ML19210B287

Safety Evaluation Supporting Amend 17 to DPR-50

ML19210B287
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Site:	Crane
Issue date:	05/18/1976
From:	Office of Nuclear Reactor Regulation
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UNITED STATES 4

NUCLEAR REGULATORY COMMISSION f=

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION

, SUPPORTING AFENDMENT NO. 17 TO FACILITY OPERATING LICENSE NO. DPR-50 METROPOLITAN EDISON COMPANY JERSEY CENTRAL POWER AND LIGHT COMPANY PENNSYLVANIA ELECTRIC C0F/ANY THREE MILE ISLAND NUCLEAR ST)IION, UNIT 1 DOCKET No. 50-289 Introduction By letter dated January 13, 1976, as supplemented by letters dated February 11, 1976 and April 2, 1976, Metropolitan Edison Company (Meted) requested a change in the Technical Specifications of License No. DPR-50 for the Three Mile Island Nuclear Station, Unit No. 1 (TMI-1). The proposed amendment is to permit operation of TMI-l as reloaded for Cycle 2 operation. The proposed 'mendment also incorporates the change, requested by letter dated August 8, 1975, which was submitted pursuant to Section 50.46 and Appendix K of 10 CFR Part 50 and the Commission's Order for Modification of License dated December 27, 1974.

Discussion The TMI-l reactor core consists of 177 fuel assemblies, each with a 15x15 array of fuel rods. The cycle 2 reload will involve the removal of all batch 1 fuel asse=blies, the relocation of once-burned batch 2 and 3 fuel assemblies and the introduction of 56 fresh batch 4 fuel assemblies. Ihe batch 4 assemblies will occupy primarily the periphery of the core and 8 locations interior to the core.

Meted has proposed changes to the present Technical Specifications as a result of:

changes and relocation of fuel assemblies as described above; use of the B&W-2 CHF correlation with a 95/95 confidence level and extended pressure application to 1750 psi; use of a reactor coolant flow rate equal to 106.5* of cycle 1 design flow; and E=ergency Core Cooling System Final Acceptance Criteria (FAC). Meted has provided technical information which includes a general description of the reload core, detailed mechanical design data on the reload fuel, nuclear and ther=al-hydraulic design data, accident and transient analysee, fuel rod bow analyses and the loss of coolant accident (LOCA) analysis in support of the reload.

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' Evaluation Fuel and Mechanical Design Creep collapse calculations were performed by Meted for three-cycle assembly power histories for TMI-1 using the Babcock & Wilcox (B&W) computer code, CROV, which we approved in our Generic Review of 3&W Cladding Creep Collapse Analysis Topical Report, BAW-10084, issued on August 9,1974. The calculations included conservative treatment of ef fects of fission gas (no credit taken), cladding thickness (lower tolerance limit), initial cladding ovality (upper tolerance limit),

and cladding temperaturc (assembly outlet temperature) on collapse time. The most limiting assembly was found to have,a collapse time which is greater than the max 1=um projected cycle 2 life of 19,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> and is therefore acceptable.

Fuel ther=al analysis calculations that account for the effects of fuel densification were performed with the approved version of the B&W analytical model TAFY as described in B&W Topical Report BAW-10044 of May 1972. Fuel densification results in increases in stored energy, increases in linear thermal output and increases the probability of local power spikes from axial gaps. During cycle 2 operation, the highest relative assembly power levels will occur in batch 3 fuel. Fuel temperature analysis for batches 2 and 3 fuel is documented in the TMI-l Fuel Densification Report, BAW-1389 of June 1973. Although the batch 4 fuel has a higher linear heat generation rate (20.15 kw/ft vs 19.6 kw/ft) due to a reduced active fuel length, the higher initial density results in a lower maxi =um predieted centerline temperature. In view of the above, we find the Meted's fuel thermal analysis acceptable.

The batch 4 fuel assemblies are not new in concept and they do not utilize different component materials. Therefore, on the bases of the analysis presented in the reports referenced, we conclude for TMI-1 cycle 2 that:

(a) The fuel rod mechanical design provides acceptable safety margins for normal operation, and (b)

The effects of fuel densification have been adequately accounted for in the fuel design.

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. Thermal-Hydraulic Analysis The thermal hydraulic calculations for the cycle 2 reload core were made using previously approved models and methods.

There were no differences due to mechanical differences since the new fuel elements are mechanically similar and flow resistances are lower than the previously analyzed cycle 1 core.

During cycle 1 the reactor coolant flow was measured for TMI-1.

With the reactor operation at 100% of full power on February 16, 1976, calorimetric and flow measurements were made and averaged. A description of the flow test and an error analysis were reported by letter dated April 8, 1976. The results of the flow test indicate a nominal flow of 109.3% of the design flow rate.

The error analysis, based on measurement errors, showed a 2a core flow error of 1.8%.

Thus, the maximum usable flow rate ;or calculations would be 107.5%

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of the design flow. To provide additional conservatism in their calculations, Meted has used a flow rate of 106.5% of design. We find that the flow test and analysis performed are acceptable and agree that this is a cbaservative flow rate.

In their letter of April 8, 1976, Meted has co=sitted to verify the flow rate for TMI-l within three months following refueling.

There-after, the flow rate will be verified every six months, plus or =inus thirty days. All verificacions will be done by the heat balance technique described in their April 8, 1976 letter.

The overpower trip, as used in the analyses of accidents and trans-ients for cycle 2 operation, has still retained the 4.6% flow penalty due to vent valves ured in the FSAR analyses. As discussed later in this evaluation, this is an additional conservatism and therefore, additional margin exists beyond that indicated in the accident analyses.

The flux / flow trip setpoint previously determined for cycle 1 was re-evaluated for the cycle 2 core. The procedure was revised to use the measured flow instead of the design flow rate. Like the pre-viously mentioned overpower trip and accident analysis, the flux / flow trip setpoint includes the penalty for a stuck open vent valve.

Thus for the pump coast down analysis the 4.6% penalty due to vent valves,

has been retained. The coast down analysis shows that with a.

flux / flow trip setpoint of 1.08, the minimum DNBR does not go below 1.30.

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On March 10, 1976, we sent s letter to Meted stating that B&W report, "B&W Operating Expe_.ance of Reactor Internals Vent Valves" had been reviewed and that sufficient evidence had been presented to assure that the vent valves will remain closed during normal operation. Based on this conclusion, it was stated that the flow penalty could be eliminated from analyses at the request of the utility; however, the corresponding modifications to the Technical Specifications must be reviewed by us prior to implementation. Meted retained the vent valve penalty in this analyses for cycle 2 and therefore additional conservatism exists.

Two further changes reflected in the cycle 2 reload report and the accompanying Technical Specifications are:

(a) The use of the B&W-2 CHF correlation down to pressures of 1750 psi instead of the previous lower pressure limit of 2,000 psi, and (b) A reduction in the minimum ellowable DNBR from 1.32 to 1.30.

We recently completed a re-evaluation of the B&N-2 CHF correlation to verify its continued suitability in relation to available rod bundle DNB data.

Ne determined that the BAN-2 correlation continues to be an acceptable correlation over the pressure, quality, mass flux, rod diameter and red spacing range of its original data base.

In conjunction with our reevaluation of the B&W-2 rHF correlation we also reviewed the Meted's proposed modificatic;s to the correlation for the cycle 2 core. The original data base for the correlation covered the pressure range 2000-2450 psia and resulted in a 1.32 minimum allowable DNB ratio to ensure with 99% confidence that 95%

of the hot rods did not experience DNB. As an attachment to their letter of February 3,1976, B5W provided information which compared the B W-2 CHF correlation with data in the low pressure range from five different test bundles. The mean measured-to-predicted ratio for all data was 1.05 and the minimum allowable DNBR was 1.29 for a 95% confidence that 95% of the hot rods at the DNBR would not experience DNB.

The 1.32 minimum DNB ratio used by B5W is based upon 95%'of the hot rods at that DNBR not experiencing DNB, with a 99% confidence.

If the confidence level is changed to 95%, which is consistent with regulatory requirements as expressed in the standard review plan, the minimum allowable DNBR becomes 1.30.

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Based on the above, we find both the extension of the B&N-2 CHF correlation to pressures down to 1750 psia and the change to a minimum DNBR of 1.30 to be acceptable. The B&W-2 CHF correlation has been shown to be conservative in the low pressure region and the change to a 1.30 minimum DNBR is consistent with the requirements of Standard Review Plan 4.4.

Nuclear Analysis Meted has provided values for core physics parameters for the TMI-1 cycle 2 core which reflect minor differences when compared to those for cycle 1.

These differences are attributable to the fact that the core has not yet reached an equilibrium cycle and such differences are to be expected. We have concluded that no significant changes exist in the core design between cycles 1 and 2.

In addition, the same calculational methods and design information were used to obtain the i=portant nuclear design parameters. Based on the above and the fact that startup tests (to be conducted prior to power operation) will verify that the critical aspects of core perfor=ance are within the assumptions of the safety analysis, we find Meted's nuclear analysis for cycle 2 to be acceptable.

Accident and Transient Analysis Accident and Transient analyses reported in paragraphs 7.1 through 7.14 of the TMI-l cycle 2 reload report submitted February 11, 1976, were exanined and we agree that the cycle 2 reload core is ther= ally and hydraulically conservative and of the sa=e design and manufacture as the cycle 1 core. We also agree that the reactivity coefficients and other input data is the same as, or is bounded by previous analyses. We have reviewed Meted's submittal and agree that in no case are the consequences of transients more severe than previously analyzed.

Fuel Rod Bow Evaluation The effect of rod bowing on DNBR was considered. Our review of Meted's submittal dated April 2,1976, indicates that the peaking penalty due io rod bowing, which the licensee has calculated as 1.6%

is acceptable. The effect of the rod bow penalty on the limits for normal operation as provided in BAW-10079, " Operational Parameters for R&W Rodded Plants," has been found by us to be within the conservatism of the current limits.

The design basis values for average linear heat rate of 5.80 kw/ft and power spike of 1.022 are greater than the actual values of 5.73 kw/ft and 1.018, respectively.

Thus adequate margin is provided to absorb the rod bow penalty.

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. Technical Specifications We have reviewed the Technical Specification changes submitted by Meted for operation with the cycle 2 core and find them to be acceptable.

ECCS Analysis On December 27, 1974, the Atomic Energy Commission issued an Order for Modification of License implementing the requirements of 10 CFR 50.46, " Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors." One of the requirements of the Order was that a re-evaluation of ECCS cooling performance calculated in accordance with an acceptable evaluation model which conforms with the provisions of 10 CFR 50.46 shall be submitted.

The Order also required that the evaluation shall be accompanied by such proposed changes in the Technical Specifications or license amend-ment as may be necessary to implement the evaluation results. As required by our Order of December 27, 1974, Meted has submitted an ECCS re-evaluation and related Technical Specification changes.

The re-evaluation and Technical Specifications were submitted by letter dated August 8,1975, using the B&W ECCS evaluation model as described in BAW-10104.

The background of our review of the B&W ECCS evaluation model and its application to IMI-l is described in our Safety Evaluation Report (SER) for TMI-1 dated December 27, 1974, issued in connection with the Order for Modification of License. The bases for acceptance of tne principal portions of the evaluation model are set forth in our Status Report of October 1974 and the Supplement to the Status Report of November 1974 which are referenced in the December 27, 1974 SER.

The December 27, 1974 SER also describes the various changes required in the earlier version of the B&W codel. Together, the December 27, 1974 SER and the Status Report and its Supplement describe an acceptable ECCS evaluation model and the basis for our acceptance of the model.

The TMI-l ECCS evaluation which is covered by this safety evaluation report properly conforms to the accepted model. Meted's July 9,1975, submittal contains docu=entation by reference to B&W Topical Reports of the revised ECCS model (with the modifications described in our December 27, 1974 SER) and a generic break spectrum appropriate to TMI-l (BAR-10103 and BAW-10104).

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. The generic analysis in BAW-10103 identified the worst break size as the 8.55 ft2 double-ended cold leg break at the pump discharge with a Cp=1.0.

The cable below summarizes the results of the LOCA limit analyses which determine the allowable linear heat rate limits as a function of elevation in the core for THI-1.

Elevation LOCA Peak Cladding Max. Local Time of (ft)

Limit Temperature ( F) 0xidation Rupture (kw/ft)

(%)

(tec) 2 15.5 2002 3.92 12.25 4

16.6 2136 4.59 13.01 6

18.0 2146 4.46 15.55 8

17.0 2110 5.19 15.01 10*

16.0 1931 2.93 39.20

• See discussion below The maximum core-wide metal-water reaction for TMI-l was calculated to be 0.557 percent, a value which is below the allowable limit of 1 percent.

As shown in the tabulation, the calculated values for the peak clad temperature and local =etal-water reaction were below the allowable limits specified in 10 CFR 50.46 of 22000F and 17 percent, respectively.

BAW-10103 has also shown that the core geometry remains amenable to cooling and that long-term core cooling can be established.

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< We noted during our review of BAW-10103 that the LOCA limit calculation at the 10-foot evaluation in the core showed reflood rates below 1 inch /second at 251 seconds into the accident (Section 7.2.5).

Appendix K to 10 CFR 50.46 requires that when reflood rates are less than 1 inch /second, heat transfer calcula-tions shall be based on the assumption that cooling is only by steam, and shall take into account any flew blockage calculated to occur as a result of cladding swelling or rupture as such blockage might affect both local steam flow and heat transfer.

As indicated by the staff in our Status Report and Supplement,.

dated October 1974 and November 1974 respectively, a steam cooling model for reflood rates less than 1 inch /second was not submitted by B&W for staff review. The steam cooling model submitted by B&W in BAW-10103 is therefore considered to be a proposed =odel change requiring further staff review. Accordingly, B&W was informed that until the proposed steam cooling model is reviewed, the heat transfer calculation at the 10-foot elevation during the period of steam cooling specified in BAW-10103 must be further justified. In lieu of using their proposed steam cooling model, B&W has submitted the results of calculations at the 10-foot elevation using adiabatic heatup during the steam cooling period, where this period is defined by BiW as the time when the reflood rate first goes below 1 inch /

seccnd to the time that it is predicted that the 10-foot elevation is covered by water.

The new calculated peak cladding temperature, local metal-water reaction and core-wide metal-water reaction at the 10-foot elevation are 19460F, 3.02%, and.647%, respectively. These values re=ain below the allowable limits of 10 CFR 50.46 and have been calculated in an acceptable manner. Until we have accepted a steam cooling model, these values will serve as the LOCA results for TMI-1 at the 10-foot elevation.

Our review of plant-specific assumptions discussed in the following paragraphs regarding the TMI-1 analyses addressed the areas of single failure criterion, long term boron concentration, potential submerged equipment, partial loop operation, ECCS valve interlocks, and the containment pressure calculation.

Singl4 Failure Criterion Appendix K to 10 CFR 50 of the Commission's regulations requires that the combination of ECCS subsystems to be assumed operative shall be those available after the most damaging single failure of ECCS equipment has occurred. Meted has assumed all containment cooling 1 5 lB 5 i 9 I

. systems operating to minimize containment pressure and has separately assumed the loss of one diesel and therefore the loss of one electrical safeguards bus to minimize ECCS cooling. We concluded in our Status Report dated October 1974 that the application of the single failure criterion was to be confirmed during subsequent plant reviews. This has been done.

A review of TMI-l piping and instrumentation diagrams indicated that the spurious actuation of certain motor-operated valves could affect the appropriate single failure assumptions. A spurious actuation of core flooding tank (CFT) vent valves CF-V3A or CF-V3B would result in a decrease in CFT pressure. Since CFT pressure is important in mitigating the consequences of a LOCA, we require that these nor= ally closed =otor-operated valves (CF-V3A and CF-V3B) have their power disconnected and associated breakers locked open except when adjusting core flooding tank pressure. To further minimize the potential for a water hammer due to the discharge of ECC water into a dry line and to ensure that air pockets have not for ed in the ECCS lines and pu=ps we require the venting of High and Low Pressure Injection pump casings and system high points during each refueling outage. Meted has committed to perform these required actions and to verify the Eigh and Low Pressure Injection pu=ps and lines are filled with water prior to cycle 2 power operation. We find this acceptable.

Manually-Controlled Electrically-Operated Valves As requested, Meted has submitted a single failure analysis for manually-controlled, electrically-operated ECCS valves. This analysis (TMI-l FSAR Chapter 6 and Meted letter dated February 11, 1976) demonstrates that no credible single failure or operator error affecting any =anually-controlled, electrically-operated ECCS valve could adversely affect ECCS perfor=ance. We find this acceptable.

Submerged Valves Meted has identified the following valves as becoming submerged when the entire contents of the BWST are discharged into the reactor contain-ment building:

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Safety MU-V2A,B Letdown cooler containment isolation IC-V2 Intermediate cooler contain-ment isolation Non-safety IC-VlA,B Letdown cooler shell side inlet isolation IC-V20 RC drain tank cooler outlet isolation MU-VlA,B Letdown cooler tube side inlet isolation WDL-V302 RC Tank recirculation WDL-V305 RC Tank recirculation Only three of the above valves (IC-V2, MU-V2A and MU-V2B) are Engineer-ed Safety Feature valves. These three containment isolation valves will have performed their safety function prior to becoming submerged and therefore,their submergence will not affect any ECCS function. It has been determined that if power were maintained or inadvertently applied, during or af ter submergence, to any one of the above valves, there would be no adverse affects on the remainder of the electrical system thus, we find this evaluation acceptable.

Electrical Independence Meted in a letter dated April 19, 1975, identified the worst single failure which could occur as the loss of one diesel resulting in the failure of the IC Engineered Safeguards Valve 480V control center.

In such case power to open valves DH-V1, DH-V2, and DH-V3 in the boron control primary flow path or RC-V4 in the alternate flow path would not be immediately available. Minimum requirements are that either RC-V4 be operable or DH-V1,2 and 3 be operable. In this case the Procedure, as detailed in' Meted's April 8, 1976 submittal, would be used to open RC-V4 to establish long-term flow for post LOCA boron control.

Since a minimum of 30-days would be available to accomplish this emetgency action, no modifications are necessary. We concur in this evaluation and find the above referenced procedure acceptable.

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. Interlocks Decay heat drop line valves DH-V1 and DH-V2 were reviewed to ensure that they are properly interlocked to prevent opening while the reactor coolant system is pressurized. DH-V1 is interlocked to RC3A-PS2 through engineered safegu trd actuation channel "A" and DH-V2 is interlocked to RC3A-PSS throuah engineered safeguard actuation channel "B".

These valves have independent and diverse interlocks to prevent them frcm being opened unless the reactor coolant system pressure is below 400 psi. We find this design acceptable.

Ccntainment Pressure The ECCS containment pressure calculations for TMI-l were done generically by B&W for reactors of this type as described in BAW-10103. We reviewed B&W's ECCS evaluation model (Status Report and Supplement dated October 1974 and November 1974, respectively) and concluded that B&W's containment pressure codel was acceptable for ECCS evaluation. We required, however, that justification of the plant-dependent input parameters used in the analysis be submitted for our review of each plant. A containment pressure calculation specific to TMI-l was contained in Meted's August 8,1975 submittal.

Justification for the containment input data was submitted for TMI-l on October 23, 1975. This justification includes a comparison of the actual containment parameters for TMI-1 with those assu=ed in BAW-10103. Meted has re-evaluated the contain=ent net-free volume, the passive heat sinks, and operation of the containment heat-removal syste=s with regard to the conservatism for the ECCS analysis.

This evaluation was based on as-built design infor=ation. The containment heat removal s:.2 ems were assumed to operate at their =aximum capacities, and minimum ope.mcional values for the spray water and service water temperature were assu=ed.

The containment pressure analysis by B&W in BAW-10103 was demonstrated to be conservative for TMI-1.

We have concluded that the plant-dependent infor=ation used for the ECCS containment pressure analysis for TMI-1 is reasonably conservative, and therefore, the calculated containment pressures are in accordance with Appendix K to 10 CFR 50 of the Commission's regulations.

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. Long-Term Boron Concentration We have reviewed the proposed procedures and the systems designed for preventing excessive boric acid buildups in the reactor vessel Meted implemented during the long-term cooling period after a LOCA.

procedures for TMI-l which would allow adequate boron dilution during the long term and which will e= ploy a concept similar to that described We noted that a failure of a diesel vill affect each in BAW-10103.

of the proposed dilution modes. Meted has indicated that the controllers for all the pertinent valves are located in the auxiliary building, thus enabling the operator to connect power to the valves with ju=per Based on calculations by B&W, which we have found acceptable, cables.

over 30 days are available, taking credit for natural circulation through the vent valves, before forced circulation is necessary; therefore, the Meted's backup procedure to a power failure is accept-able.

As initially proposed by Meted in their letter dated May 27, 1975, dilution Mode 1 was to be first attempted to establish suction from the reactor vessel outlet pipe through the decay heat drop line with one LPI string. It is our position that Mode 1 should not be attempted as a method to control boron concentration in the core during long-term cooling. As stated in Meted's letter of May 27, 1975 and BAW-10103 the success of Yode 1 is not ensured because of the possibility of gas or steam entrain =ent in the decay heat suction nozzle. Such gas or steam entrainment can result in severe damage to the decay heat re= oval Long-term heat removal requirements can exist for long durations pump.

(days or months) after the accident and continuous operation of one In the event of train of the decay heat removal system is required.

equipment malfunction in this train, no method is available to remove the decay heat if the other train has been previously damaged. Therefore, since initiation of Mode 1 is not allowed, Modes 2 and 3 (as proposed in Meted's May 27, 1975 letter) must be single failure proof in combination.

Mode 2, using the hot leg drain approach is satisfactory as one of, the two methods of preventing excessive concentration of boric acid in Mode 3 is a backup to Mode 2 and e= ploys hot leg injection the core.

through the pressurizer. This backup method required installation of a 1 1/2 inch check valve in the decay heat pressurizer auxiliary spray line and upgrading of the motor operator of valve RC-V4 so that it is Meted has made the now qualified for the post LOCA environment.

We find necessary modifications during the present refueling outage.

this to be acceptable.

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. We have reviewed the operating procedures associated with this proposal and conclude that these procedures are acceptable.

Partial Loop Analyses To allow an operating configuration with less than four reactor coolant pumps on the line (partial loop), we required an analysis of the predicted consequences of a LOCA occurring during the proposed partial loop operating mode (s). Meted submitted an analysis for partial loop operation with one idle reactor coolant pump (three pumps operating) in their August 8, 1975 submittal. This analysis assumes that the worst break was the 8.55 fc2 guillotine at the reactor coolant pu=p discharge, with Co =1.0 as reported in BAW-10103. The worst break selected was located in the active leg of the partially idle loop.

Placing the break at the discharge of the pu=p in an active cold leg of the partially idle loop (instead of at the discharge of the pump in an active cold leg of the fully active loop) yields the most degraded positive flow through the cor, during the first half of the blevdown and results in higher cladding temperatures.

Tu. =aximum cladding temperature for the one-idle-pump mode of operation was 17660F. Meted's analysis used an initial pin ' ressure of 1500 psi. The results of a new analysis were submitted 1.o reflect a more appropriate value of initial pin pressure. As was demonstrated in the time-in-life sensitivity study in BAW-10103, the worst pin pressure for this analysis is 760 psi.

The maximum cladding temperature for this analysis is 17840F, a value which is within the criterion of 10 CTR 50.46.

This analysis may be used to support Meted's proposed operation with one idle reactor coolant pump.

Since an analysis of ECCS cooling performance with one idle reactor coolant pump in each loop has not been submitted, power operation in this configuration will be limited by Technical Specifications to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Single loop operation (i.e., operation with two idle pumps in one loop) is prohibited.

Technical Specifications We have reviewed the proposed Technical Specification changes to assure that operation of TMI-l during cycle 2 will be within the limits imposed by the Final Acceptance Criteria (FAC) for ECCS performance. Changes in the allowable heat generation rates as a function of height in the core have been accommodated by revision of the power-flow-imbalance specifications (Figures 3.5 2G-I). Only minor change has been made by Meted in the rod position limit specification.

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. The reduction in allowable heat generation rate (kw/ft) in the lower half of the core has been accoc=odated by reducing the allowable negative imbalance at full power by approximately 4%.

The increase in allowable heat generation rate at the top of the core (as compared to that permitted by the Interim Acceptance Criterfon) permits relaxing the positive axial i=halance by approximately 3%.

On the basis of our review, we find the technical specification changes as proposed to be acceptable.

We have completed our review of the TMI-1 ECCS performance re-analyses and have concluded:

(a) The proposed Technical Specifications are based on a LOCA analysis performed in accordance with Appendix K to 10 CFR 50.

(b) The ECCS minimum contain=ent pressure calculations were perfor ed in accordance with Appendix K to 10 CFR 50.

(c) The single failure criterion will be satisfied orovided that the

' requirements as specified in this Safety Evaluation Report are implemented.

(d) The modified procedures for long-term cooling after a LOCA are acceptable. The necessary plant modifications to provide assurance that the ECCS can be operated in a manner which would prevent excessive boric acid concentration from occurring have been =ade during the present refueling outage.

(e) The proposed mode of reactor operation with one idle reactor coolant pump is supported by a LOCA analysis performed in accordance with Appendix K to 10 CFR 50.

Operation with one idle pu=p in each loop is restricted to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

We have completed our evaluation of the TMI-l cycle 2 reload application and conclude that the licensee has perfor=ed the required analyses and has shown that operation of the cycle 2 core vill be within applicable fuel design and performance criteria. In addition, we conclude that the licensee's proposed Technical Specification changes cect the Final Acceptance Criteria based on an acceptable ECCS model conforming to the requirements of 10 CFR 50.46 and that the restrictions i= posed on the facility by the Cocmission's December 27, 1974 Order for Modification of License should be :erminated and replaced by the limitations established in accordance with 10 CFR 50.46.

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7 Conclusion We have concluded, based on the considerations discursed above, that:

(1) there is reasencble assurance that the health and safety of the public will not be endangered by operation in the proposed canner, and (2) such activitics will be conducted in eccpliance with the Cc ission's regulations and the issuance of this atendeent will not be initical to the cctton defcase and security or the health and safety of the public.

Dated: PA Y ] 3 }g g S

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