1-1, and at zero power is shown in 7.3.1-2. The results of the two CEA ejection cases analyzed (Table 7.3.1-2) show that the maximum total energy deposited during the events are less than the criterion for clad damage (i.e., 200 cal /gm). Consequently, no fuel pin failurcs occur. e 9 O

TABLE 7.3.1-1 CEA EJECTI0ti If1CIDENT ASSUMPTIOil5 Units Cycle 2 Cycle 3 Parameter Full Power MWt 2611 2754 Core Power Level

                                              -4 op/ UF           +.5           +.5 Moderator Temperature Coefficient         10
                                                                   .90          .92 K Factor
                                           %Ap                     .31           .29 Ejected CEA Worth
                                                                   .0048         .0047 Delayed Neutron Fraction, S 5.53           5.83 Post-Ejected 3-D Fuel Rod Power Peak *
                                            %Ap                    3.0           3.0 CEA Bank Worth at Trip 1.045         1.054 Augmentation Factor Zero Power MWt                     1.0           1.0 Core Power Level
                                                                    .84           .88 K Factor
                                             %Ap                    .74           .65 Ejected CEA Worth 18.6        14.5 Post-Ejected 3-D Fuel Rod Power Peak

• 1.47 1.47 CEA Bank Worth at Trip %Ap
• Includes all uncertainties, tilt allowance and waterhole peaking effects.

TABLE 7.3.1-2 CEA EJECTION RESULTS Cycle 2 Ref. Cycle 3 Val ue Value Full Power _ 197.7 190.8 Total Average Enthalpy of Hottest Fuel Pellet (cal /gm) 272.2 287.2 Total Centerline Enthalpy of Hottest Fuel Pellet (cal /gm) Fraction of Rods that Suffer Clad Damage (Average Enthalpy 0.0 0~. 0 >200 col /gm)

                                                                 .003        .021 Fraction of Fuel Having at least Incipient Centerline Melting (Centerline Enthalpy >250 cal /gm) 0.0        0.0 Fraction of Fuel Having a Fully Molten Centerline Condition (Centerline Enthalpy >310 cal /gm)

Zero Power 198.2 184.5 Total Average Enthalpy of Hottest Fuel Pellet (cal /gm) 198.2 184.5 Total Centerline Enthalpy of Hottest Fubl Pellet (cal /gm) 0.0 0.0 Fraction of Rods that Suffer Clad Damage (Average Enthalpy

>200 cal /gm) 0.0        0.0 Fraction of Fuel Having at least Incipient Centerline Melting (Centerline Enthalpy >250 cal /gm) 0.0       0.0 Fraction of Fuel Having a Fully Molten Centerline Condition (Centerline Enthalpy >310 cal /gm)

7.3.2 STEAM LINE RUPTURE EVENT The Steam Line Rupture event was analyzed for MP-2 due to operation at 2700 MWt for Cycle 3. As stated in the FSAR, the analysis was performed to demonstate that the critical heat flux is not exceeded during this event. The analysis assumed that the event is initiated by a circumferential rupture of a 34 inch (inside diameter) steam line at the steam generator - main steam line no zle. This break size is the most limiting, since this causes the greatest rate of temperature reduction in the reactor core region. The analysis of Steam LineThe Rupture event was performed with the two steam line rupture events methodology reported in the FSAR. considered during stretcn power operation were:

1. 2 loop full load - 2754 MWt case

2. 2 loop - no load case The one-loop full-load and one-loop no-load cases have not been analyzed since these operating conditions are not allowed by the Technical Specifications (paragraph 3.4.1).

Two-Loop - 2754 MNt The Two Loop - 2754 MWt case was initiated at the conditions listed in Table 7.3.2-1. The Moderator Temperature Coefficient (MTC) of reactivity assumed in the analysis ccrresponds to end of life, since this MTC results in the greatest positive reactivity change during the RCS cooldown caused by the Steam Line Rupture. Since the reactivity change associated with moderator feedback varies significantly over the moderatcr temperatures covered in the analysis, a curve of reactivity insertion versus temperature rather than 6 single value of MTC is assumed in the analysis. The moderator cooldown curve given in Figure 7.3.2-1, is based in a model which has an associated MTC of -2.2x10-4 Ap/ F at hot full power conditions. The reactivity defect 0 associated with fuel temperature decreases is also based on end of life Doppler defect. The Doppler defect based on an end of life Fuel Temperature Coefficient (FTC), in conjunction with the decreasing fuel temperatures, causes the greatest po3itive reactivity insertion during the Steam Line Rupture event. The uncertainty on the FTC assumed in the analysis is given

    'in Table 7.3.2-1. The 4 fraction assumed is the maximum absolute. value including uncertainties. This. too is conservatlye since it max 1mized the' positive reactivity insertion during cooldowns.

The minimum CEA worth assumed to be available for shutdown at the time of reactor trip at the maximum allowed power (2754 MWt), is 5.3%, assuming that the most reactive CEA is stuck in the fully withdrawn position during a scram. The analysis conservatively assumed that.the boron injected from the safety injection tank is worth -1.0%Ap per 90 ppm. Table 7.3.2-3 presents the sequence of events for the case initiated at the limiting conditions given in Table 7.3.2-1. The reactivity insertion as a

function of time is given in i tgure 7.3.2-2. As seen from the figure, the Steam Line Rupture event initiated at the maximum allowed power level (2754 MWt) remains at least .15%ap subcritical during the event compared to +.005%Ap for Cycle 2. The transient behavior of core power, heat flux RCS pressure, RCS temperatures, and steam generator pressere are presented in Figures 7.3.2-3 to 7.3.2-7. ' Two Loon - No Load The two-loop - no load case was initiated at the conditions given in Table 7.3.2-2. The moderator cooldown curve given in Figure 7.3.2-1 corresponds to an initial MTC of -2.2x10-4 Ap/0F. The end of life MTC was used for the reasons given in the two loop - 2754 MWt case. The FTC used in the analysis also corresponds to end of life for the reasons previously given for the two loop - 2754 MWt case. The minimum CEA shutdown worth available is conservatively assumed to be

  -3.2%Ap. A maximum inverse boron worth of 85 ppm /%Ap was conservatively assumed for the no load case for the safety injection.

The sequence of events for the twc-loop no load case is given in Table

7. 3. 2-4. The reactivity insertion as a function of time is given in Figure 7.3.2-8. The results of the analysis show a peak total reactivity of 0.21%Ap at 115.2 seconds for a period of 45.2 seconds, in canparison to

   +.18%Ap for Cycle 2. As a result of the low peak total reactivity and the short time duration of positive reactivity, there is no return to power during a Steam Line Rupture event initiated at zero power for stretch power operations. The transient behavior of core power, core heat flux, RCS pressure, RCS temperature, and steam generator pressure are presented in Figures 7.3.2-9 to 7.3.2-13.

The results of the full power Steamline Rupture indicate that the core remains subcritical by .15% Ap. For the zero power case, the core does not return to power. Hence, in both cases, the results confirm that the critical heat flux is not exceeded. h e e

TABLE 7.3.2-1 KEY PARAMETERS ASSUMED IN THE STEAll LINE RUPTURE ANALYSIS 2 LOOP - 2754 MWt Units Cycle 2 Cycle 3 Parameters MWt 2611 2754 Initial Core Power Level O F 544 551 Core Inlet Temperature psia 2200 2200 RCS Pressure Core Mass Flow Rate x106 lbm/hr 134.9 133.7 psia 812.2 860.4 Initial Steam Generator Pressure

                                               %Ap            -5.25              -5.31 CEA Worth at Trip                                                         .

1.0 1.15 Doppler Multiplier See Figure 7.3.2-1 See Figure 7.3.2-1 Moderator Cooldown Curve ppm /%Ap 65 90 Inverse Boron Worth x10-4 F/%Ap -2.2 -2.2 Effective MTC 8 fraction (including +10% unc.) .00583 .00576 e e e

TABLE 7.3.2-2 KEY PARAMETERS ASSUMED IN THE STEAM LItiE RUPTURE ANALYSIS 2 LOOP - NO LOAD Units Cycle 2 Cycle 3 Parameters MWt 0.0 0.0 Initial Core Power Level O 532 532 Core Inlet Temperature F psia 2200 2200 RCS Pressure Core Mass Flow Rate x106 lb/hr 134.9 133.7 psia 900 895 Initial Steam Generator Pressure CEA Worth at Trip %Ap -3.2 -3.2 Doppler Multiplier 'l:6 0.85 Figure 7.3.2-1 Figure 7.3.2-1 Moderator Cooldown Curve ppm /%Ap 65 85 Inverse Boron Worth x10-4 UF/%Ap -2.2 -2.2 Effective MTC 8 fraction (including +10% unc.) .00583 .00576 0 6 e e e

TABLE 7.3.2-3 SEQUENCE OF EVENTS FOR STEAM LINE RUPTURE 2 LOOP - 2'54 MWt Setpoint or Event Value Time (Sec) , 0.0 Steam Lin Rupture Occurs Low Steam Generator Pressure Signal 478 psia 3.4 4.8 CEAs Begin to Drop Into Core 10.7 Isolation Valves Closed 15.9 Pressurizer Empties 68.7 Steam Generator Blows Dry

TABLE 7.3.2-4 SEQUENCE OF EVENTS FOR STEAM LINE RUPTURE 2 LOOP - NO LOAD Setpoint or Event Value Time (Sec) , 0.0 Steam Line Rupture Occurs -- 3.7 Low Steam Generator Pressure Trip Signal 478 psia 5.1 CEAs Begin to Drop into Core -- 10.7 1 solation Valve Closed 12.4 Pressurizer Empties -- 111.8 Steam Generator Blows Dry --

7.3.3 STEAM GENERATOR TUBE RUPTURE EVENT The Steam Generator Tube Rupture (SGTR) event was reanalyzed due to operation at 2700 MWt for Cycle 3. The design basis SGTR is a double ended break of one steam generator U tube. Table 7.3.3-1 lists the key transient related parameters used in this analysis. In the analysis, it is assumed that the initial RCS pressure is as high as 2300 psia. This initial RCS pressure maximizes the amount of primary coolant transported to the steam system since the amount of leak is directly proportional to the difference between the primary and secondary pressures. Also, the higher pressure delays the low pressurizer pressure trip. For this event, the DNBR SAFDL is not exceeded due to the action of the Thermal Margin / Low Pressure trip which provides a reactor trip to maintain the DNBR above 1.19. Therefore, no fuel failure occurs during the transient. Inherent in the Thennal Margin / Low Pressure trip is the explicit calculation of the limiting radial and axial peaks, maximum inlet temperature, P.CS pressure, core power, and conservative CEA scram characteristics. The sequence of events for this transient is given in Table 7.3.3-2. Figures 7.3.3-1 through 7.3.3-6 present the transient behavior of core power, core heat flux, the RCS pressure, the RCS coolant temperatures, the steam generator pressure and the rupturco tube leak rate.

TABLE 7.3.3-1 KEY PARAMETERS ASSUMED IN THE STEAM GENERATOR TUBE RUPTURE EVENT Ref. Cycle 2 Cycle 3 Value Value Parameter Units 2611 2754 MWt Initial Core Power Level U 544 551 Core Inlet Temperature F 2250 2300 RCS Pressure psia Core Mass Flow Rate x106 lbm/hr. 134.9 133.7 812 860 Initial Steam Generator psia Pressure

                              %Ap                      -5.41      -4.32 CEA Worth at Trip x10-4 Ap                  -2.5      -2.5 Moderator Temperature Coefficient 1.15      1.15 Doppler Multiplier G

S O

s TABLE 7.3.3-2 3EQUENCE OF EVENTS FOR THE STEAM GENERATOR TUBE RUPTURE INCIDENT Event Setpoint or Value Time (Sec)_ 0.0 Tube Rupture Occurs 825.2 Pressurizer Empties -- Low Pressurizer Pressure Trip Condition 1728 psia 825.2 825.6 Dump Valves Open 825.6 Bypass Valves Open 826.6 CEA3 Begin to Drop into Core -- Maximum Steam Generator Pressure 901.4 psia 839.0 900.0 Dump Valves Close 913.2 Bypass Valve Closes 1800.0 Operator Initiates Appropriate Action and -- Begins Cooldown to 300 F a

7.3.4 SEIZED ROTOR EVENT The Seized Rotor event was analyzed at 2700 MWt for Cycle 3 to ensure that only a small fraction of fuel pins are predicted to fail during this event. The single reactor coolant pump shaft seizure is postulated to occur as a consequence of a mechanical failure. The effect of a single reactor coolant pump shaf t seizure is a rapid reduction,in the reactor coolant flow to the

                                                                    ~

three-pump value. A reactor trip for the seized rotor postulated accident is initiated by a low coolant flow rate as determined by a reduction in the sum of the steam generator hot to cold leg pressure drops. This signal is compared with a setpoint which is a function of the number of energized reactor coolant pumps. For a complete loss of flow at full power operating conditions a trip will be initiated when, or before, the flow rate drops to 89 percent of full flow. The conservative assumptions used in this analysis are listed in Table 7.3.4-1. These parameters were chosen to maximize the number of predicted fuel pin failures. The analysis was carried out in the following steps: A. Upon initiation of this transient, the core flow rate is assumed to drop immediately to the asymptotic three pump core flow value of 77.2% of four pump flow. For conservatism in the analysis, it is assumed that the flow at the core inlet is instantaneously reduced to the three pump core flow value. B. The resultant flow is used as input to CESEC, a digital computer code described in Reference 4 which simulates the NSSS to demonstrate that the reactor coolant system (RCS) pressure will remain below the upset limit of 2750 psia (110% of design). C. A set of limiting axial power distributions,-including a consistent set of scram reactivity curves, is selected using the same method as described in the Loss of Flow Section 7.2.1. D. The RCS flow, axial power distributions and corresponding scram curves are then input into STRIKIN II (Reference 6) for each limiting case to determine the Hot Channel and core average heat fluxes versus time for the transient. See Loss of Flow Section 7.2.1 for a description of STRIKIN II. E. TORC /CE-1 was used for calculation of the minimum DNBR for the transient. For each limiting ASI, the seized rotor transient is initiated at the Limiting Condition for Operation for determination . of the minimum DNBR. Although the transient results in a minimum DNBR that is below 1.19, this condition exists for only a very short period of time so that consequences for this postulated accident are acceptable. . 9

TABLE 7.3.4-1 KEY PARAMETERS ASSUMED IN SEIZED ROTOR ANALYSIS Cycle 2 Cycle 3 Parameter Units 2611 2754 MWt Initial Core Power Level 544 551 Core Inlet Coolant Temperature F Core Mass Flow Rate 106 lbm/hr 134.9 133.7 2200 2200 Reactor Coolant System Pressure psia 857 psia 861 Steam Generator Pressure x10~4 Ap/UF +.5 +.5 Moderator Temperature Coefficient

                                       --            .85             1.0 Doppler Coefficient Multiplier CEA Worth on Trip                 %Ap               4.98         -5.00

TABLE 7.3.4-2 StQUENCE OF EVENTS FOR SElZED ROTOR Setpoint or Value Time (Sec) Event 0.0 Seizure of One Reactor Coolant Pump Low Coolant Flow Trip 89% of 4-Pump Flow 0.0 0.04 Dump Valve Opens 67 Trip Breakers Open 1.17 Shutdown CEAs Begin Dropping into Core 3233 Maximum RCS Pressure 2276 9

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REFERENCES (Section7)

1. Hi11 stone 2 Cycle 2, Proposed License Amendment Related to Refueling Docket No. 50-336, September 1977.

2. CENPD-199-P, "C-E Setpoint Methodology," April 1976.

3. CENPD-98, "C0AST Code Descriptien." April 1973.

4. CENPD-107, "CESEC - Digital Simulation of a C-E Nuclear Steam Supply System," April 1974.

5. CENPD-161-P, " TORC Code - A Computer Code for Determining the Thermal Margin of a Reactor Core," July 1975.

6. CENPD-135-P, "STRIKIN II, A Cylindrical Geometry Fuel Rod Heat Transfer Program," August 1974.

7. Letter from A. E. Lundvall to D. Davis, July 13, 1977.

8. A) Letter from G. W. Gore to E. G. Case, December 1,1977. '

B) Letter fran A. E. Lundvall to E. G. Case, March 20, 1978. C) Letter fran A. E. Lundvall to E. G. Case, March 17, 1978.

9. WAPD-Th-479, J. A. Redfield, " CHIC-KIN - A Fortran Program for Intermediate and Fast Transients in a Water Moderator Reactor,"

January 1965,

10. WAPD-TM-743, J. B. Yasinsky, M. Natelson, and L. A. Hageman, "TWIGL - A Program to Solve the Two-Dimensional, Two Group, Space-Time Neutron Diffusion Equations with Temperature Feedback," February 1968.

11. CENPD-190A, "CEA Ejection, C-E Method for Control Element Assembly Ejection," July 1976.

12. GEMP-482, H. C. Brassfield, et. al., " Recommended Property and Reactor Kinetics Data for Use in Evaluating a Light Water-Cooled Reactor Loss-of-Coolant Incident Involving Zircaloy-4 or 304-SS Clad U02 ," April 1968.

13. Idaho Nuclear Corporation, Monthly Report, Ny-123-69, October 1969.

14. Idaho Nuclear Corporation, Monthly Report. Hai-127-70, March 1970.

15. CENPD-183, "C-E Methods for Loss of flow Analysis," July 1975. -

16. Letter from W. G. Counsil to R. Reid, " Millstone Nuclear Power Station, Unit No. 2, Reactor S olant Pump Speed Sensing System", Docket No. 50-336, Novenber 8, 1978.

8.0 TECHNICAL SPECIFICATION CHANGES The purpose of this Section is to provide an explanation for each of the changes proposed in support of plant operation at 2700 MWt. 8.1 The maximum thermal power level allowed by the license is increased to 2700 megawatts thermal consistent with NNECO's application for a power uprating beginning with Cycle 3 operation. 8.2 Superfluous license conditions D, E, and F are deleted. These conditions relate to the mini-flow bypass line, the neutron streaming phenomenon. and RCP flow test data. All love been or will be fulfilled prior to the start of cycle 3 operation. 8.3 The definition of RATED THERMAL POWER is revised to reflect the proposed power increase. 8.4 Revised Figures 2.1-1, 2.2-2, 2.2-3, 2.2-4, and B2.1-1 are provided. These new figures, which relate to the Thermal Margin Trip setpoints, are the result of the Cycle 3 setpoint analysis and are justified by the safety analyses provided. 8.5 Bases Pages B2-1, 32-3, B2-5, B2-6, and B2-7 are revised to reflect the use of the CE-1 correlation to relate reactor power, rcactor coolant tempera-ture and pressure to DNBR. The CE-1 correlation limits the calculated minicum DNBR to 1.19 which provides a 95/95 probability / confidence level that DNB will not occur. Previoun cycle analyses used the W-3 correlation which limited the minimum DNBR to 1.30. On Page B2-1, the reference to a reactor coolant inlet temperature of 570 F at the 1750 psia isobar is deleted since a 580*F temperature limit was the value used in the analyses which generate all the isobars shown on Figure 2.2-1. On Page B2-2, Cycle 3 specific values are provided. On Page B2-3, the reference to Table 2.1-1 has been corrected to Figure .221-1. On Page B2-7, the pressure measurement allowance and time delay allowance are revised for consistency with Cycle 3 safety analysis. 8.6 The CEA drop time limit is increased to 3.1 seconds due to the use of four fuel assemblics with reduced flow guide tubes. The drop time for the CEA's above these four assemblics may be greater than the Cycle 2 limit of 2.75 seconds and, therefore, a drop time of 3.10 seconds wa s conservatively chosen for the Cycle 3 safety analysis. The drop times for all other CEA's are ex-pected to show considerable margin to the 3.1 second limit. The CEA drop time for all CEA's will be demonstrated prior to Cycle 3 criticality, in

             ^

accordance with Technical Specifications. 8.7 The Power Lependent insertion Limit curve shown in Figure 3.1-2 was changed to maintain operating flexibility in developing the DNB and Linear Heat Rate LCO's.

8.8 On Pages 3/4 2-1 and 3/4 2-2, a revised treatment of linear heat rate is provided. These specifications are consistent with the safety analyses performed for Cycle 3 operation. 8.9 Figures 3.2-2, Excore Linear Heat Rate LCO and 4.2-1, Augmentation Factors, have been changed to reflect the Cycle 3 analyses. 8.10 The calculated value of Fxy, the total planar radial peaking factor is limited to 1.615. The fay curve, Figure 3.2-3a, is also modified as shown on Page 3/4 2-8a. the total integrated radial peaking factor ThecalculatedvalueofFf,fcurve, is limited to 1.630. The F Figure 3.2-3g,isalsomodifiedas shown on Page 3/4 2-8b. The limits on Fxy and Frensure that the assump-tions used in the safety analysis and setpoint generation remain valid. 8.11 The azimuthal power tilt specification has been modified to be applicable above 50% of rated thermal power. Meaningful measurements cannot be made at low power; therefore, the applicability statement has been revised to remove Mode 2 below 50% of rated thermal power. 8.12 Analyses have shown, as explained in Section 4.1, that the cladding collapse resistance of all fuel rods is sufficient to preclude collapse during their design lifetime. Therefore, a limitation on Cycle 3 burnup is not warranted and Section 3.2.5 and 4.2.5 have been deleted accordingly. The bases section is also deleted. 8.13 Table 3.2-1 is revised to reflect a maximum allowable cold leg temperature of 549'F, consistent with Cycle 3 safety analyses. 8.14 The DNB LCO, Figure 3.2-4, has been modified to reflect Cycle 3 safety analysis and ensures that the minimum DNBR limit of 1.19 is not exceeded. 8.15 Cycle 3 safety analyses have assumed an RTD time constant of 8.0 seconds. 8.16 Response time specifications have been included for the charging pumps and containment air recirculation system. Credit for flow from the charging pumps will be assumed in the small break LOCA analyses beginning with Cycle 3 operation. Assuming a loss of of fsite power, a response time of 40 seconds is required and is the value assumcd in the small break LOCA analyses for Cycle 3. This time interval can be justified by the following sequence of events. Time, Secends Action 0 Rupture 1 SIAS 21 Diesel Breaker Closes 29.4 Sequencer starts charging pumps 34.4 Pumps performing design function

The response time of 20.0 seconds, which assumes that offsite power is availabl., is conservative and provides assurance that charging pump flow will be available well before the required 40 seconds. Justification for the 31.0 second response time for the containment air recirculation system is provided in the W. G. Counsil letter to R. Reid dated October 11, 1978. The response time of 15 seconds, which assumes that offsite power is available is conservative and prevides assurance that the CAR fans will be performing their design function within the required 31.0 seconds. 8.17 The supplementary inspection program and related Tables 4.4-7 and 4.4-8 are deleted. The basis for deleting this section is that the supplementary inspection program was intended to be performed only once, i.e., during the upcoming outage inspection. As stated in Section 4.4.5.2.3, Frequency of Surveillance, "The supplementary inspection requirements described in this specification are to be implemented during the next Scheduled Inspec-tion". As described in the definition of Scheduled Inspection, this inspec-tion is to be performed not less than 12 nor more than 24 months after the May, 1977 Outage Inspection. Due to continued excellent condenser integrity and secondary water chemistry, operation of the full flow condensate polisher during Cycle 2, and the steam generator repairs performed last outage, NNECO anticipates a successful inspection program this outage. Therefore, steam generator specifications have been revised to reflect the provisions of Reg. Guide 1.83. Note also an editorial change has been made to Table 4.4-6. 8.18 Specification 4.4.5.1.4.a.8 has also been revised to reflect Reg. Guide 1.83 recommendations, and is consistent with current CE-STS. Cold leg side entry requirements are not required to adequa,tely inspect a steam generator tube. C.19 In the D. C. Switzer letter to G. Lear dated December 16, 1977, NNECO proposed to allow continued operation at specified reduced power levels with inoperable main steam line code safety valves. The subject reference pro-vided Table 3.7.0 which specified allowable fractions of rated thermal power for a given number of inoperable valves. Since the thermal power is proposed to be increased to 2700 MWt for Cycle 3, these fractions are inaccurate. Revised fractions corresponding to a thermal power level of 2700 MWt are provided. The above mentioned letter also proposed to revise Bases Page B3/4 7-1 to reficct 2560 MWt operation. Since uprating to 2700 MWt is proposed, the existing bases for this specification are correct. The values provided, therefore, supersede the information provided in the December 16, 1977 letter. G.20 Specification 3/4.7.8, Hydraulic Snubbers, is revised to eliminate Cycle 2-specific requirements. In addition, Specification 4.7.8.1 has been revised to eliminate wording applicable only to the initial snubber inspection.

8.21 On Page 3/410-1, clarification of an acceptable boron concentration is provided by the inclusion of ">" signs. 8.22 Bases Page B3/4 2-1 is revised for consistency with the specifications. 8.23 Bases Pages B3/4 4-2 and B3/4 4-2a are revised to delete Cycle 2 specific rcquirements.

PROPOSED CilANGES License License p. 3 - Maximum Power Level to 2700 MWt License p. 4 - Delete completed conditions D), E), and F) Technical Specifications P. 1-1 - Rated Thermal Power of 2700 MWt P. 2-2 - As shown. P. 2-7 - As shown. P. 2-8 - As shown. P. 2-9 - As shown. P. B2-1 - CE-1 correlation, DNBR of 1.19, eliminate 570*F restriction. P. B2-2 - As shown. P. B2-3 - Editorial, DNBR of 1.19. P. B2-5 - DNBR of 1.19. P. B2-6 - DNBR of 1.19. P. B2-7 - TM/LP changes. P. 3/4 1 CEA Drop Time. P. 3/4 1 As r,hown. P. 3/4 2 LilR ACTION statement. P. 3/4 2 Ae shown. P. 3/4 2 As shown. P. 3/4 2 As shown. P. 3/4 2-6 -F[y P. 3/4 2-8a - As shown. P. 3/4 2-8b - As shown. P. 3/4 2-9 -F[ P. 3/4 2 Applicability of Tq P. 3/4 2 Eliminate Fuel Residence Spec. P, 3/4 2 Cold Leg Temperature Increase. P. 3/4 2 As shown. P. 3/4 3 RTD response tima. P. 3/4 3 Add Response Times for Charging Pumps and CAR System. P. 3/4 3 Add Response Times for Charging Pumps and CAR System. P. 3/4 4 Delete Supplementary Inservice Inspection Program. P. 3/4 4-7a - Remove Cold Leg Side Entry Requirement. P. 3/4 4-7b through P. 3/4 4-7d - Delete Supplementary Inservice Inspection Program. P. 3/4 4-7f - Editorial. P. 3/4 4-7g and 3/4 4-7h - Delete Supplementary Inservice Inspection Program. P. 3/4 7 Revise previous TSCR, based on percentage of 2560 MWt. P. 3/4 7 Delete Cycle 2 Specific Snubber Surveillance and Cycle 1 wording. P. 3/4 7 Delete Cycle 2 Specific Snubber Surveillance. P. B3/4 2 Change Measurement - Calculational Uncertainty Factor to 1.07 P. B3/4 2 Delete Fuel Residence Bases P. B3/4 4-2 and P. B3/4 4-2a - Modity S/G Bases. P. 3/4 10 Addition of ">" sign.

(2) Pursuant to the Act and 10 CFR Part 70, to receive, possess and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended; (3) Pursuant to the Ace nd 10 CFR Parts 30, 40, and 70, to receive, possess and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4) Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument and equipment calibration or associated with radioactive apparatus or components; C. This amended license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter 1: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54, and 50.59 of Part 50, and Section 70.32 of Part 70; is Lubject to all applicable provisions of the Act and to the rules, regula-tions, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below: (1) thximum Power Level The licensees are authorized to operate the facility at steady state reactor core power levels not in excess of 2700 megawatts thermal provided that the preoperational test items identified in Enclosure 1 to this amendment have been completed in sequence. Enclosure 1 is an integral part of Amendment No. 4 to DPR-65. (2) Technical Specifications (2) The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 46, are hereby incorporated in the license. The licensee shall operate the facility in acecrdance with the Technical Specifica-tions. (3) Steam Generator Feedwater Flow When the stear -jnerator water temperature is above 212*F and the steam ger,~ tator water level falls below the feedwater sparger, feedwater flow shall be limited to 600 gpm. If feedwater is not reestablished within 15 minutes from the time that the steam generator water level falls below the feedwater sparger, feedwater flow shall be limited to 168 gpm.

I (4) Fire Protection , The licensee may proceed with and is required to complete the modifications identified in Section 3 of the NRC's Fire Protection Safety Evaluation on the facility dated September 19, 1978. These modifications shall be completed by the end of the ref ueling outage presently scheduled for summer,1980. The licensee is required to implement and maintain the administrative controls identified in Section 6 of the NRC's Fire Protection Safety Evaluation on the facility dated September 19, 1978. The administra-tive controls shall be in effect by December 31, 1978. , G. This amended license is ef fective as of its date of issuance and shall expire at midnight December 11, 2010. FOR THE NUCLEAR RECUlATOKY CGMISSIGt . f Roger S. B d, Acting D F DivisiwJ of Reactortcensing Office et Nuclear Reactor Regulatloc Date of Issuance: g{p 2 6 S75 -

 .-..y,                          _                    _ - . _ . _ _ _      . _ - - . . . _ . . _ _ _

March 31, 1976 1.0 DEFINITIONS DEFINED TERMS 1.1 The DEFINED TERMS of this section appear in capitalized type and are applicable throughout these Technical Specifications. THERMAL POWER 1.2 THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant. RATED THERMAL POWER 1.3 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 2700 MWt. l OPERATIONAL MODE 1.4 An OPERATIONAL MODE shall correspond to any one inclusive com-bination of core reactivity condition, power level and average reactor coolant temperature specified in Table 1.1. ACTION 1.5 ACTION shall be those additional requirements specified as corollary statements to each principle specification and shall be part of the specifications. OPERABLE - OPERABILITY 1.6 A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function (s). Implicit in this definition shall be the assumption that all necessary attendant instrumentation, controls, electric power, cooling or seal water, lubrication or other required auxiliary equipment is also OPERABLE. REPORTABLE OCCURRENCE 1.7 A REPORTABLE OCCURRENCE shall be any of those conditions specified as a reportable occurrence in Revision 4 of Regulatory Guide 1.16,

" Reporting of Operating Information - Appendix "A" Technical Specifications."

MILLSTONE - UNIT 2 1-1

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February 2, 1976 2.1 SAFETY LIMITS BASES 2.1.1 REACTOR CORE The restrictions of this safety limit prevent overheating of the fuel cladding and possible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel is prevented by maintaining the steady state peak linear heat rate at or less than 21 kw/ft. Centerline fuel melting will not occur for this peak linear heat rate. Overheating of the fuel cladd hg is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cledding surface temperature is slightly above the coolant saturation temperature. Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient. DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temper-ature and Pressure have been related to DNB through the CE-l correlation. l The CE-1 irs correlation has been develoDed to predict the Di4B flux and the location of DNB for axially uniform and non-uniform heat flux distri-butions. The local DNB heat flux ratio, DNBR, defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB. The minimum value of the DNBR during steady state operation, normal operational transients, and anticipated transients is limited to 1.19. l This value corresponds to a 95 percent probability at a 95 percent con-fidence level that DNB will not occur and is chosen as as appropriate margin to DNB for all operating conditions. The curves of Figure 2.1-1 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and maximum cold leg temperature vith four Reactor Coolant Pumps operating for which the minimum DNBR is no less than 1.19 for the family of axial shapes and corresponding l radial peaks shown in Figure B2.1-1. The limits in Figure 2.1-1 were calculated for react'or coolant inlet temperatures less than or equal to 580 F. The dashed line at 580 F coolant inlet j tt.,nperatures is not a safety limit; however, operation above 580 F is not possible because of the actuation of the main steam line safety valves which limit the maximum value of reactor inlet temperature. Reactor operation at THERMAL PulaER levels higher than 112% of RATED THERMAL MILLSTONE - UrHT 2 B 2-1

February ,l 1976

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                     'AO8lyjg,o U3MOgTVlxy NILLSTogg UNIT 2                   8 g,,

February 2, 1976 SAFETY LIMITS BASES POWER is prohibited by the high power level trip setpoint specified in Figure 2.1-1. The area of safe operation is below and to the left of l these lines. The conditions for the Thermal Margin Safety Limit curves in Figure 2.1-1 to be valid are shown on the figure. The reactor protective system in combination with the Limiting Conditions for Operation, is designed to prevent any anticipated com-bination of transient conditions for reactor coolant system temperature, pressure, and thermal power level that would result in a DNBR of less than 1.19 and preclude the existence of flow instabilities. 2.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the integrity of the Reactor Coolant System from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere. The reactor pressure vessel and pressurizer are designed to Section III of the ASME Code for Nuclear Power Plant Components which permits a maximum transient pressure of 110% (2750 psia) of design pressure. The Reactor Coolant System piping, valves and fittings, are designed to 4NSI B 31.7, Class I which permits a maximum transient pressure of 110% (2750 psia) of component design pressure. The Safety Limit of 2750 psia is therefore consistent with the design criteria and associated code require-ments. The entire Reactor Coolant System is hydrotested at 3125 psia to demonstrate integrity prior to initial operation. MILLSTONE - UNIT 2 B 2-3 I

LIMITING SAFETY SYSTEM SETTINGS BASES Reactor Coolant Flow-Low (Continued) operation of the reactor at reduced power if one or two reactor coolant punps are taken out of service. The low-flow trip setpoints and Allowable Values for the various reactor coolant pump combinations have been derived in consideration of instrument errors and response times of equipment involved to maintain the DNBR above 1.19 under normal operation l and expected transients. For reactor operation with only two or three reactor coolant pumps operating, the Reactor Coolant Flow-Low trip set-points, the Power Level-High trip setpoints, and the Thermal Margin / Low Pressure trip setpoints are automatically changed when the pump condition selector switch is manually set to the desired two- or three-pump position. Changing these trip setpoints during two and three pump operation prevents the minimum value of DNBR from going below 1.19 during l normal operational transients and anticipated transients when only two or three reactor coolant pumps are operating. Pressurizer Pressure-High - The Pressurizer Pressure-High trip, backed up by the pressurizer code safety valves and main steam line safety valves, provides reactor coolant system protection against overpressurization in the event of loss of load without reactor trip. This trip's setpoint is 100 psi below the nominal lift setting (2500 psia) of the pressurizer code safety valves and its concurrent operation with the power-operated relief valves avoids the undesirable operation of the pressurizer code safety vaives. Containment Pressure-High The Containment Pressure-High trip provides assurance that a reactor trip is initiated concurrently with a safety injection. The setpoint for this trip is identical to the safety injection setpoint. Steam Generator Presrure-Low The Steam Generator Pressure-Low trip provides protection against an excessive rate of heat extraction from the steam generators and subsequent cooldown of the reactor coolant. The setting of 500 psia is sufficiently below the full-load operating point of 815 psia so as not to interfere with normal operation, but still high enough to provide the required protection in the event of excessively high steam flow. This setting was used with an uncertainty factor of 1 22 psi in the accident analyses. MILLSTONE - UNIT 2 B 2-5 e 5

December 8, 1978 LIMITING SAFETY SYSTEM SETTINGS BASES Steam Generator Water Level - Low The Steam Generator Water Level-Low Trip provides core protection oy preventing operation with the steam generator water level below the minimum volume required for adequate heat removal capacity and as;ures that the design pressure of the reactor coolant system will not be exceeded. The specified setpoint provides allowance that there will be suf ficient water inventory in the steam generators at the time of trip to provide a margin of more than 10 minutes before auxiliary feedwater is required. Local Power Density-High The Local Power Der,sity-High trip, functioning from AXIAL SHAPE INDEX monitoring, is provided to ensure that the peak local power density in the fuel which corresponds to fuel centerline melting will not occur as a consequence of axial power maldistributions. A reactor trip is initiated whenever the AXIAL SHAPE INDEX exceeds the allowable limits of Figure 2.2-2. The AXIAL SHAPE INDEX is calculated from the upper and lower ex-core neutron detector channels. The calculated setpoints are generated as a function of THERMAL POWER level with the allowed CEA group positior, being inferred from the THERMAL POWER level. The trip is automatically bypassed below 15 percent power. The maximum AZIMUTHAL POWER TILT and maximum CEA misalignment permitted for continuous operation are assumed in generation of the setpoints. In addition, CEA group sequencing in accordance with the Specifications 3.1.3.5 and 3.1.3.6 is assumed. Finally, the maximum insertion of CEA banks which can occur during any anticipated opera-tional occurrence prior to a Power Level-High trip is assumed. Thermal Margin / Low Pressure The Thermal Margin / low Pressure trip is provided to prevent opera-tion when the DNBR is less than 1.19. l

                                                                        *~

MILLSTONE - UNIT 2 B 2 -6

April 19, 1978 LIMITING SAFETY SYSTEM SETTINGS BASES Thermal Margin / Low Pressure (Continued) The trip is initiated whenever 'he reactor coolant system pressure signal drops below either 1750 psia or a computed value as cescribed below, whichever is higher. The computed value is a function of the higher of LT power or neutron power, reactor inlet temperature, the number of reactor coolant pumps operating and the AXI AL SHAPE INDEX. The minimum value of reactor coolant flow rate, the maximum AZIM'JTH*L POWER TILT and the maximum CEA deviation permitted for continuous operation are assumed in the generation of this trip function. In addition, CEA group sequencing in accordance with Specifications 3.1.3.5 and 3.1.3.6 is assumed. Finally, the maximum insertion of CEA banks which can occur during any anticipated operational occurrente prior to a Power Level-High trip is assumed. The Thermal Margin / Low Pressure trip setpoints are derived from the core safety limits through application of appropriate allowances for equipment response time measurement uncertainties and processing error. A safety margin is provided which includes: an allowance of 55 of RATED THEP?.AL POWER to compensate for potential power measurement error; an allowance of 2*F to compensate for potential temperature measurement uncertainty; and a further allowance of 67 psi to compensate for pres- l sure measurement error, trip system processing error, and time delay associated with providing effective termination of the occurrence that exhibits the most rapid decrease in margin to the safety limit. The 67 psi allowance is made up of a 22 psi pressure measurement allowance and a 45 psi time delay allowance. Loss of Turbine A Loss of Turbine trip causes a direct reactor trip when operating above 15% of RATED THERMAL POWER. This trip provides turbine protection, reduces the severity of the ensuing transient and helps avoid the lifting of the main steam line safety valves during the ensuing transient, thus extending the service life of these valves. No credit was taken in the accident analyses for operation of this trip. Its functional capability at the specified trip setting is required to enhance the overall reliability of the Reactor Protection System. w. MILLSTONE - UNIT 2 B 2-7

Apri l 19, 1978 ' -~ ---- REACTIVITY CONTROL SYSTEMS CEA DROP TIME LIMITING CONDITION FOR OPERATION 3.1.3.4 The individual full length (shutdown and control) CEA drop time, from a fully withdrawn position, shall be 1 3.1 seconds from when l electrical power is interrupted to the CEA drive mechanism until the CEA reaches its 90 percent insertion position with:

a. T,yg > SI5*F, and

b. All reactor coolant pumps operating.

APPLICABILITY: MODE 3. ACTION:

a. With the drop time of any full length CEA determined to exceed the above limit, restore the CEA drop time to within the above limit prior to proceeding to PODE 1 er 2.

b. With the CEA drop times within limits but determined at less than full reactor coolant flow, operation raay proceed provided THEP. MAL POWER is restricted to less than or equal to the maximum THERMAL POWER level allowable for the reactor coolant pump combination operating at the time of CEA drop time determination.

SURVEILLANCE REQUIREMENTS 4 .1. 3. 4 Thf CEA drop time of full length CEAs shall be demonstrated through measurement prior to reactor criticality:

a. For all CEAs following each removal of the reactor vessel head,

b. For specifically affected individual CEAs following any main-tenance on or modification to the CEA drive system which could affect the drop time of those specific CEAs, and

c. At least once per 18 months.

ilLLSTONE - UNIT 2 3/4 1-26

                )5% (1.00, 7- 25) 1.00 P      0.90 G  m                i a  d                '

@ y 0.80 ' (0.73, 7- 46) en s M l 1 0.70 (0.65, 6- 25) c h i i s s , 5 0.60 ', ', (0.56, 6- 50)

                    '      LONG TERM
  @                 '    S   DY STATE h   0.50          ,

TRANSIENT INSERTION LIMIT 4 ' LIMIT' ' O.40 8 7: 8

                                    . SHORT TERM O                  {              I STEADY STATE D   0.30           '

INSERTION

                     '.                        LIMIT                 (0.20, 5- 60) h 0.20            -             -

0.10 (0.0, 4- 60) $ 0 = h MhTS j l l7 l l l l l l 5l l l l l l3 l l 0 20 40 60 80 100 0 20 40 60 80 100 0 20 40 60 80 100 l l 61 I I I I I4 l l l 0 20 40 60 80 100 0 20 40 60 80 100

                                                        % OF CEA INSERTION FIGURE 3.1-2 CEA Insertion Limits vs. THERMAL POWER with 4 Reactor Coolant Pumps Operating

April 19, 1978 3/4.2 POWER DISTRIBUTION LIMITS LINEAR HEAT RATE LIMITING CONDITION FOR OPERATION 3.2.1 The linear heat rate shall not exceed the limits shown on Figure 3.2-1. APPLICABILITY: MODE 1. ACTION: During operation with the linear heat rate being monitored by the Incore Detector Monitoring System, comply with the following ACTION: With the linear heat rate exceeding its limit, as indicated by four or more coincident incore channels, within 15 minutes initiate corrective action to reduce the linear heat rate to within the limits and either:

a. Restore the linear heat rate to within its limits within one hour, or

b. Be in at lee.st HOT STANDBY within the next 6 hours.

During operation with the linear heat rate being monitored by the Excore Detector Monitoring System, comply with the following ACTIONS: With the linear heat rate exceeding its limit, as indicated by the AXIAL SHAPE INDEX being outside of the power dependent limits on the Power Ratio Recorder and with the THERMAL POWER:

a. Above 100t of the allowable power level determined by the expres-sion (M)U4) in Specification 4.2.1.2.b, within 15 minutes i either restore the AXIAL SHAPE INDEX to within the limits of Figure 3.2-2 or reduce THERMAL POWER to < 100% of the allowable power level determined by the expression (M)(N) in l Speci fication 4.2.1.2.b.

b. < 100% of the allowable power level determined by the expression (71)(N) in Specification 4.g.l.2.b, either restore the [

AXIAL SHAPE INDEX to within the limits of Figure 3.2-2 within I hour from initially exceeding the linear heat rate limit or be in HOT STANDBY within the next 4 hours. SURVEILLANCE REQUIREMENTS 4.2.1.1 The linear heat rate shall be determined to be within its limits by continuously monitoring the core power distribution with either the excore detector monitoring system or with the incore detector monitoring system. MILLSTONE - UNIT 2 3/4 2-1 .

April 19, 1978. POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) 4.2.1.2 Excore Detector Monitoring System - The excore detector monitoring system may be used for monitoring the core power distribution by:

a. Verifying that the full length CEA's are withdrawn to and maintained at or beyond the Long-Term Steady-State Insertion Limit of Specification 3.1.3.6.

b. Verifying at le:st once per 31 days that the AXIAL SHAPE INDEX alarm setpoints are adjusted to within the limits shown on Figure 3.2-2.

c. Verifying at least once per 31 days that the AXIAL SHAPE INDEX 1s maintained within the allowable limits of Figure 3.2-2, where 100 percent of the allowable power represents the maximum THERMAL POWER allowed by the following expression:

MxN where

1. M is the maximum allowable THERMAL POWER level *or the existing Reactor Coolant Pump combination.

2. N is the maximum allowable fraction of RATED THERMAL POWER asdeterminedbytheFJy curve shown in Figure 3.2-3 of Specification 3.2.2.

4.2.1.3 Incore Detector Monitoring System - The incore detector monitoring system may be used for monitoring the core power distribution by verifying that the incore detector Local Power Density alarms:

a. Are adjusted to satisfy the requirements of the core power dis-tribution map which shall be updated at least once per 31 days.

b. Have their alarm setpoint adjusted to less than or equal to the limits shown on Figure 3.2-1 when the following factors are appropriately included in the setting of these alarms:

1. Flux peaking augmentation factors as shown in Figure 4.2-1,

2. A measurement-calculational uncertainty factor of 1.07, l

3. An engineering uncertainty factor of 1.03,

4. A linear heat rate uncerte.inty factor of 1.01 due to axial fuel densification and thermal expansion,

5. A THERMAL POWER measurement uncertainty factor cf 1.02.

MILLSTONE - l Jilt 2 3/4 2-2

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         -0.6                                           -0.4                                                    -0.2                                              -0.0                                                       0.2                                                      0.4                                                0.6 AXIAL SHAPE INDEX FIGURE 3.2-2 AXIAL SHAPE INDEX vs. Fraction of Allowable Power Level per Specification *.2.1.2b MILLSTONE - UNIT 2                                                                                                                                                                   3/4 2-4

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a s a a s e S'dOIDVd Nb! WIN 3HDflV MILLSTONE - UNIT 2 3/4 2-5

April 19, 1978 POWER DISTRIBUTION LIMITS TOTAL PLANAR RADIAL PEAKING FACTOR - Th LIMITING CONDITION FOR OPERATION 3.2.2 The calculated value of F ,definedasFfy=F,y(1+T),shallbe q limited to < l .615. l APPLICABILITY: MODE 1*. ACTION: With Ffy > l.615 within 6 hours either:

a. Reduce THERMAL POWER to t. ring the combination of THERMAL POWER andFfy to within the limits of Figure 3.2-3 and withdraw the full length CEAs to or beyond the Long Tem Steady State Inser-tion Limits of Specification 3.1.3.6; or

b. Be in at least HOT STANDBY.

SURVEILLANCE REQUIREMENTS 4.2.2.1 The provisions of Specification 4.0.4 are not applicable. T 4.2.2.2 F,, shall be calculated by the expression F ,=F,y(1+T)andFfy q shall be detemined to be within its limit at the following intervals:

a. Prior to operation above 70 percent of RATED THERMAL POWER

            ,after each fuel loading,

b. At least once per 31 days of accumulated operation in MODE 1, and

, c. Within four hours if the AZIMUTHAL POWER TILT (T q ) is > 0.02. See Special Test Exception 3.10.2. MILLSTONE - UNIT 2 3/4 2-6

x - 5 G H UNACCEPTABLE o ' OPERATION 5 (1.615, 1.0) (1.656, 0.975) REGION

1.0  :

E - M (1.712, 0.8725) s 5 N k a 0.8 , (1.776, 0.8025) m a O W 0*6 H ACCEPTABLE .  ! w N OPERATION l ~ 2 g REGION ~ z E 3 0.4 , t3 2 m 0.2 , B  : j , . - . , 0.0 1.60 1.62 1.64 1.66 1.68 1.70 1.72 1.74 1.76 1.78 Ff: x [Fxy x (1 + Tq)] FIGURE 3.2-3a Total Planar Radial Peaking Factor versus Allowable Fraction of RATED THERMAL POWER

3 r E U:%CCEPTABLE o OPER TION REGION 8 (1.630, 1.0) h 1.0 H ~ g (1.659, 0.86) I

r R (1.644, 0.90)

                                   -                                                 (1.717, 0.755) a  0.8          -   -   --

h (1.689, 0.80) N (1.776, 0.675) a g 0.6 - w I N w ~ & $ .j. ECUTEE v_ 2- ' 0*4 y

1. i OPERATION  ; j , '

o ' l  ! _! i REGION  ; ,..t._.._, '. . - g

                                                                                     ~

N  :  ; U g 0.2 - B - d 0.0 1.60 1.62 1.64 1.66 1.68 1.70 1.72 1.74 1.76 1.78 F[: [Fr x (1 + Tq)} FIGURE 3.2-3b i Total Integrated Radial Peaking Factor Versus Allowable Fraction of RATED THERMAL POWER 4

April 19, 1978 POWER DISTRIBUTION LIMITS TOTALINTEGRATEDRADIALPEAKINGFACTOR-Ff LIMITING CONDITION FOR OPERATION T 3.2.3 The calculated value of F r defined as Ff = Fr (1+Tq ), shall be limited to 1 1.630. APPLICABILITY: MODE 1*. ACTION: Witt, Ff > 1.630, within 6 hours either:

a. Reduce THERMAL POLFR to bring the combination of THERMAL POWER and FT to within the limits of Figure 3.2-3 and withdraw the 7

full length CEAs to or beyond the Long Term Steady State Insertion Limits of Specification 3.1.3.6; or

b. Be in at least HOT STANDBY.

SURVEILLANCE REQUIREMENTS 4.2.3.1 The provisions of Specification 4.0.4 are not applicable. T 4.2.3.2 F T 7 shallbecalculatedbytheexpressionFf=F(1+T)andF r q y shall be detemined to be within its limit at the following intervals:

c. Prior to operation above 70 percent of RATED THERMAL POWER after each fuel loading,

b. At least once per 31 days of accumulated operation in MODE 1, and

c. Within four hours if the AZIMUTHAL POWER TILT q(T ) is > 0.020.

T 4,2.3.3 F shall be detennined each time a calculation of F7 is required by 7 using the incore detectors to obtain a power distribution map with all full length CEAs at or above the Long Term Steady State Insertion Lirhit for the existing Reactor Coolant Pump Combination. T 4.2.3.4 Tq shall be determined each time a calculation of F' is required T and the value of Tqused to determine Frshall be the measured value of Tq.

• See Special Test Exception 3.10.2.

MILLSTONE - UNIT 2 3/4 2-9

                                                                                              ~

April 19, 1978 POWER DISTRIBUTION LIMITS AZIMUTHAL POWER TILT - T q LIMITING CONDITION FOR OPERATION 3.2.4 The AZIMUTHAL POWER TILTq (T ) shall not exceed 0.02. APPLICABILITY: MODE 1, above 50% of RATED THERMAL F0WER*. l ACTION:

a. With the indicated AZIMUTHAL POWER TILT determined to be > 0.02 but < 0.10, either correct the power tilt within two hours or determine within the next 2 hours and at least once per sub-sequent 8 hours,thattheTOTALPLANARRADIALPEAKINGFACTOR(Ffy) and the TOTAL INTEGRATED RADIAL PEAKING FACTOR r (F ) are within the limits of Specifications 3.2.2 and 3.2.3.

b. With the indicated AZIMUTHAL POWER TILT cietermined to be > 0.10, operation may proceed for up to 2 hours provided that the TOTAL T

INTEGRATED RADIAL PEAKING FACTOR (F 7 ) and TOTAL PLANAR RADIAL PEAKING FACTOR (F y) are within the limits of Specifications 3.2.2 and 3.2.3. Subsequent operation for the purpose of measurement and to identify the cause of the tilt is allowable provided the THERMAL POWER level is restricted to < 20% of the maxinum allowable THERMAL POWER level for the existing Reactor Coolant Pump combination. SURVEILLANCE REQUIREMENT 4.2.4.1 The provisions of Specification 4.0.4 are not applicable. 4.2.4.2. The AZIMUTHAL POWER TILT shall be determined to be within the limit by: l

a. Calculating the tilt at least once per 7 days when the Channel High Deviation Alarm is OPERABLE, See Special Test Exception 3.10.2.

MILLSTONE - UNIT 2 3/4 2-10 l32

THIS PAGE INTENTIONALLY LEFT BLANK MILLSTONL - LNIT 2 3/4 2-12

April 19, 1978 TABLE 3.2-1 DNB MARGIN LIMITS Four Reactor Coolant Pumps Parameter __ Operating Cold Leg Temperature 1 549 F  ! Pressurizer Pressure 1,2225 psia

• Reactor Coolant Flow Rate 3,370,000 gpm AXI AL SHAPE INDEX Figure 3.2-4
• Limit not applicable during either a THERMAL POWER ramp increase in )

excess of Si of RATED THERMAL POWER per minute or a THERMAL POWER step increase of greater than 10% of RATED THERMAL POWER. 6 MILLSTONE - UNIT 2 3/4 2-14 l

1.2 1.1 (-0.10, 1.0) (0.15, 1.0) 1.0 UNACCEPTABLE g UNACCEPTABLE OPERATION a OPERATION REGION REGION O 0.9 d a ll (-0.30, 0.80) (0.30, 0.80) @ ACCEPTABLE H OPERATION I REGION b g 0.7 p U N w 0.6 - 0.5 0.4

       -0.6     -0.4      -0.2       0.0       0.2      0.4        0.6 Y1 FIGURE 3.2-4 AXIAL SHAPE INDEX Operating Limits With 4 Reactor Coolant Pumps Operating MILLSTONE - UNIT 2                   3/4 2-15

j , ' b December 28, 1978 TAlllE 3.3-2 x

       ;:                               REACTOR PROTECTIVE INSTRUt1F_NTA_TIO_N_R.E_SPONSE TIMES 8

N TUllCTIONAL UNIT RESPONSE Tit 1E b 1. Manual Reactor Trip < 2.0 seconds d 2. Power Level - High < 0.40 seconds * # and ; 8.0 seconds.' ll37

3. Reactor Coolant Flow - Low < 0.65 seconds

4. Pressurizer Pressure - High < 0.90 seconds

5. Containment Pressure - High Not Applicable

6. Steam Generator Pressure - Low < 0,90 seconds w 7. Steam Generator Water Level - Low < 0.90 seconds

8. Local Power Density - High < 0.40 seconds *# and 1 8.0 seconds #8 37

     &       9. Thermal Margin / Low Pressure                                 < 0.90 seconds *# and c 8.0 secordsdf

10. Loss of Turbine--Hydraulic -

Fluid Pressure - Low Not Applicable

• Neutron detectors are exempt from response time testing. Response time of the neutron flux signal portion of the channel shall be measured from detector output or input of first electronic ce"renent in c!onnel .

tResponse time does not include contribution of RTDs. FFRTO response time only. This value is equivalent to the time interval required for the RID's cuf put f n achieve 63.2% of its total change when subjected to a step change in RTD temperaturn.

December 8, 1978 TABLE 3.3-5 ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECON;5

1. Manual ,

a. SIAS Safety Injection (ECCS) Not Applicable Containment Isolation Not Applicable Enclosure Building Filtration System Not Applicable

b. CSAS Containment Spray Not Applicable

c. CIAS Containment Isolation Not Applicable

d. SRAS Containment Sump Recirculation Not Applicable

e. EBFAS Enclosure Building Filtration System Not Applicable

2. Pressurizer Pressure-Low

a. ECCS

1) High Pressure Safety Injection < 30.0*/5.0**

2) Low Pressure Safety Injection < 50.0*/5.0**

3 Charging Pumps < 40.0*/20.0** 4)) Containment Air Fecirculation System < 31.0*/15.0**

b. Containment Isolation < 7.5 Enclosure Building F iltration System

   ~

c. < 50.0*/35.0**

MILLSTONE - UNIT 2 3/4 3-21

December 8, 1978 TABLE 3.3-5 (Continued) ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECC' IDS

3. Containment Pressure-H;ph

a. Safety Injection (ECCS)

1) High Pressure Safety Injection -< 30.0*/5.0**

2) Low Pressure Safety Injection < 50.0*/5.0**

Charging Pumps

3) < 40

4) Containment Air ibcirculation System E 31 0*/T5.0**

0*/20.0**

b. Containment Isolation < 7.5

c. Enclosure Building F iltration System < 50.0*/35.0**

4 Containment Pressure--High-High

a. Containment Spray 1 35.6*II)/35.6**(I)

5. Containnent Radiation-Hich

a. Containment Purge Valves Isolation < Counting period plus 7.5

6. Steam Generator Pressure-Low

a. Main Steam Isciation 1 6.9

b. Feedwater Isolation 1 60

7. Refueling Water Storage Tank-Low

a. Containment Sump Recirculation 1 120 TABLE NOTATION Diesel generator starting and sequence loading delays included.

Diesel generator starting and sequence loading delays not included. Offsite power available. II) Header fill time not included.

                                                                          *~

MILLSTONE - UNIT 2 3/4 3-22

April 14, 1978 REACTOR COOLANT SYSTEM STEAM GENERATORS LIMITING CONDITION FOR OPERATION 3.4.5 Each steam generator shall be OPERABLE. APPLICABILITY: MODES 1, 2 and 3. ACTION: With one or more steam generators inoperable, restore the inoperable gen'.:rator(s) to OPERABLE status prior to increasing T,yg above '.00 F. SURVEILLANCE REQUIREMENTS 4.4.5.0 Each steam generator shall be demonstrated OPERABLE by perfor-mance of the following Augmented Inservice Inspection Program, 4.4.5.1 Augmented Inservice Inspection Program 4.4.5.1.1 S_t_eam Generator Sample Selection and Inspection - Each steam generator shall be detennined OPERABLE during shutdown by selecting and inspecting at least the minimum number of steam generators specified in Table 4.4-5. 4.4.5.1.2 Steam Generator Tube Sample Selection and Inspection - The steam generator tube minimum sample size, inspection result classifica-tion, and the corresponding action required shall be as specified in TaV e 4.4-6. The inservice inspection of steam generator tubes shall be perfonned at the frequencies specified in Specification 4.4.5.1.3 and the inspected tubes shall be verified acceptable per the acceptance criteria of Specification 4.4.5.1.4. The tubes selected for each inservice inspection shall include at least 3% of the total number of tubes in all steam generators; the tubes selected for these inspections shall be selected on a random basis except:

a. Where experience in similar plants with similar water chemistry indicates critical areas to be inspected, then at least 50%

of the tubas inspected shall be from these critical areas. t b. The first . ample of tubes selected for each inservice inspec-tion (subt quent to the preservice inspection) of each steam generator sr.all include: MILLSTONE - UNIT 2 3/4 4-5 I

April 14, 1978 REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 4.4.5.1.4 Acceptance Criteria l31

a. As used in this Specification

1. Imperfection means an exception to the dimensions, finish or contour of a tube from that required by fabrication drawings or specifications. Eddy-current testing indica-tions below 20% of the nominal tube wall thickness, if detectable, may be considered as imperfections.

2. Degradation means a service-induced cracking, wastage, wear or general corrosion occurring on eith< - inside or outside of 1 tube.

3. Degraded Tube means a tube containing imperfections >20%

of the nominal wall thickness caused by degradation. _

4.  % Degradation means the percentage of the tube wall thick-ness affected or removed by degradation.

5. sefect mearis an imperfection of such severity that it exceeds the plugging limit. A tube containing a defect is defective.

6. Plugging Limit means the imperfection depth at or beyond which the tuce shall be removed from service because it may become unserviceable prior to the next inspection and is equal to 40% of the nominal tube wall thickness.

7. Unserviceable describes the condition of a tube if it leaks or contains a defect large enough to affect its structural integrity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break as specified in 4.4.5.1.3.c, above. l31

8. Tube Inspection means an inspection of the steam generator tube from the point of entry (hot leg side) completel drQUnd Ine d-Deno T.o the top support of tne cold leg.y 31

b. The steam generator shall be determined OPERABLE after completing the corresponding actions (plug all tubes exceeding the plugging limit and all tubes containing through-wall cracks) required by Table 4.4-6.

MILLSTONE - UNIT 2 3/4 4-7a l

April 14, 1978 REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 4.4.5.1.5 Reports

a. Following each inservice inspection of steam generator tubes, the number of tubes plugged in each steam generator shall be reported to the Comission within 15 days.

b. The complete results of the steam generator tube inservice inspection shall be included in the Annual Operating Report for the period in which this inspection was completed. This report-shall include:

1. Number and extent of tubes inspected.

2. Location and percent of wall-thickness penetration for each indication of an imperfection.

3. Identification of tubes plugged.

c. Results of steam generator tube inspections which fall into Category C-3 and require prompt notification of the Comission shall be reported pursuant to Specification 6.9.1 prior to resumptiort of plant oper ation. The written followup of this report shall provide a otscription of investigations conducted to determine cause of the tube degradation and corrective l

measures taken to prevent recurrence. MILLSTONE - UNIT 2 3/4 4-7b i l

THIS PAGE INTENTIONALLY LEFT BLNiK MILLSTONE - UNIT 2 3/4 4-7c

THIS PAGE INTENTIONALLY LEFT BLANK MILLSTONE - UNIT 2 3/4 4-7d

x T ABLE 4.4 6 E r-o STEAM GENERATOR TUBE INSPECTION 2 IST SAMPLE INSPECTION 2ND SAMPLE INSPECTION 3RD SAMPLE INSPECTION c Sample Site Result Action Requircel Result Action Required Result Action Required H A minimum of C-1 None N/A N/A N/A N/A N S Tubes per

• S. G.

C-2 Plug defective tubes C-1 None N/A N/A and inspect additonal Plug defective tubes C-1 None 2S tutes in this S. G- C-2 and inspect additional C-2 Plug defective tubes 45 tubes in this S. G. p ggn ,, C-3 C-3 result of first sample Perform action for w C-3 C-3 result of first N/A N/A N sample 3 3 C 't inspect all tutes in All other

  '                             this S. G., pluq de-     S. G.s are           None            N/A             N/A
 %                              fective tubes and        C-1 inspect 2S tubes i" Some S. G.s Perform action for       N/A             N/A each other S. G-        C-2 but no C-2 result of second additional   sample Prompt notification     S. G. are to NRC pursuant        C-3 to specification                                                                            3 6.9.1 Additional inspect all tubes in                                    y S. G. is C-3 each S. G. and plug                                    "

defective tubes. Prompt notification N/A N/A to NRC pursuant - to specification - 6.9.1 $ a> S=3-% Where N is the number of steam generators in the unit, and n is the number of steam generators inspected n during an inspection T a

THIS PAGE INTENTIONALLY LEFT BLANK MILLSTONE - l Jilt 2 3/4 4-79

THIS PAGE INTENTIONALLY LEFT BLANK MILLSTONE - LNIT 2 3/4 4-7h

TABLE 3.7-0 MAXIMUM ALLOWABLE POWER LEVEL WITH ONE OR MORE MAIN STEAM SAFET/ VALVES INOPERABLE MAXIMUM NUMBER OF IN0PERABLE SAFETi VALVES ON ANY ONE MAXIMUM ALLOWABLE STEAM GENERATOR POWiR LEVEL (% POWER) I 90.0 2 76.8 3 63.5 4 51.2 5 37.9 6 25.6 ( ( MILLSTONE - UNIT 2 3/4 7-2

 '                                                                Hay'15, 1978 PLANT SYSTEMS 3/4.7.8 HYDRAULIC SNUBBERS LIMITING CONDITION FOR OPERATION 3.7.8.1    All hydraulic snubbers listed in Tabl- 3.7-1 shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: dith one or more hydraulic snubbers inoperable, restore the inoperable snubber (s) to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the follow-ing 30 hours. SURVEILLANCE REQUIREMENTS I 4.7.8.1 Hydraulic snubbers shall be demonstrated OPERABLE by performance of the following augmented inservice inspection program:

a. Each hydraulic snubber with seal material fabricated from ethylene propylene or other materiais demonstrated compat'ible with the operating environment and approved as such by the NRC, shall be determined OPERABLE in accordance witn the inspection schedule of Table 4.7-3

                               , by a visual inspection of the snubber. Visual inspections of the s,.ubbers shall include, but are not necessarily limited to, inspection of the hydraulic fluid reservoirs, fluid connections, and linkage connections to the piping and anchors. Initiation of the Table 4.7-3 inspection schedule shall be made assuming the unit was previously at the 6 month inspection interval.
            .b. Each hydraulic snubber with seal material not fabricated from ethylene propylene or other materials demonstrated compatible  l with the operating environment shall be determined OPERABLE at least once per 31 days by a visual inspection of the snubber.

Visual inspections of the snubbers shcIl include, but are not necessarily limited to, inspection of the hydraulic fluid reservoirs, fluid connections, and linkage connections to the piping and anchors.

     , 4ILLSTONE - UNIT 2                3/4 7-21

May 15, 1978 PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

c. At least once per 18 months during shutdown, a representative sample of at least 10 hydraulic snubbers or at lust 10% of all snubbers listed in Table 3.7-1, whichever is less, shall be selected and functionally tested to verify correct piston movement, lock up and bleed. Snubbers greater than 50,000 lb.

capacity may be excluded from functional testing requirements.. Snubbers s' elected for functional testing shall be selected on a rotating basis. Snubbers identified as either "Especially Difficult to Remove" or in "High Radiation Zones" may be exempted from functional testing provided these snubbers were. demonstrated OPERABLE during previous functional tests. Snubbers found inoperable during functional testing shall be restored to OPERABLE status prior to resuming operation. For each snubber found inoperable during these functional tests, an additional minimum of 10% of all snubbers or 10 snubbers, whichever is less, shall also be functionally tested until no more failures are found or all snubbers have been functionally tested. MILLSTONE - UNIT 2 3/4 7-22 I

April 19, 1978 3/4.10 SPECIAL TEST EXCEPTIONS SHUTDOWN MARGIN LIMITINS CONDITION FOR OPERATION 3.10.1 The SH' J TDOWN MARGIN requirement of Specification 3.1.1.1 may be suspended for measurement of CEA worth and shutdown margin provided reactivity equivalent to at least the highest estimated CEA worth is available for trip insertion from OPERABLE CEA(s). 32 APPLICABILITY: MODES 2 and 3. ACTION:

a. With any full length CEA not fully inserted and with less than the above reactivity equivalent available for trip insertion, imediately initiate and continue boration at > 44 gpm ofz.1720 ppm boric acid solution or its equivalent untiT the L 1TDOWN MARGIN required by Specification 3.1.1.1 is restored.

b. With all full length CEAs inserted and the reactor subtritical 32 by less than the above reactivity equivalent, immediately initiate and continue boration at > 44 gpm ofc1720 ppm boric l acid solution or its equivalent until the SHUTDOWN MARGIN re-quired by Specification 3.1.1.1 is restored.

SURVEILLANCE REOUIREMENT3 4.10.1.1 The position of each full length CEA required either partially 132 or fully withdrawn shall be determined at least once per 2 hours. 4.10.1.2 Each CEA not fully inserted shall be demonstrated capable of full insertion when tripped from at least the 50% withdrawn positon with-in 24 hours prior to reducing the SHUTDOWN MARGIN to less than the limits of Specification 3.1.1.1. 32 MILLSTONE - UNIT 2 3/4 10-1

April 19,1978 3/4.2 POWER DISTRIBUTION LIMITS BASES 3/4.2.1 LINEAR HEAT RATE The limitation on linear heat rate ensures that in the event of a LOCA, the peak temperature of the fuel cladding will not exceed 2200 F. Either of the two core power distribution monitoring systems, the Excore Detector Monitoring System and the Incore Detector flonitoring System, provide adequate monitoring of the core power distribution and are capaole of verifying that the linear heat rate does not exceed its limits. The Excore Detector Monitoring System perfoms this function by continuously monitoring the AXIAL SHAPE INDEX with two OPERABLE excore neutron flux detectors and verifying that the AXIAL SHAPE INDEX is maintained within the allowable limits of Figure 3.2-2. In conjunction with the use of the excore monitoring system and in establishing the AXIAL SHAPE INDEX limits, the following assumptions are made: 1) the CEA insertion limits of Specifications 3.1. 3. 2, 3.1. 3. 5 a nd 3.1. 3.6 are satisfied 2) the flux peaking augmentation factors are as shown in Figure 4.2-1, 3) the AZIMUTHAL POWER TILT restrictions of Specification 3.2.4 are satisfied, and 4) the TOTAL PLANAR RADIAL PEAKING FACTOR does not exceed the limits of Specification 3.2.2. The Incore Detector Monitoring System continuously provides a direct measure of the peaking factors and the alarms which have been established for the individual incore detector segments ensure that the peak linear heat rates will be maintained within the allowable limits of Figure 3.2-1. The setpoints for these alartns include allowances, set in the conservative directions, for 1) flux peaking augmentation factors as shown in Figure 4.2-1, 2) a measurement-calculational uncertainty factor of 1.07, 3) an engineering uncertainty factor of 1.03, 4) an allowance of l 1.01 for axial fuel densification and themal expansion, and 5) a THERMAL POWER measurement uncertainty factor of 1.02. 3]4.2.2, 3/4.2.3 and 3/4.2.4 TOTAL PLANAR AND INTEGRATED RADI AL PEAKING FACTORS - F y AND Ff AND AZIMUTHAL POWER q TILT - T T The limitiations on F xy and Tq are provided to ensure that the assumptions used in the analysis for establishing the Linear Heat Rate and Local power Density - High LCOs and LSSS setpoints remain valid during operation at the various allowable CEA group insertion limits. The limitations on Ff and T are provided to ensure that the assumptions used in the analysis establishing the DNB Margin LCO, and Thermal f;argin/ Low Pressure LSSS setpoints remain valid during ope.2 tion at the various MILLSTONE - UNIT 2 B 3/4 2-1

April 19, 1978 POWER DISTRIBUTION LIMITS BASES allowable CEA group insertion lir,its. IfFfy,FforTq exceed their basic limitations, operation may continue under the additional restrictions imposed by the ACTION statements since these additional restrictions provide adequate provisions to assure that the assumptions used in establishing the Linear Heat Rate, Thermal Margin / Low Pressure and Local Power Density - High LCOs and LSSS setpoints remain valid. An AZIMUTHAL POWER TILT > 0.10 is not expected and if it should occur, subsequent operaticn would be restricted to only those operations required to identify the cause of this unexpected tilt. The value of T that must be used in the equation F q

                                                                  =F xy (1 + Tq) and Fr =Fr (1 + T q) is the measured tilt.

The surveillance requirements for verifying that F ,FfandTq are within their limits provide assuran:e that the actualT valuesTof F*7, T Fr and T qdo not exceed the assumed values. Verifying F xy and F r after each fuel loading prior to exceed ng 75% of RATED THERMAL POWER provides additional assurance that the core was properly loaded. 3/4.2.6 DNB MARGIN The limitations provided in this specifict. tion ensure that the assumed margins to DNB are maintained. The limiting values of the parameters in this specification are those assumed as the initial condi-tions in the accident and transient analyses; therefore, operation must be maintained within the specified limits for the accident and transient ,ana. lyses to remain valid. I MILLSTONE - UNIT 2 B 3/4 2-2

REACTOR COOLANT SYSTEM BASES 3/4.4.4 PRESSURIZER A steam bubble in the pressurizer ensures that the RCS is not a hydraulically solid system and is capable of accomodating pressure surges during operation. The steam bubble also protects the pressurizer code safety valves and power vperated relief valve agai'nst water relief. The power operated relief valve and steam bubble function to relieve RCS pressure during all design transients. Operation of the power operated relief valve in conjunction with a reactor trip on a Pressurizer-- Pressure-High signal, minimizes the undesirable opening of the spring-loaded pressurizer code safety valves. 3/4.4.5 STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be maintained. The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1. Inservice inspection of steam generator tubing is essential in order to maintain surveillance of the conditicas of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion. Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken. The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those chemistry limits found to result in negligible corrosion of the steam generator tubes. If the secondary coolant chemistry is not maintained within these limits, localized corrosion may likely result in stress corrosion cracking. The extent of cracking during plant operation would be limited by the limitation of steam generator tube leakage between the primary coolant system and the secondary coolant system (primary-to-secondary leakage = 0.5 GPM per steam generator). Cracks having a primary-to-secondary leakage less than this limit during operation will have an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents. Operating plants have demonstrated that primary-to-secondary leakage of 0.5 GPM per steam generator can readily be MILLSTONE - lRIT 2 B 3/4 4-2

                                                       ..e.     , . . . .

REACTOR COOLANT SYSTEM BASES detected by radiation monitors of steam generator blowdown. Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located and plugged. Wastage-t/pe defects are unlikely with preper chemistry treatment of the secondary coolant. However, even if a defect should develop in service, it will be found during scheduled inservice steam generator tube examinations. Plugging will be required for all tubes with imper-fections exceeding the plugging limit of 40% of the tube nominal wall thickness. Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect degradation that has penetrated 20% of the original tube wall thickness. Whenever the results of any steam generator tubing inservice inspec-tion fall into Category C-3, these results will be promptly reported to the Commission pursuant to Specification 6.9.1 prior to the resumption of plant operation. Such cases will be considered by the Commission on a case-by-case basis and may result in a requirement for analysis, laboratory examinations, tests, additional eddy-current inspection, and revision of the Technical Specifications, if necessary. MILLSTONE - MiIT 2 B 3/4 4-2a

AUG 1 1975 INTRODUCTION The Northeast Nuclear Energy Company, Millstone Unit I has been producing power commercially since December 28, 1970 with no significant adverse environmental impact evidenced as a result of plant operation. It is a boiling water reactor (BWR) licensed at 2011 megewatts thermal ((MWt) with a corresponding net electrical generation of 652.1 megawatts electrical (MWe). Millstone Unit 2 is a pressurized water reactor (PWR) designed to be licensed at2700< MWt, with a corresponding net electrical generation of about 870 MWe. Mill-stone Unit 2 is scheduled to begin commercial operation in 1975. A third unit, Millstone Unit 3 is under construction on this site and is presently scheduled for commercial operation in 1984. The Environmental Technical Specifications contained herein reflect the operating experience of Unit 1 and the need for continuing control and monitoring of plant operations to maintain the environmental impact to a level as low as practicable and in all cases within acceptable limits. In this regard, the specifications delineated are responsive to provisions of the National Environmental Policy Act of 1969. Such conditions and limitations necessary for controlling the operations of the Millstone Station have been incorporated so that the station will

   .have an acceptable environmental impact. Design features and operating practices bearing a direct relationc' nip to the effect on the environment have been described. An environmental-surveillance program provides a means to permit an assessment of the impact of the plant on the environ-ment. The surveillance program incorporates reporting levels and instruc-tions so that prompt notification of the U.S. Nucicar Regulatory Commission (NRC) shall be made in the event of an observable effect on a key parameter exceeding a specific level.

The administrative structure for the organizatien and management, review, audit, reports and records is described. The org 'izstional structure provides the emphasis on the protection of the safetn health, and environment. s e i

9.0 STARTUP PROGRAM The principal tests in the proposed startup program for Cycle 3 at listed below. These tests are sufficient to show that the as-loaded cote is safe for continued operation and is within the bounds assumed in the safety analysis. Acceptance Criteria and action to be taken if the acceptance criteria are not met are discussed below. The predicted values of the parameters being measured are not calculated until the end of cycle when the actual shutdown assembly burnups are known. Hot Functional Tests (1) CEDM Performance Verification Initial Criticality and Low Power Physics Tests (1) Initial Criticality (2) CEA Symmetry Check

a. Hot Zero Power (3) Critical Boron Concentration

a. Hot Zero Power, All Fods Out

b. Hot Zero Power, Banku 2 through 7 inserted (4) CEA Croup Worths

a. Hot Zero Power, Groups 7, 6, 5, 4, 3, and 2, non-overlap worth

b. Accumulated Total Worth of CEA Groups 7, 6, 5, 4, 3, and 2 (5) Iso-Thermal Temperature Coefficient

a. Hot Zero Power, 100% PDIL

b. Hot Zero Power, Banks 2 through 7 inserted Power Ascension Tests (1) Critical Boron Concentration

a. 50% Power, All Rods Out, Equilibrium Xenon

b. 100% Power, All Rods Out, Equilibrium Xenon (2) Power Distribution Verification

a. 50% Power, All Rods Out, Equilibrium Xenon

b. 100% Power, All Rods Out, Equilibrium Xenon (3) Iso-Thermal Temperature coefficient

a. 100% Power, 100% PDIL, Equilibrium Xenon

(4) Power Coefficient

a. 100% Power, 100% PDIL, Equilibrium Xenon Acceptance Criteria (1) CEA Group Worths The Greater of a) 15% of predicted b) i0.06% Ap (2) Total CEA Worth 10% of predicted (3) Isothermal Temp. Coef. 0.3 x 10-4 Ap/ F of predicted (4) Power Coef. 1 0.3 x 10-4 Ap/% PWR of predicted (5) Boron Concentration 75 PPM of predicted (6) CEA Symmetry Check 1.5c deviation from group average *

(7) Power Distribution Within technical specifications (8) Core Flow Determination 370,000 GPM at rated temperature and pressure

• This is not intended to be a go-no go criterion, but an indication of possible flux tilts during power operation. This test is for information only.

The following plan of action is provided if a measured parameter differs from the predicted valued by more than the acceptance criteria. (1) The physics test program will be extended to reconfirm the measured value, and/or (2) The predicted value will be reviewed to ensure that it accurately reflects the particular plant conditions under which the measurement was made and re-fined if appropriate. (3) It, after the above two steps, the disagreement persists, the safety analysis will be reviewed to determine whether the measured value of the par-ticular parameter in question, when combined with all of the other safety related parameters, increases the severity or consequences of accidents or anticipated operational occurrences. If equivalent safety for the plant can be demonstrated, the test results will be deemed acceptable. (4) The other phycies related safety parameters will be verified to be within acceptable limits by additional measurements if necessary. (5) The the combination of safety parameters determined above fall outside of range of safety parameters used to support the proposed operation of the plant, the plant operating limits will be adjusted to prevent conditions which could result in exceeding the specified acceptable fuel design limits. The acceptance criteria for the CEA Symmetry checks will be 1.5c or 15% deviation from the group average, whichever is greater. The reason that two

types of criteria are stated is that for high worth rods, a percent devia-tion is appropriate as is applied to other rod worth measurements. For small worth rods, however, an absolute deviation is required which is the same type of allowance as specified for reactivity coefficient measurements. This is not intended to be a go-no-go criterion, but rather an indication of the degree of tilt that might cause the Azimuthal Power Tilt specification to be exceeded. If the criterion is not met during the test, the test pro-gram will be extended to reconfirm the measured value. If the values still fall outside the stated criteria, the results will be reviewed to determine the potential impact upon plant operations. The acceptance criteria for power distribution verifications are the limits cited in the Technical Specifications and include the following: 3.2.1 Linear Heat Rate T 3.2.2 Total Planar Radial Peaking Factor Fxy 3.2.3 Total Integrated Radial Peaking Factor Fr 3.2.4 Azimuthal Power Tilt Tq If after review of the data, it is determined that a Technical Specification limit has been exceeded, then appropriate action as required by Technical Specifications will be taken.

10.0 EVALUATION OF RADIOLOGICAL CONSEQUENCES Control Room Dose to Operators Post-LOCA In Section 14.18.3.3 of the Millstone Unit No. 2 FSAR, the Control Room Dose to Operators was reported. This analysis was based on a power level of 2700 MWt as stated in Section 14.18.2. 2sults indicated that the calculated dose was well within the limits of GDC-19. In Section 6.S of the Millstone Unit No. 2 SER, the Staff documented their con-clusion regarding the habitability of the control room post-LOCA. On Page 6-56, the Staff stated:

     "We have calculated the potential radiation doses to control room personnel following a LOCA. The resultant doses are within the guidelines of GDC-19."

In light of the above information, no further analyses are provided.

Radiological Consequences of Steam Generator Tubc Rupture Accideat_ The radiological consequences for this accident were based on the analysis presented in section 7.3.3. The intent of this analysis is to verify that the site boundary doses do not exceed the guidelines of 10CFR100. The steam generator tube rupture incident is a penetration of the barrier between the reactor coolant system and the main steam system. The integrity of this barrier is significant from the standpoint of radiological safety in that a leaking steam generator tube allows the transfer of reactor coolant into the main steam system. Radioactivity contained la the reactor coolant mixes with water in the shell side of the affected steam generator. This radioactivity is transported by steam to the turbine and then to the condenser, or directly to the condenser via the main steam dump and bypass s ys t er:. Non condensible radioactive gases in the condenser are removed by the condenser air ejector discharge via the Unit 1 stack. To maximize the coolant transported from the primary to the secondary and; consequently, the releases to the atmosphere, the analysis was performed assuming the reactor does not trip until the minimum setpoint of the thermal Margin /Lew Pressure trip is reached. This prolongs the steam releases to the atmosphere and thus maximizes the site boundary doses. The methodology to calculate the site boundary doses for Steam Generator Tube Rupture is indicated below. In determining the site boundary dose, the thyroid and whole body doses were conservatively calculated. For purposes of the thyroid dose calculation, it is assumed that all leakages and releases during a given period of time occur instantaneoasly at the end of the period. In addition, the concentration in the steam generators is based on the minimum liquid mass occurring during that period. Also, for the purpose of added conservatism a preaccident iodine spike was assumed to occur. The concentration of I-131 in the steam generators was calculated by the following equation: Initial Conc. (Ci/LB) + [ Amount of activity leaked to steam generators f rom Tech Spec primary to secondary leak rate 4 minimum steam generator liquid mass during period). Thyroid dose is then calculated using the following equation: Dose (REi) = Concentration of I-131 (dose equivalent curie) X spiking factor X amount of steam released X steam generator partition factor X breathing rate X the 0-1 hour atmospheric dispersion coefficies X dose conversion factor

In determining the whole body dose, the major assumption made is that all noble gases leaked to the steam generators will be released to the atmosphere. The major periods of time for nobic gas releases are the same as those indicated for the thyroid dose. Therefore, the whole body dose is calculated by the following equation: Dose (REM) = 0.25 X average energy of betas and gammas per disintegration X primary coolant act'.vity concentration X amount of primary to secondary leak during period X amount of primary to secondary leak during period X the 0-2 hour atmospheric dispersion coefficient The radiological release criterion for this analysis is that the 2-hour dose at the site boundary should not exceed 10CFR100 guidelines. Table 1 shows the assumptions used in the calculation for the radiological release and Table 2 lists the assumptions used in the analysis. The results of the analysis are that 48292 lbs of primary coolant are trans-ported to the steam generator secondary side. Based on this release and the values presented in Table 1, the site coundary doses calculated are: Thyroid (DEQ I-131): 6.0 x 10-3 rem Whole Body (DEQ XE-133): .1 rem The radiological release criterion for this analysis is that the 2-hour dose at the site boundary should not exceed 10CFR100 guidelines.

Table 1 Assumptions for the radiological evaluation for the steam generator tube rupture accident. Conservative Assumption Basis

1) Reactor Coolant System Tech Specs

     !!aximum Allowable Concentration 1-131 (DEQ) = 1.0uci/gm

2) Steam Generator Maximum Tech Specs Allowable Concentration I-131 (DEQ) = .1 uCi/gm

3) Reactor Coolant System Tech Specs Maximum Allowable Concentration of Noble Cases Xe-133 (DEQ) = 100/E uCi/gm

4) Steam Generator Partition Factor = .01

5) Air Ejector Partition Factor = .0005

6) Atmospheric Dispersion Coefficient: 95% maximum X/Q's for the years Site Boundary = 1.03 x 10-4 sec/m3 1974-1976

7) Breathing Rate = 3.47 x 10-4 m3/see SRP 15.6.3

8) I-131 dose conversion factor = 1.49 x 106 rem /Ci Reg. Guide 1.109

9) Dose Increase Factor Due to Iodine NRC Criterion Spike = 60.

          -                             Tabic 2 SEQUENCE OF EVENTS FOR THE STEAM GENERATOR TUBE RUPTURE INCIDENT Event                                        Setpoipt or Value Time (Sec) 0.0      Tube Rupture Occurs L                                                                          --

825.2 Pressurizer Empties Low Pressurizer Pressure Trip Condition 1728 psia 826.1 825.6 Dump Valves Open 825.6 Bypass Valves Open 826.6 CEAs Begin to Drop into Core Maximum Steam Generator Pressure 901.4 psia 8 39.0 900.0 Dump Valves Close 913.2 Bypass Valve Closes 1800.0 Operator Initiates Appgopriate Action and Begins Cooldown to 300 F O}}