ML19210B120

Safety Evaluation Supporting Amend 29 to DPR-50

ML19210B120
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Site:	Crane
Issue date:	04/22/1977
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UMTED STATES j

NUCLEAR REGULATORY COMMISSION

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\\ *% g SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION AMENDMENT NO. 29 TO FACILITY OPERATING LICENSE NO. DPR-50 METROPOLITAN EDISON COMPANY JERSEY CENTRAL POWER AND LIGHT COMPANY PENNSYLVANIA ELECTRIC COMPANY THREE MILE ISLAND NUCLEAR STATION, UNIT NO. 1 DOCKET NO. 50-289 Introduction By letters dated October 29, 1976, January 26, 1977, and February 23, 1977, Metropolitan Edison Company (the licensee) requested several amendments to Appendix A to Facility Operating License DPR-50 for Three Mile Island Nuclear Station Unit No.1 (TMI-1). The change requested by the licensee's letter of October 29, 1976, would revise the TMI-l Technical Specifications to permit irradiation of the TMI-1 reactor vessel material surveillance specimens in the Three Mile Island Nuclear Station Unit No. 2 (TMI-2) reactor vessel pursuant to the provisions of 10 CFR 50, AppeMix H,Section II.C.4.

By letter dated December 17,1976, we requested additional information concerning this proposed amendment.

This information was furnished by the licensee in letters dated December 29, 1976, and January 20, 1977.

The change requested by the licensee's letter of January 26, 1977, would revise the TMI-1 Technical Specifications to reflect plant operating limitations for the fuel loading to be used during Cycle 3.

By letter dated March 23, 1977, we requested additional information concerning this proposed amendment. This information was furnished by the licensee's letter of March 31, 1977.

The change requested by the licensee's letter of February 23,1977, would revise the TMI-1 Technical Specifications to update the reactor coolant system pressure limits during system heatup and cooldown.

In support of this request the licensee also submitted Babcock and Wilcox Report

" Analysis of Capsule TMI-lE from Metropolitan Edison Company Three Mile Island Nuclear Station Unit 1," BAW-1439 (January 1977).

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Discussion and Evaluation 1.

Reactor Vessel Material Surveillance Procram The original TMI-l design included three reactor vessel surveiDance specimen holder tubes (SSHTs) located near the reactor inside vessel wall.

Each of these SSHTs housed two capsules containing reactor vesse surveillance specimens.

During a refueling outage in the Spring of 1976 it was discovered that the SSHTs had suffered severe damage.

To prevent further damage, all surveillance capsules and all parts of the SSHTs that had failed or were deemed likely to fail were removed from the vessel.

The NRC granted the licensee an exemption which permitted operation of TMI-l in the following Cycle (Cycle 2) without the surveillance specimens present. This exemption was granted :,n the basis that the surveillance specimens had already received a neutron exposure in excess of that which the reactor vessel would receive by the end of Cycle 2.

At the time the exemption was granted it was expected that SSHTs of improved design would be reinstalled in TMI-1.

Since the discovery of the damage to the SSHTs, Babcock & Wilcox Company (B&W), the reactor supplier, has undertaken the design, manufacture and testing of an improved SSHT.

SSHTs of this improved design are presently installed in Davis-Besse Unit No. 1, Crystal River Unit No. 3 and Three Mile Island Unit No. 2 (THI-2). All three of these plants have reactors supplied by B&W and all are in the process of beginning initial operation within the next few months.

In addition, all of these reactors are of the same basic B&W 177 fuel assembly vessel design as TMI-1.

The acceptability of the redesigned SSHTs has been demonstrated by a test program reviewed and approved by the NRC staff and performed at Davis-Besse Unit No. 1.

Installation of the redesigned SSHTs in'the Davis-Besse Unit No.1, Crystal River Unit No. 3 and TMI-2 reactor vessels did not present any unusual difficulties because it was performed prior to neutron activation of the reactor internals.

Studies of methods to install the redesigned SSHTs in the irradiated B&W reactors, however, indicated that substantial difficulties would be experienced, primarily because precision machining, alignment and inspection must be performed remotely and under water.

" ' Although such problems do not in themselves justify relief from a require-ment to reinstall the SSHTs in TMI-1, they would cause significant radiation to personnel.

Based on their experience in removing the SSHTs at TMI-l and other reactors, B&W estimated that installing SSHTs in irradiated reactors would result in personnel exposure of about 100 man-rem per reactor.

In the interest of maintaining the radiation exposure of plant personnel as low as reasonably achievable, the licensee, in cooperation with B&W and the owners of other B&W 177 fuel assembly plants, has proposed an, alternative program that does not require reinstalling the SSHTs in TMI-l.

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The proposed plan involves integrating the interrupted surveillance program at operating reactors which suffered damage to SSHTs into the programs for similar new plants.

For TMI-l this method for carrying out the program is in accord with the Site Integrated Surveillance Program permitted by Appendix H, 10 CFR 50 Paragraph II.C.4, in that the TMI-l program would be conducted in TMI-2 and, therefore, would be conducted at the same site.

There are three distinct features of these proposed programs:

1.

A host-reactor feature, in which the original surveillance materials from one or more reactors that have been in service will now be irradiated in a new host reactor which has been fitted with the*

newly-designed capsule holders; 2.

An augmented surveillance feature in which more weld metal specimens and some larger fracture mechanics (compact tension (CT)) specimens will be included in the program; and 3.

A data-sharing feature in which all available irradiation data for all of the beltline welds of a given reactor vessel will be considered by the licensee or his consultants in predicting the adjusted reference temperature and in making any fracture analyses for that vessel.

Typically, several of the welds in any one vessel were made with the same weld wire and flux as those used on some other reactors.

The data sharing feature is required because the welds in these reactors have high radiation sensitivity due to high copper content, large and random variation of copper from point to point in the weld, and low initial upper shelf energy.

The specific program proposed for TMI-l involves irradiating the remaining original TMI-l surveillance capsules (one has been removed and tested) in some o' the locations provided in the TMI-2 vessel.

This plan will accomplish the 04 aginal purpose of obtaining information on the effect of radiation on a material that is one of the controlling materials in the TMI-l reactor vessel on a schedule that provides an appropriate lead time over the vessel irradiation rate. The overall integrated program also will provide infor-mation from surveillance programs in Crystal River Unit No. 3, and Davis-Besse Unit No.1 on material considered to be representative of the welds in the TMI-l vessel.

It is also important to note that still more information relevant to the TMI-l vessel materials will be obtained from the NRC sponsored Heavy Section Steel Technology (HSST) irradiation programs. Details are provided below:

Three weld materials are of primary interest for the TMI-l vessel.

Procedure Qualification (P.Q.) numbers

• WF 70 and WF 25 are used in the top and center

• Weld materials are specifically identified by the ASME Code by the procedure qualification test number. A procedure qualification test is required on each combination of heat of weld wire and batch of flux.

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. circumferential welds. The end of life (EOL) fluence for both of these welds is estimated to be 1.2 x 1019 nyt, and both have compositions that are Weld P.Q.

expected to make them relatively sensitive to radiation damage.

No. SA-1526, used for two of the longitudinal welds, also has high copper.

Further, the EOL fluence at the azimuthal locations of these longitudinal welds is 9 x 1018 so they may become limiting during the service life. Another shell weld, the lower circumferential, is made of a material that is expected to be radiation sensitive (P.Q. No. WF 67), but the E0L fluence at this location is estimated to be at least an order of magnitude lower than that of the other circumferential welds, so it is not expected to be limiting.

The original TMI-1 surveillance material, WF 25, was the same as the center circumferential weld, so information on its radiation behavior is already Two available from the test results from one capsule irradiated in TMI-1.

additional capsules containing this weld material will be irradiated in THI-2 in accordance with the proposed program.

In addition to this TMI-1/TMI-2 integrated program, "research" capsules containing tensile, Charpy (Cy) and two sizes of compact tensile (CT) spe-cimens of B&W archive materials will be included in the overall B&W power reactor surveillance program.

Weld materials included in the "research" capsules that may be limiting for the THI-l reactor vessel are: PQ SA-1526, PQ WF-25 and PQ WF-70.

In addition, specimens of WF-67 material will also be included in the "research" program.

These capsules will be irradiated in the Davis-Besse 1, Crystal River Unit No. 3 and TMI-2 reactor vessels. The presently planned withdrawal schedule calls for the first capsules to be withdrawn in 1981 to 1982, at which time the upper shelf energy of the specimens is predicted to be about 50 ft-lbs.

Other "research" capsules will be withdrawn in about 1989 when the specimens will have received a fluence approximately equal to that at the inner surface of the vessel at EOL.

Also, research programs being sponsored by NRC will provide additional information on the effect of radiation on these specific weld materials and on several additional B&W weld materials expected to respond to radia-tion in a similar manner. These programs, HSST-2 and HSST-3, consist of many tensile, Cv, and CT specimens irradiated in a test reactor. Although information on the shift in RTNDT will be obtained, the main emphasis of the HSST programs is to develop methods that can be used to better evaluate low shelf toughness using the rather small specimens used in the power reactor programs.

The following table shows where samples of the pertinent weld materials will be irradiated in the proposed integrated and "research" programs, what kinds of specimens will be used, and when the information will be available according to the present schedule.

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Capsule Irradiation Information Specimen Weld Designation

• Locations Available.

Types **

WF-70 R-1 Davis-Besse 1981 Cv, CT Unit No. 1

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R-2 Davis-Besse 1989 Cv, CT Unit No. 1 R-1 Crystal River 1982 Cv, CT Unit No. 3 R-2 Crystal River 1989 Cv, CT Unit No. 3 HSST-3 Test Reactor 1978 Cv, CT WF-25 TMI-lE THI-1 1976 Cv, Tenstle (already removed)

TMI-1A THI-2 1982 Cv, Tensile TMI-1C THI-2 1990 Cv, Tensile R-1 THI-2 1982 Cv, CT R-2 TMI-2 1989

' Cv, CT HSST-2 Test Reactor 1977 Cv, CT HSST-3 Test Reactor 1978 Cv, CT SA-1526 R-1 TMI-2 1982 Cv, CT R-2 THI-2 1989 Cv, CT The following welds are not controlling for TMI-1:

WF-67 R-l Davis-Besse 1981 Cv, CT Unit No. 1 R-2 Davis-Besse 1989 Cv, CT Unit No. 1 R-1 Crystal River 1982 Cv, CT Unit No. 3 R-2 Crystal River 1989 Cv, CT Unit No. 3 WF-8 SA-1494 The irradiation schedule and withdrawal dates shown may be modified on Note:

the basis of initial test results to optimize the scheduled availability of information.

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• & ** See top of page 6 Juu

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• R-1, R denotes "research" capsule THI-lE - denotes capsule E from Three Mile Island Unit No. 1 HSST-2, denotes capsule irradiated as part of the HSST program.

• • Cy - denotes Charpy V-notch specimen.

CT - denotes compact tensile specimen.

We have evaluated the effectiveness of this overall program plan, and have concluded that the informatien to be developed that is directly and indirectly relevant to the TM:-l reactor vessel will be sufficient to provide assurance of safety margins against vessel fcilure that comply with Appendix G,10 CFR 50.

Until data become available from the surveillance program, a conservative prediction of radiation damage can be made by using R.G.1.99* for at least the next five years of operation.

This Regulatory Guide is based on the NRC staff's analysis of all data available at the time it was written.

New data, in particular the results of the augmented integrated surveillance program described above, will be used to periodically update the Regulatory Guide. Predictions of the adjustment of reference temperature and the drop in upper shelf energy are given graphically in R. G.1.99 as functions of copper and phosphorus content and of fluence.

In addition there is an " Upper Limit" line on each graph, which is to be used when information about the copper and phosphorus contents is inadequate.

Because the chemical analyses of the B&W welds have shown considerable variation, the NRC staff intends to use the Upper Limit lines as the basis for any predictions required at this time.

We have also considered the uncertainties involved in applying radiation effects information obtained in other reactors to the TMI-l vessel. The major uncertainties involved are:

1.

Accuracy of neutron fluence calculations; 2.

Magnitude and effect of variation in neutron spectra between reactors; 3.

Magnitude and effect of variations in irradiation temperature between reactors; 4.

Magnitude and effect of variations in rate of irradiation on material properties.

• Regulatory Guide 1.99, " Effects of Residual Elements on Predicted Radiation Damage to Reactor Vhssel Materials", July 1975.

Revision 1 is to be published in April 1977.

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. The effects of these variables have been studied for at least 20 years.

Although some uncertainties still remain, the effects are fairly well established and understood as discussed below.

1.

Calculational methods for estimating the neutron flux at the reactor vessel wall and at irradiation capsule locations have been developed over many years.

The dosimetry used in irradiation capsules has furnished information that was used to check out and refine the calculational methods. As a result, the fast neutron flux and fluence in these locations can generally be calculated to an accuracy of + 20%, particularly if some dosimetry checks are available.

DosTmeters from the original TMI-l surveillance program were removed and tested, so the fluence calculations for the vessel have been verified.

In addition, it should be noted that the effect of neutron radiation on reactor vessel steel varies as the square root of the fluence; hence, uncertainties of 20 to 50% in fluence are not highly significant.

We have also considered the fact that the design of the TMI-l vessel, internals, and core is nominally identical to that of the other reactors which will be used to obtain radiation effects information.

These considerations are the basis for our conclusion that uncertainties in the calculation of neutron fluence will be small, and the effect of such uncertainties on the assessment of the radiaticn effects on the vessel material will also be small.

2.

Although differences in neutron energy spectra can cause uncertainties in the effects of radiation on material when this is evaluated without considering spectrum effects, only very large differences in spectra are significant. The variations from one B&W reactor to another are stated to be relatively minor, because they have similar geometry.

We considered the possible differences in neutron spectra that could occur between the B&W power reactors involved in the integrated program.

Such effects can be dealt with, if necessary, through methods that are being developed for that purpose.

However, the worst expected differences are judged inconsequential based on present knowledge of irradiation effects.

The neutron spectrum uncertainty will be kept under active scrutiny by the NRC staff and if additional developments (theoretical or experimental) suggest that the effect might be significant under some conditions, appropriate adjustments in reference temperature, drop in upper shelf energy or other suitable parameter can be made.

3.

The effect of the temperature of irradiation has also been the subject of considerable reserach.

It is well known that radiation damage is less severe at 6000F than at 5000F (the temperature range of concern). The differences in effect on the steel appear to be noticeable and should be taken into account if the irradiation temperature difference is over about 250F.

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integrated program are expected to be less than 50 F, and can be conservatively evaluated.

4.

The effect of irradiation rate has been evaluated by research programs at the Naval Research Laboratory (NRL) and other laboratories. Although the consensus of experts on this subject is that there will be no major differences in material property changes by irradiation rates varying over 2 to 3 orders of magnitude, more data from surveillance programs are needed to provide verification.

However, the differences in the rates of irradiation of specimens in the integrated program and the limiting material in the walls of the affected vessels will be less than one order of magnitude.

There-fore, we have concluded that there will be no significant uncertainties in this program associated with differences in rate of irradiation.

To implement the proposed program for TMI-1, the licensee had proposed Technical Specifications which would require the remaining T:4I-l capsules to be irradiated in TMI-2 on a schedule consistent with the overall integrated and augmented surveillance program.

We have found that certain revisions should be made in the proposed Technical Specifications.

These revisions have been discussed with and found acceptable by the licensee.

In addition, to assure that timely accumulation of neutron exposure by the specimens is not unduly affected by extended outages at TMI-2, we have suggested to the licensee, additional Technical Specifications which would require the proposal of alternate irradiation plans if irradiation of the specimens was not being accomplished in a timely manner at TMI-2.

The licensee has agreed to these suggested additional Technical Specifications.

Finally, it is noted that the proposed overall integrated, augmented program (with possible minor modification yet to be finalized) should provide more useful information than could have been extracted fram the original surveillance program.

The proposed program will ala provide information of the type needed to meet the requirements of Paragraph V.C of Appendix G,10 CFR 50.

Conclusion Regarding Reactor Vessel Material Surveillance Program We have evaluated the adequacy of the proposed integrated, augmented reactor vessel material irradiation program for TMI-l as an alternative to the original program that was interrupted by failure of the associated hardware. We conclude that the proposed program will provide the information required to comply with Appendix G,10 CFR 50, and that the degree of commonality between TMI-l and TMI-2 and the predicted severity of irradi-ation is such that the uncertainties involved in using data obtained from sur-veillance specimens irradiated in Three flile Island Unit 2 to establish TMI-l vessel operating limitations are small and can be accounted for

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We also conclude that the by imposition of appropriate margins.

associated Technical Specification changes to implement the program

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are acceptable.

2.

Fuel Reload for Cycle 3 Operation Reload Description The TMI-l reactor core consists of 177 fuel assemblies, each with a The reloading for Cycle 3 operation

'15x15 array of fuel rods.

(2, 3) consists of the removal of all batch 2 assemblies, the relocation of batch 3 and 4 assemblies, and the introduction of 13 batch la and 48 new batch 5 assemblies. The batch 5 assemblies will be located at the core periphery and the batch la assemblies will occupy 13 positions within the mixed central zone.

Fuel Mechanical Desian The outside dimensions and configuration of the Mark B-4 (batch 4 & 5) fuel assemblies and older Mark B-3 (batch 3) fuel assemblies are identical except that the Mark B-4 have spring-type flexible spacers and the Mark B-3 have corrugated-type flexible spacers.

This Mark B-4 fuel rod spacer has been previously reviewed and found acceptable by the NRC staff on the basis of no significant mechanical or material change to the reactor operation (4) and has been successfully operating in similar cores for a substantial time (Section 4.5 of Reit.ence 1). The Mark B-4 fuel assemblies, therefore, do not represent any unreviewed or untested change in mechanical design from the refere.'ce cycle and are therefore acceptable.

This mechanical design change has been taken into account in the various analyses which are discussed in the following sections.

The results of these analyses have shown that this fuel design difference in the TMI-l core is of negligible effect.

Fuel rod cladding creep collapse analyses were performed for the The CROV computer code was used to calculate the Cycle 3 core.

time to fuel rod cladding creep collapse. (1, 5) The calculational methods, assumptions, and data have been previously reviewed and The analysis assumed a 2000 hour0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> approved by the NRC staff (6).

densification time which maximizes creep; no fission gas production which maximizes differential pressure; and a lower tolerance limit on clad thickness and an upper tolerance limit on cladding ovality, both of which maximize cladding creep deformation.

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. The batch 3 fuel was found to be more limiting than the batch 4, 5, and la fuel due to the lower prepressurization, lower pellet The most limiting assembly density, and previous power history.

in batch 3 was found to have a collapse time longer than the maximum projected threc-cycle core exposure (24,288 EFPH).

From the viewpoint of cladding stress due to differential pressure, thermal stress due to fuel temperature gradients, and bending stress, neither the yield stress nor the B&W 1% total strain criterion for the cladding is predicted to be exceeded in the Cycle 3 core.

The batch 5 fuel assembly design is based upon established concepts and utilizes standard component materials. Therefore, on the bases of the analyses presented and previously successful operations with equivalent fuel, we conclude that the fuel mechanical design for Cycle 3 operation is acceptable and its application to Cycle 3 operation will not endanger the health and safety of the public.

Fuel Thermal Design The fuel thermal design analysis was conducted with the TAFY-3 computer code, as discussed in reference 7.

The analysis considered ensification, as the effect of a power spike from fuel pelle)8(1 Modifications to modeled in the " Fuel Densification Report".t the " Fuel Densification Report" on the fuel pellet void probability, and fuel grain size distribution, Fk ') ave been previously Fq,iewed and approved by the NRC staff. 19) rev Based on the analyses presented in reference 1 and comparison with allowable Linear Heat Generation Rate (LHGR) for fuel centerline melt considerations, (12) the fuel thermal design for the TMI-l Cycle 3 core is acceptable and can be applied with reasonable assurance that the health and safety of the public will not be endangered.

Fuel Material Design The fuel material design for Cycles 2 and 3 operation is not significantly different from that of Cycle 1 operation.

The only difference is that Zircaloy-4 is used as the fuel assembly tubular spacer material in Mark B-4 fuel instated of zirconium dioxide (Zr02), which is used in Mark B-3 fuel. This change does not affect the fuel system chemistry. This change has been reviewed and has a substantial amount of previous experience (Section 4.5 of reference 1). Therefore, the fuel material design for TMI-l Cycle 3 operation is acceptable.

Nuclear Design The TMI-l reactor has completed two operating cycles and is thus sufficiently close to equilibrium cycle to show only minor changes in physics parameters.

The Cycle 3 core will consist of four distinct fuel types:

fresh batch 5 assemblies located at periphery, once-burned batch 4 assemblies located generally in

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, an intermediate zone and also near the core center, twice-burned batch 3 assemblies located between the periphery and the inter-mediate zone, and located between the intermediate zone and central zone, plus 13 batch la assemblies loaded with the batch 4 assemblies. Thus, although the Cycle 3 core is a four batch loading, the physics parameters are quite close to those of the Cycle 2 core.

In addition, these parameters will be verified during the startup testing program described later.

The only significant procedural change from the reference cycle (Cycle 2) is the specification of axial power shaping rod (APSR) position limits.

The APSR position limits will provide additional control of power peaking through an improved definition of the core power distribution.

The calculational methods used by the licensee arc the same as were used for cycle 2.(10)

Because of this, and because of the verifi-cation provided by the physics testing which will be performed during the Cycle 3 startup, we find the nuclear design for Cycle 3 to be acceptable.

Thermal-Hydraulic Analysis Major acceptance criteria for the thermal-hydraulic design are specified in the NRC's Standard Review Plan Section 4.4 ("Thernal and Hydraulic Design"). These criteria establish the acceptable limits for DNBR (Departure from Nucleate Boiling Ratio).

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thermal-hydraulic analyses for the TMI-l Cycle 3 reload core were made with prevTdusly approved models gnd methods, as stated in the TMI-1 Final Safety Analysis Report.tll)

The reactor coolant flow rate was accurately measured during Cycle 1 operation and a minimum measured value of 108". of the system design flow was determined.

The licensee has taken credit in Cycles 2 and 3 thermal-hydraulic analyses for the fact that the actual system flow and conservatisms in this analysis.(i,and has also included uncertainties is greater than the design flow rate

10) The new design flow is 106.5% of the Cycle 1 design flow.

In the past, a reactor coolant flow penalty had been assumed in the thermal-hydraulic design analysis for TMI-1. This penalty was associated with the potential for a core internal vent valve to be stuck open during normal operation.

The core internal vent valves are incorporated into the design of the reactor internals to preclude potential vapor lock during a postulated cold-leg break Loss-of-Coolant Accident (LOCA).

We have concluded that by application of a surveillance program the vent valve flow ?enalty may be remnved. The sprveillance requirements demonstrate that the vent valves are not stuck open and that the vent valves operate freely.

A separate review of the licensee's surveillance program for the j

vent valves has concluded that the program adequately meets our requirements, and that the vent valve penalty was properly I'O

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JVO eliminated (12).

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. The effect of fuel rod bow was evaluated by the licensee with consideration given to both the hot channel power spike and the effect on DNBR. This evaluation was also separately reviewed and accepted by the NRC staff (12).

There are differences in the flow resistant.e between the Mark B-3 fuel assemblies and the Mark B-4 assemblies. The flow resistance for a Mark B-4 fuel assembly is sliv ly less than that for the For the Cycle.i loading, the highest assembly),

Mark B-3 assemblies.

The Cycle 2 analysisllU power always occurs in a Mark B-4 assembly (.also used for Cycle 3 refere assembly to be a Mark B-3 type. This analysis is conservative for Cycle 3 because the predicted hot assembly coolant flow rate is less than that of a corresponding Mark B-4 as::cebly.

Because of the analyses discussed above, we have found the thermal-hydraulic analysis to be acceptable and the proposed Technical Specifications related to the thermal-hydraulic analysis also acceptable.

Accident and Tre1sient Analyses A generic LOCA analysis for a B&W 177 assembly lowered-loop plant has been performed using the Final Acceptance Criteria ECCS evaluation model (13, 3).

This analysis has been reviewed by the NRC staff (14), and found applicable to the TMI-1 Cycle 3 core.

All other accidents and transients (loss of flow, dropped rod, inadvertent bank withdrawal, etc) have been examined by the licensee for Cycle 3 and found to fall within the bounds of the FSAR analyses, as updated for Cycle 2 operation. We have reviewed the various input parameters for Cycle 3, and have found the licensee's conclusion acceptable.

Startup Program The licensee has proposed a startup program which will verify:

Critical boron concentration.

Temperature reactivity coefficient at two point'.

Control bank worth by boron swap.

More than half the required shutdown reactivity will be verified using this method.

Control bank worth by bank drop.

The remainder of the banks will be checked by this method.

Ejected rod worth.

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In addition, during the power escalation phase, the startup program will verify:

power distribution at three plateaus.

Dropped-rod power distribution.

Incore/excore imbalance correlation.

Doppler coefficient at 100% power.

Moderator temperature coefficient at 100% power.

We have reviewed this proposed startup program and have found it acceptable.

Technical Specifications The licensee nas proposed revisions to the Technical Specifications to implement the changes due to the Cycle 3 reloadtll. We have reviewed the revised Technical Specifications relating to Cycle 3 operation and found them acceptable except in two instances:

(1) the Technical Specifications did not define an acceptable range of values for the moderator temperature coefficient below 95% power, and (2) the intervals at which core power maps would be obtained required These deficiencies were brought to the attentign clarification.

of the licensee who proposed revised Technical Specifications (3) which we find acceptable.

Conclusion Regarding Fuel Reload for Cycle 3 Operation Based on our evaluation of the application as set forth above, we conclude that the proposed changes in Technical Specifications associated with operation in Cycle 3, as revised, are acceptable.

Based on the foregoing evaluation of fuel and nuclear design, thermal-hydraulic performance and accident analyses, we also conclude that the operational characteristics of TMI-1 in Cycle 3 will be substantially the same as those in Cycle 2 and that operation of TMI-l as proposed, will not involve a significant increase in the probability or con-sequences of accidents previously considered nor involve a significant decrease in safety margin. Accordingly, we have concluded that operation of TMI-l does not involve a significant hazards consideration.

3.

Pressure Limits Durina Reactor Coolant System Heatuo and Cooldown By letter dated February 23, 1977, the licensee proposed reactor coolant system (RCS) pressure-temperature operating limits to be applicable thrcugh six effective full power years (EFPY) of operation.

The calculated maximum fluence on the vessel wall intgrior surface at the end of 6 EFpY was stated to be 3.2 x 1018 n/cm'.

There are six different weld materials in the reactor vessel beltline region. These are designated WF-8, WF-25, WF-67, WF-70, SA-1494 and SA-1526. Based on the chemical compositions and locations of these welds in the beltline, WF-25 WF-70 and SA-1526 are expected to be the most limiting materials.

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Weld material WF-25 is included in the THI-l material surveillance program. It contains approximately 0.34% copper and 0.015% phosphorus, and its initial reference temperature (RTNDT; was -14*F.

Capsule THI-lE, containing this weld material, was removed from THI-l following the first operating cycle. The dosimeters installed in this capsule indicated that the specimens had received a fast fluence of approximately 1.07 x 10 8 n/cm2 during their exposure. Examination of the WF-25 1

specimens which had received this fluence indicated that the upper shelf energy of the material had decreased from 81 ft-lbs to 62 ft-lbs, and that the shif t in RT DT, measured at 50 ft-lbs, was ll70F. Both N

of these experimental values agree clossly with the values predicted by Regulatory Guide 1.99.

The other possibly limiting weld materials are WF-70 and SE-1526.

WF-70 contains approximately 0.27% copper and 0.014% phosphorus and an estimated initial RTNDT of 20*F.

SA-1526 contains approximately 0.36%

cepper and 0.016% phosphorus and also has an estimated initial RTNDT of 20*F.

Based on the composition, initial RT DT and neutron fluence at the N

locations of the various beltline materials, the licensee estimated, using the curves in Regulatory Guide 1.99, Revision 1*, the adjusted RT DT at N

6 EFPY for each of these matertals. From these adjusted values of RTNDT.

the licensee then developed the proposed operating limit curves. The most limiting material according to the licensee's analysis was SA-1526 which had an adjusted RTNDT of 145*F at the end of 6 EFPY.

We have reviewed the licensee's submittal and have determined that the for material WF-70 was estimated by the licensee based adjusted RTNDT on the measured cepper content of this material. However, in view of the variances we have found in the chemical composition of some welds (particularly copper content), we believe the upper limit lines of Regulatory Guide 1.99 should be used to predict changes in RTNDT-When this is done for the WF-70 material, it is found to have an adjusted RTNDT of 160*F at 6 EFPY and, thus, is more limiting than the SA-1526 used by the licensee.

At 5 EFPY, however, WF-70 would have an adjusted RTNDT of %:s than 145'F. Therefore, since the proposed operating limit curves were based on a maximun adjusted RTNDT of 145*F, reduction of the term of applicability of the curves from 6 EFPY to 5 EFPY would allow the curves to conservatively represent all of the limiting materials -

including WF-70. Such a reduction has been discussed with and found acceptable by the licensee.

To be pubHshed April,1977.

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Conclusion Recardino Revised pressure-Temperature Limit for Heatup and Cooldown Based on the above considerations, we conclude that the proposed operating limit curves, limited in applicability to 5 EFPY, are acceptable and conform to the requirements of 10 CFR 50, Appendix G, No changes were proposed for Technical Specification 3.1.3, " Minimum Conditions for Criticality". This specification requires that the reactor coolant temperature be above 525*F prior to criticality except for low power physics tests.

This specification meets the requirements of paragraph 4.A.2.c of Appendix G,10 CFR Part 50 and is acceptable.

Confonnance with Appendix G,10 CFR Part 50 in establishing safe operating limitations will ensure adequate safety margins during operation, testing, maintenance and postulated accident conditions and constitutes an acceptable basis for satisfying the requirements to NRC General Design Criterion 31, Appendix A, 10 CFR Part 50.

Because this proposed change is in conformance with Appendix G to 10 CFR 50 which is designed to maintain an adequate safety margin, we conclude that the proposed change, as revised, does not involve a significant increase in the probability or consequences of accidents previously considered and does not involve a significant decrease in safety margin. We therefore, further conclude that the proposed change, as revised, does not. involve a significant hazards consider-ation.

Environmental Consideration We have determined that the amendment does not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact. Having made this determination, we have further concluded that the amendment involves an action which is insignificant from the standpoint of environmental impact and pursuant to 10 CFR 551.5(d)(4) that an environmental impact statement or negative declaration and environmental impact appraisal need not be prepared in connection with the issuance of this amendment.

Conclusions We have concluded, based on the considerations discussed above, that:

(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in tne proposed manner, and (2) such activities will be conducted in compliance with the Commission's reculations and the issuance of this amendment will not be. inimical to the common defense and security or to the health and safety of the public.

Dated: April 22, 1977 15G8 337

Q References 1.

Letter, R. C. Arnold (Metropolitan Edison) to Director of Nuclear Reactor Regulation, dated January 26, 1977, enclosing Technical Specification Change Request No. 45 and BAW-1442.

2.

BAW-1442, Three Mile Island Unit 1 Cycle 3 Reload Report, November, 1976 3.

Letter, R. C. Arnold (Metropolitan Edison) to Director of Nuclear Reactor Regulation, dated March 31, 1977, enclosing responses to Round 1 Questions concerning the TMI-1 Cycle 3 Reload Application.

4.

SE on Oconee Nuclear Station, Units 1,2,&3, dated June 30, 1976, Amendment Nos. 27, 27, and 23 for License Nos. DPR-38, DPR-47, and DPR-55.

5.

BAW-10084P, Rev. 1, Prccram to Determine In-Reactor Performance of B&W Fuels - Claddina Creep Collapse, October, 1976 6.

Letter, A. Schwencer (NRC) to J. F. Mallay (B&W), dated January 29, 1975 7.

BAW-10044, TAFY - Fuel Pin Temoerature and Glass Pressure Analysis, May, 1972 8.

BAW-10055, Rev.1, Fuel Densification Report, June,1973 9.

Memorandum from R. Lobel to D. F. Ross, "Present Status of B&W Power Spike Model," July 23, 1974.

10. Three Mile Island Unit 1 - Cycle 2 Reload Report, Revision 1, July 1,1976.

11. Three Mile Island Unit 1 Nuclear Station, Final Safety Analysis Recort, USNRC Docket No. 50-239.

12 Safety Evaluation by the Office of Nuclear Reactor Regulation, Amendment No. 25 to Facility Operating License No. OPR-50, dated March 7, 1977.

13. BAW-10103, Rev. 1, ECCS Analysis of B&W's 177-FA Lowered Looo NSS, September, 1975.

14. Letter, D. F. Ross (NRC) to D. B. Vassallo (NRC), Re: Topical Report Evaluation BAW-10104, ECCS Evaluation Model. Revised Nucleate Boilina Lockout Model, dated February 2,1977.

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UNITED STATES NUCLEAR REGULATORY COMMISSION DOCKET NO. 50-289 METROPOLITAN EDISON COMPANY JERSEY CENTRAL POWER AND LIGHT COMPANY PENNSYLVANIA ELECTRIC COMPANY NOTICE OF ISSUANCE OF AMENDMENT TO FACILITY OPERATING LICENSE The U. S. Nuclear Regulatory Commission (the Comission) has issued Amendment No. 29 to Facility Operating License No. DPR-50m issued to Metropolitan Edison Company, Jersey Central Power and Light Comoany and Pennsylvania Electric Company (the licensees), which revised Technical Specifications for operation of the Three Mile Island Nuclear Station, Unit No.1 (TMI-1) located in Dauphin County, Pennsylvania. The amendment is effective as of its date of issuance.

This amendment revises the Technical Specifications to:

1) permit irradiation of TMI-l reactor vessel material surveillance specimens in the Three Mile Island Nuclear Station, Unit No. 2 reactor vessel; 2) reflect plant operating limitations for the fuel loading to be used during Cycle 3; and 3) update the reactor coolant system pressure limits during system heatup and cooldown.

The applications for the amendment comply with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations. The Commission has made appropriate findings as required by the Act and the Commission's rules and regulations in 10 CFR Chapter I, which are set forth in the license amendment.

Notice of Proposed Issuance of Amendment to Facility Operating License in conne: tion with Item 1, above, was published in the FEDERAL

,O p 7'll f22f WV 1533 539

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r REGISTER on February 3, 1977 (42 F.R. 6652). No request for a hearing or petition for leave to intervene was filed following notice for this action. Prior public notice of Items 2 and 3, above, was not required since they do not involve a significant hazards consideration.

The Commission has determined that the issuance of this amendment will not result in any significant environmental impact and that pursuant to 10 CFR 551.5(d)(4) an environmental impact statement or negative declaration and environmental impact appraisal need not be prepared in connection with issuance of this amendment.

For further details with respect to this action, see (1) the applications for amendment dated October 29, 1976, as supplemented December 29, 1976 and January 20, 1977; January 26, 1977, as supple-mented March 31, 1977; and February 23,1977,(2) Amendment No. 29 to License No. OPR-50, and (3) the Commission's related Safety Evaluation. All of these items are available for public inspection at the Commission's Public Document Room, 1717 H Street, N.W.

Washington, D.C. and at the Government Publications Section, State Library of Pennsylvania, Box 1601'(Education Building), Harrisburg, Pennsylvania.

A copy of items (2) and (3) may be obtained upon request addressed to the~ U. S. Nuclear Regulatory Commission, Washington, D.C. 20555, Attention: Director, Division of Operating Reactors.

Dated at Bethesda, Maryland, this 22nd day of April 1977.

FOR THE NUCLEAR REGULATORY C0t' MISSION

k f-Lp Robert W. Reid, Chief Operating Reactors Brar.ch d4 Division of Operating Reactors 1588 340