ML20153G967

Corrected Safety Evaluation Supporting Amend 97 to License DPR-61

ML20153G967
Person / Time
Site:	Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date:	04/28/1988
From:	Office of Nuclear Reactor Regulation
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ML20153G944	List: ML20153G942 ML20153G959 ML20153G967
References
NUDOCS 8805120074
Download: ML20153G967 (21)

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UNITED STATES NUCLEAR REGULATORY COMMISSION

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CORRECTED SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMEN 0 MENT NO. 97 TO FACILITY OPERATING LICENSE NO. OPR-61 CONNECTICUT YANXEE ATOMIC POWER COMPANY HA00AM NECX PLANT 00CXET NO. 50-213

1.0 INTRODUCTION

By letter dated June 1,1987, as modified by letter dated July 22, 1987, the Connecticut Yankee Atemic Power Ccmpany.(CYAPCO, licensee) submitted a request for changes to the Haddam Neck Plant Technical Specifications (TS).

The license amendment supports operation of the Haddam Neck Plant for Cycle

' 15 and reflects major efforts in upgrading design basis analyses.of accidents and in refomatting the existing technical. specifications as part of the

).'.anned conversion to the Westinghouse standard technical specifications.

More specifically, the license amendment deletes existing Technical Specifi-cations 3.15, "Reactivity Ancmalies"; .'.16. "Isothemal Coefficient of Reactivity"; 3.18, ' Power Distribution Monitoring and Control"; and 3.20, 1

' Reactor Coolant System Flow, Temperature and Pressure." The infomation contained in those specifications has been reorganized along with additional limiting conditions of operation, action statements and surveillance requirements, which are. consistent with the Westinghouse standard technical specifications, into revised Technical Specifications 3.3, "Reactor Coolant System Operational Components"; 3.10. "Reactivity Control"; and 3.11, "Limiting Linear Heat Generation Rate"; as required, to assure completeness with the existing technical specification. One new specification (3.24, "Special Test Exceptions") has been added to fomalize the special test exceptions required to perform various startup physics-tests.

In addition, existing technical specifications in sections 1.0, "Definitions";

2.2, "Safety Limits"; 2.4, "Maximum Safety Settings - Protective Instrument-ation"; 3.11. "Containment"; 3.13. "Refueling"; 3.5, "Chemical and Volume Control System'; 3.7, "Minimum Water Volume and Boron Concentration in the Refueling Water Storage Tank"; and 4.9, "Main Steam Isolation V:lves' have been revised to account for the revised design basis analyses in support of the safe operation of the Haddam Neck Plant for Cycle 15.

2.0 -tCXGROUND l

By letter dated June 1,1987, as modified by letter dated July 22, 1987, Connecticut Yankee Atomic Power Ccmpany submitted revisions to the Haddam Neck Plant TS to reflect planned operation during Cycle 15.

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2-l The non-LOCA accident analysis methodology, including th.e use of RETRAN and VIPRE, was submitted for staff review by letter dated August 1, 1984 .

I The physics methodology for Cycle 15 was provided for staff review by letter dated Septemeer 12, 1985. The plant design basis analyses using the referenced methodologies were submitted by letters dated June 30, 1986, March 10, and May 8, 1987.

The VIPRE methodology for thennal-hydr'aulic parameters was reviewed and a Safety Evaluation forwarded to CYAPC0 by letter dated October 16, 1986. 4 The core physics methodology has been reviewed and a Safety Evaluation forwarded to the licensee by letter dated August 3, 1987.

The staf' review and approval of the RETRAN transient cethodology and the limiting design bases transient analyses as they apply to Cycle 15 operation at the Haddam Neck Plant is described in Section 3.A of this Safety Evaluation.

3.0 EVALUA7t0N ,

A. Cycle 15 Technical Review

1.0 INTRODUCTION

r By letter dated June 1, 1987, CYAPCO submittee its safety evaluation to support Cycle 15 operation for the Haddam Neck Plant. The submittal included the proposed Technical Specification (TS) changes, a safety evaluation for the proposed TS changes and a technical report supporting ,

Cycle 15 operation in the areas of nucitar design and accident analyses.

The licensee's objective for this submittal was to demonstrate that the limiting conditions resulting from the rean+1yses of the non-LOCA design 3 basis accidents are properly reflected in the proposed TS and to support its position that the Haddam Neck Plant can be operated safely at the rated power level of 1825 MWt threughout Cycle 15.

2.0 EVALVATION .

I 2.1 Reload Description i

i The Cycle 15 core consists of 157 15x15 fuel assemblies, each containing 204 fuel rods, 20 control red guide tubes and 1 instrument guide tube.

All fuel pins have the same outside diameter of 0.422 inches. Four lead l

test assemblies contain zircaloy clad fuel rods (batch ISB) with a clad l wall thickness of 0.027 inches compared to 0.0165 inches for the stainless i

steel 304 clad fuel rods of all other assemblies. The four assembites containing zircaloy clad fuel rods were first loaded into the core for l Cycle 13. A LOCA analysis for these assemblies was performed at that time.

The vinimum uranium theoretical density is 94.9% for all Cycle 15 fuel.

The ititial uranium-235 enrichment for all new Cycle 15 assemblies is 4.0 percent. The assembly loading is of the out-in type, i.e., with the new assemblies loaded on the ptriphery of the core. This is a high neutron <

leakage loading.

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All 45 control rods are being replaced in Cycle 15 due to previously observed cracking and wear problems. The new control rods are neutronically identical with the old except for extended life features such as Inconel

cladding for improved wear resistance.

2.2 Fuel Design ,

The Haddam Neck core for Cycle 15 will contain fuel assemblies of previously irradiated batches 15A,150 and 16 and the 56 new assemblies of batch 17.

In addition, one once-burned assembly from batch 15C is to be reinserted 4

in the core's central position. All fuel assemblies are mechanically i

interchangeable. The four zircaloy clad assemblies (batch ISB) have a  ;

1. shorter fuel stack, thicker cladding, smaller diameter fuel pellets and higher prepressurization in order to give fuel performance equivalent to the 304 stainless steel clad fuel rods.

The evaluation of the fuel performance is discussed below for cladding -

collapse, stress, strain and fatigue.

Batch 15A is estirated to be the most limiting in tems of core exposure time and burnup to the end of Cycle 15. The power histories of all assemblies were analyzed 1,o determine the worst pcwer history for creep collapse.

All Cycle 15 assemblies were analyzed under the worst power history conditiens.

For all five fuel batches, analyses predict collapse times and burnups exceeding the maximum expected residence time and burnup in Cycle 15. For '

betch 15A in particular, the cladding collapse time is 31,500 effective full power hours (EFpH) while the design residence time to the end of Cycle ,

1 15 will be 28,190 EFPH. The analysis was perfomed with TACO 2, which is i l an approved code, and the staff has detemined that the cladding creep

! collapse analysis is acceptable.

I All fuel rods were evaluated for stresses following the ASME guide' lines '

' for pressure vessels which require that the primary membrane stress intensity must be less than two-thirds of the minimum unirradiated yield strength, i In all cases, the calculated stress values are acceptable. .

The fuel desi n criteria specify a limit of one percent or less on cladding

! plastic tenst a circumferential strain. The licensee perfomed an evaluation of the cladding strain using the TACO 2 code and assumed the worst Cycle 1$ 1 l

heat generation rate and fuel burnup. It was detemined that the design strain limit is alch higher than the predicted value for Cycle 15 and, l therefore, it is acceptable.

r l The fatigue usage factor was calculated following the ASME pressure vessel code which defines a maximum allowable factor of 0.9. Using conservative i

conditions for Cycle 15, the fatigue usage factor was found to be 0.2 for I the stainless steel 304 cladding and 0.4 for the Zircaloy, which are I acceptable.

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2.3 Nuclear Desien The Cycle 15 reload is the first reload which has been designed by- Northeast Utilities personnel for CYAPCO, one of their member utilities. The physics methodology is documented in the topical report NUSCO-152, "Physics Methodology for PhR Reload Design."

The major differences between the Cycle 15 design and the Cycle 14 design are:

(a) a 5'F higher vessel average operating temperature, and (b) a new set of control rods The higher operating temperature was adopted following the cold-leg resistance temperature detector (RTO) replacement and relocation during the 1986 outage (Cycle 14) to eliminate cold leg streaming effects.

Previous fuel cycles operated at a vessel average temperature about 5'F higher than the indicated value. CYAPC0 submitted an analysis for Cycle 15 which showed that a 5'F higher temperature at full power would have a negligible effect on the fuel cycle design.

All 45 control rods have been replaced due to previously observed cracking and wear problems. The new control rods are mechanically improved but they '

are neutronically identical with the old set. Calculations of their reactivity worth show them to be equivalent to the old ones.

The licensee has perfomed analyses to show that during Cycle 15, the limits of the moderator temperature ccef ficient, the Doppler temperature coefficient, the delayed neutren fraction, the prempt neutron lifetime and the maximum differential rod worth, would be acceptable. These limits are discussed in greater detail in Section 2.4 below.

The staff has reviewed the nuclear design submitted by CYAPC0 in support of Cycle 15 of the Haddam Neck Plant. We conclude that the methods used

' have been previously approved and that the results are within the range of Cycle 14 parameters. Therefore, we find the nuclear design of Cycle 15 to be acceptable.

j 2.4 Transient and Accident Analyses Several physics parameters for Cycle 15 will be more limiting than for Cycle 14. These parameters include reduced shutdowi margin,' increased differential rod worth, revised reactivity (Doppler and moderator tempera.

ture) coefficients, and revised axial peaking f actors. These These changes analyses are: t.w required that six accident analyses 'se revised.

uncontrolled rod withdrawal, dropped rod, main steam line break, boren dilution, loss of flow, and red cluster control assembly (RCCA) ejection.

The licensee stated that operation through 180 effective full power days (EFPD) of Cycle 15 is bounded by the results and assumptions of the l

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5-current design basis steam line break analysis which was submitted in 1980 and approved by the NRC. Therefore, approval of the revised steam line break analysis is not required for Cycle 15 startup. The staff has reviewed this pocition and has concluded that the existing SLB analyses would bound operation of the Haddam Neck Plant for the first 180 EFP0 of Cycle 15.

Operation beyond 180 EFPD requires an approved steam line break (SLB) analysis.

2.4.1 Red Withdrawal Transient -

Due to changes in some Cycle 15 parameters, CYAPCO performed a reanalyses

! of the rod withdrawal accident. Technical specifications are also proposed to require the startup rate trip to be operable whenever the reactor trip breakers are closed and the control red drive lif t coils are energized, up

! to the P-7 interlock. In addition, TS are proposed to recuire a different

! (more restrictive) number of operating RCS loops during subtritical conditions.

I Uncontrolled red withdrawal transient analyses were perfomed by NUSCO for four loop operation comencing at 100% power and 65t pcwer, and comencing at 55% power for three loop operation and at suberitical with and without a functional rod stop. In all suberitical transient cases, the licensee reasonably concluded that such transient would be terminated by the startup rate trip or by operator action before a significant power level was reached and, therefore, that fuel themal limits would not be challenged nor would DNBR limits be reached.

Parametric studies made for those transient cases starting from power included using both positive and negative axial offsets, varying the reactivity insertion rates; and using both maximum and minimum feedback.

The following specific reactivity assumptions were used: (i) least negative Doppler and most positive moderator temperature coefficient (MTC) vs most negative Doppler and MTC; (ii) highest RCCA stuck out; and (iii) reactivity insertion rates up to 22.5 percent milli K per second (pem/sec) (which was the maximum obtainable from any single or two sequer,tial control Banks in controlled overlap). Conservatisms were introduced by the use of limiting values of fuel rod conductivity, maximum core inlet temperature and minimum RCS pressure and flow.

Core power, Inlet temperature and flow, and primary pressure were used as input to VIPRE-01 in which the ONBR computations were made.

No parametric computations were done to verify the accuracy of the RETRAN nodalization for this analysis. Nevertheless, since (i) this is a particularly short transient in which only the primary pressure, core flow, and inlet temperature and pcwer are important, and since (ii) comparison by the licensee of the pressure and temperature computed by RETRAN to actual plant pressure and temperature data for a 30% load rejection and a partial loss of feedwater event was good, the staff has reasonable assurance that the licensee's cceputation of core flow, core inlet temperature and primary pressure are accurately predicted by the RETRAN model. Furthemore, the staff has reasonable assurance that core power was conservatively computed in this analysis.

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6-The licensee concluded that the minimum ONBR is always greater than 1.3 and, therefore, that fuel themal limits would not be exceeded. Based upon the foregoing, the staff has reasonable assurance that the fuel thermal limits would not be exceeded. In addition, the staff concludes that the licensee has imposed sufficient conservatism to the appropriate parameters such that the staff has adequate assurances of conservative results for the red withdra'wal analysis. .

2.4.2 Boron Dilution The reanalysis of the boron dilution accident was submitted because the shut-down margin is being reduced due to reload physics parameters. The licensee has stated that the maximum possible dilution rate is 180 gpm and perfomed a simpis ccmputation of the reactivity insertion rate for such a dilution rate. The licensee reached the conclusion that such a dilution rate is well within the reactivity insertion rates of the uncontrolled RCCA withdrawal analysis. Upon review of the licensee's submittal, the staff concurs with the licensee's conclusien that the minimum DNBR will not be challenged by the boren dilution accident.

The boron dilution analyses was used to establish shutdown margins which would enable the operator to have at least 15 minutes for the time frem the first safety alam until criticality for Modes 1 through 5 (i.e., all modes except the refueling mode) and 30 minutes in Mode 6 (the refueling mode), as required in Standard Review Plan (SRP) 15.4.6. These new shutdown margins have been reflected in proposed TS. However, according to the licensee, the steam line break accident beccmes the controlling accident for shutdown margin after 180 EFPD. The staff has reviewed this issue and

. agrees with the licensee's assessment. The staff will evaluate the forth-ccming reanalyses of the main steam line break accident to support plant operation after 180 EFPO. However, the current analyses provide reasonable assurances that the proposed shutdcwn margins are conservative for the period ,

between startup and 180 EFPO.

2.4.3 RCCA Ejection The RCCA ejection accident represents the potential Itccident with the most rapid reactivity insertion rate. The licensee's analytical methedology for this accident is similar to that described above for the uncontrolled red withdrawal accidents, with the licensee stating that they employed the following conservatists: (i) the limiting burnup dependent physics pars-meters were combined to generate the most severe system response, and (ii) point kinetics was used with no Doppler weighting multiplier. More specifically, the burnup dependent physics parameters used: (1) took no credit for flux flattening effects of reactivity feedback; used maximum bankinsertionateachpowerlevel;(2)assumedadversexenon;(3)added margins to ejected rod worth to account for calculational uncertainties; (4) Assumed fuel temperature feedback to be at its minimum value over the

, entire burnup range; (5) used the most positive MTC; and (6) used the j smallest delayed neutron fraction over the entire burnup range (to minimize time to prompt criticality). In addition, trip reactivity was computed without including the ejected rod and using trip and trip response delays.

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The licensee performed analyses for four loop hot full power (HFP) and hot zero power (HZP) and three loop HFP and HZP operations. The licensee ,

estimated that 18% of the fuel rods reached DNBR of less than 1.3, ano no rod had fuel melting at the center line in the four loop HZP case and none in the HFP cases.

, The staff has reviewea the licensee's anelyses and has cencluded that the licensee's analyses of the RCCA ejection'is acceptable.

2.4.4 Droceed Red Accident The licensee's analysis for this event used the same methodology as described for the RCCA ejection and rod withdrawal accidents described above. This transient is a power reduction transient for which the licensee used to have a turbine runback feature. Given the large number of plant transients caused by spurious dropped red signals, the licensee has decided to disable the turbine runback feature and, therefore, needed to reanalyze this transient.

Parametric analyses were conducted varying the dropped rod worth from 0 to 250 pcm (minimum to maximum expected values) and with turbine load runback l in manual and in automatic. The RCS was, in contrast to the reactivity insertion transient discussed above, assumed to be initially in conditions with maximum core inlet temperature, but with minimum primary pressure and core flow, in each case, intended to produce the minimum computed DNBR.

In addition, the most negative Doppler and MTC are used (except in the analysis of the transient with the automatic turbine runback. since there is no trip in that case) to maximize the power, thus, also tending to produce a lower DNBR.

The licensee concluded that the minimum DNBR was obtained for the full power four loop operation without crediting either the turbine runback or the rod l stop protective features, in that case, the licensee concluded that the minimum DNBR remained well above 1.3. On the basis of the foregoing, we conclude that there are adequate assurances that the dropped red accident ,

analyses for Haddam Neck are conservative and, therefore, acceptable.

2.4.5 Loss of Flow The licensee felt that the current TS limit for low RCS flow rate for four loop operation may not be met since there is very little available margin now and some additional steam generator tube plugging is anticipated. In order to justify the icwer preposed RCS ficw rate TS, the licensee submitted the less-of-flow (LOF) analysis assuming a lower core flow rate. In 1 addition, the new ar,alysis included a change of the three loop flew trip setpoint in the governing TS.

The licensee stated that the flow rate requirements were based upon a steam generator plugging level consistent with 500 eauivalent plugged tubes per steam generator, and that an evaluation has been performed with l

the bypass flow fraction equal to 4.5 percent l i

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8-The licensee further indicated that the minimum RCS flow rate used is 246,000 gpm (4-loop) and 194,000 gpm (3-loop) and that the core flew rate used in the VIPRE calculations was the same as that used in the RETRAN calculations.

The licensee analyzed the worst case LOF event (ccmplete loss-of-flow from full power), which results in the most severe power-to-flow ratio and, therefore, the lowest DNB. In Sectiert II.E of topical report NUSCO 140-1, "Thermal Hydraulic Model Qualification Volume 1 (RETRAN)." the licensee presented LOF sensitivity studies of the impact of variation of RCP inertia, junction inertia, rod insertion time, delay of scram signal and reduction of RCS flow loss coefficients. They concluded that variation parameters had a minimal impact upon CNBR calculated using the W-3L ccrrelation anJ, therefore, that normal values could be used.

Although no experimental data was presented to verify the flow coastdown curves for this transient, we believe that good comparison to the results in the Facility Description and Safety Analysis (FDSA) gives adequate assurances of accuracy of the RETRAN model to permit its use in this event.

1 The licensee conservatively assumed that the transient comences frem 102%

pcwer for the four loop transient and 67% for the three loop transient and the RCP inertia reduced by 10% to reduce RCS flow more rapidly. In addition, the licensee assumed minimum initial RCS pressure and flew but assumed the maximum core inlet temperature. Each of these assumptions is conservative, tending to lower the computed minimum DNBR. In addition, the computed RCS pressure rise was minimized by assuming minimum initial pressurizer level, maximum initial SG 1evel, maximum turbine stop valve closure time, the pressurizer heaters off, maximue pressurizer spray flow, and the charging systems isolated but letdown flew available. Finally, reactivity insertion was raximized through the use of the least negative Doppler ccefficient and the most positive MTC and the reactor trip ccmput-ation included instrument response and delay times.

CYAPCO concluded that the minimum DNBR was approximately 1.4 in the worst case LOF transient. Since the VIPRE-01 code has beert approved by NRC, and since the minimum DNBRs computed with VIPRE-01 using the RETRAN cutput described above are above 1.3 for both analyses, the staff has adequate assurance that the required DNBR limits will be met. Therefore, the staff

' concludes that this analyses is acceptable, l

In sumary, six transient and accident analyses have been reanalyzed by tne c

licensee and found to be c'enservative and acceptable. The assumptions and conditions used in the analyses are found to be consistent with the proposed TS. The steam line break accident will be reanalyzed by the licensee and may impact the acceptability of TS Sections 3.10.1.1, 3.10.1.2, and 3.10.1.4 during the second half of the Cycle 15 operation. In addition, use of RETUN models in these analyses is accepttble.

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B. Technical Scecifications Given that extensive modifications have been proposed to the format of the existing technical specifications, the staff has reviewed the proposed Cycle 15 reload technical specifications to assure that the proposed technical specifications (1) retain all previous requirements, as appro-riate, and (2) incorporate the appropriate parameters to reflect the revised safety analyses performed in support of Cycle 15 operation and which provide reasonable assurance of safe operation during Cycle 15.

- Table 1 contains a comparison of the existing technical specifications with the proposed new and refomatted technical specifications. This comparison assured that all existing requirements have been retained in the proposed refomatted technical specifications except for those which have been deleted as discussed below. The proposed mod.ifications to the technical specification sections are summarized below.

B.1 Section 1.0 "Cefinitions" ji Three changes have been proposed for Cycle 15 operation. The first is a W change to the definition of shutdown margin to reflect the Westir.ghouse methodology of calculating shutdown margin. Westinghouse methodology calculations for shutdown margin assume a core with all control rods

[' inserted except for the maximum worth rod. This is also consistent with current review practice in the staff's Standard Review plan.

The second proposed change is the definition of Operational M00ES. The definition of the reactivity condition (Xeff) for the refueling mode is being revised from the original technical specification value of Aeff 60.92 to a new value (Xefff 0.94). The new Keff value is based upon a revised boren dilution accident analysis which was perfomed in support of Cycle I' 15 operation and is more restrictive than the current Westinghouse TS (Xefff0.95). As discussed earlier, this accident has been reviewed by the staff and was found to be acceptable, i

In addition, this amendment proposes to revise the definition of Operational MODES 2-5. Even though the numbers in Table 1-1 change, the reactivity conditions for MODES 2-5 in the Westinghouse STS are identical to the reactivity cor.ditions proposed for Cycle 15. The reactivity condition definitions have been intended to identify whether the core is critical (MODES I and 2) or suberitical (HODES 3-5) within the traditional 1% delta K/K reactivity uncertainty allowance. Since shutdown margin may be provided by control rods actually inserted, conual rods available to insert and/or soluble boron, the core reactivity cendition cannot be equated directly to shutdown margin. The proposed TS change for Cycle 15 returns the reactivity condition to the original intent, and provides a contistent definition of shutdown margin and shutdown margin requirements that were identified by the non-LOCA accident reanalysis.

The staff has reviewed the three proposed modifications discussed above and concludes that they are consistent with the analyses submitted in The staff has found the submitted analyses support of Cycle 15 operation.

to be acceptable, as described above, and therefore the staff concludes that the proposed modifications are acceptable.

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4 B.2 Section 2.0, ' Safety Limits and Maximum Safety Settings" l I The licensee has proposed two changes to Section 2.2, "Safety Limits -

. Reactor Core." The first change is to provide Applicability and Action j i Statements to this specificatien. The existing specification currently a does not include these items. The staff has reviewed these additiens and i

found them to be consistent with plant operation and the Westinghouse Standard Technical Specifications (STS) and, therefore, concludes that the t j proposed additions are acceptable. -

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! The second change to Section 2.2 are proposed revisions to Figures 2.1 1 ,

and 2.1-2. The above figures have been revised to account for the use of  :

the staff approved V! PRE code as well as a measured reduction in reactor  !

coolant system (RCS) flow rate. The minimum departure frem nucleate toiling ratio (MDMR) limit and the co e exit void fraction limit values remain unchanged. The Bases for Specification 2.2 have also been revised to ,

reflect the use of the newer ccde (methodology) and includes a discussion '

of tha design peaking factors for both three- and four-loop operation.

4 The staff has reviewed the proposed changes and concludes that the licensee  :

! has used previously approved methcdology in calculating the revised figures.

l Further, the licensee has assured that all previously approved design .

i margins have not been reduced. Therefore, the staff concludes that the  !

proposed modifications are acceptable.  :

The licensee has also proposed three modifications to Section 2.4, "Maximum  !

, Safety Settings - Protective Instrumentation." The first proposed modifi-cation is a lowering of the reactor trip setpoint for Icw coolant flew.

3 l The lower setpoint is directly obtained frem the reanalysis of the less-of-  !

! flow accident. The staff has reviewed the revised less-of-flow accident  :

l and fcund the analyses to be acceptable (see Section 3.A.2.4.5 above). The l

Itcensee has also added a new trip setpoint for a high start-up rate reactor j trip. This trip is from safety-related equipment and is necessary to assure '

'I that assumptions made in the revised analysis of the control rod withdrawal i from suberitical accident are preserved. The revised control rod withdraw &1 ,

i frem suberitical accident for Cycle 15 has been reviewed and found to be .

j acceptable (see Section 3.A.2.4.1 above). >

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The last modtfication to this section is the deletion of the requirement <

1 for determining the quadrant pcwer tilt ratio (QPTR). The requirement to i

! maintain a QPTR less than 1.02 has been celeted from Section 2.4 but has i been preserved as Specification 3.17.4 in the revised fonnat. t l The licensee has also proposed revisions to the Bases section of Specifica-

tion 2.4 to reflect the revised accident analyses and the addition of a new

! L reactor trip.

! The staff has reviewed the proposed mcdifications to Sections 2.2 and 2.4 [

j and the Bases to Section 2.4. The proposed modifications are consistent l with the revised analyses which have been reviewed by the staff and found l to be acceptable. Therefore, the str.ff concludes that the proposed redifica- ,

tiens to Section 2.0 are consistent with the analyses in support of Cycle 15 (

operation and are acceptable.

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B.3 Specification 3.3, "Reactor Ceolant System" Proposed Specification 3.3, "Peactor Coolant Sp tem," replaces in entirety the current Specification 3.3, "Reactor Coolant System Cperational Com-ponents," with several exceptions as described below. As stated earlier, Tabit 1 contains a one for one listing of the existing requirements of tne current TS and the refomatted and renumbered section in the proposed TS for Cycle 15. Review of this table has &ssured that all existing specifica-tions, surveillanc6 requirements and a'ction statements, have been preserved in the proposed refomatted technical specifications with four exceptions.

These are existing Specifications 3.3.C.2, 3.3.C.3, 3.3.C.5 and 3.3.H.

The bases for their deletion are discussed below.

Existing Specifications 3.3.C.2, 3.3.C.3 and 3.3.C.5 specify a required number of reactor coolant pumps and steam generators operating above certain power levels. These specifications are not consistent with the design reanalyses perfomed in support of Cycle 15. Further, the number of reactor coolant loops to be operating has been specified by Mode Applicability statements. These requirements have been retained in proposed Specification 3.3.1.1 for MODES 1 and 2. Proposed Specification 3.3.1.1 would require a l larger number of reactor coolant loops operating than previously recuired in existing Specifications 3.3.C.2, 3.3.C.3 and 3.3.C.S. The requirement to have additional loops operating is more restrictive than the existing specifications and is consistent with the revised accident analyses for the Haddam Neck Plant. Therefore, the deletion of current Specifications 3.3.C.2, 3.3.C.3 and 3.3.C.5 is acceptable.

Current Specification 3.3.H was a one-time requirement during the 1981 refueling outage to demonstrate natural circulation cooldown capability at the Haddam Neck Plant and for operator training in natural circulation cooldown. All requirements of this specification were adequately demenstrated during the 1981 outage and, therefore, this specification is no longer needed. The staff has reviewed the proposed modification to delete current Specification 3.3.H and concludes that deletion of the existing specification is acceptable.

In addition to the deletion of the specifications dircussed above, the proposed reformatted Specification 3.3 has added new applicability, actien ctatements and surveillance requirements. New limiting conditions of operation have been proposed to assure that the assumptions used in the safety analyses concerning operation of the plant are preserved.

The staff has reviewed the proposed revision to existing Technical Specification 3.3 and has concluded that (1) a comparison of the current specification and the proposed limiting conditions of operation have verified that all previous requirements, with the exception of four items discussed earlier, have been retained in the proposed refomatted specifi-cation, and (2) that the proposed specification is supported by the plant Therefort, design safety analyses perfomed for operation during Cycle 15.

the staff concludes that the proposed revhions to Specification 3.3 are acceptable.

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t B.4 Specification 3.5, "Chemical and Volume Control System" i

5pecification 3.1, "Minimum Water Volume and coron Concentration in the Refuelino Water Storage Tanx" i Specification 3.1, "containt ent"  ;

3pecification J.m3. "Re fueling"  ;

Consistent with the tiiscussion in Section 3.8.1 above, the licensee '

calculated the required shutdown margi.n for Cycle 15 using assumptions which have previously been reviewed and approved by the staff for other Westinghouse designed pressurized water reactors. The shutdown margin requirements determine the minimum refueling boron concentration required beassure to that the Keff values given in Table 1.1 OPERATIONAL N00ES, can satisfied. Modifications to the above specifications have been made to reflect the revised shutdewn margin analyses for Cycle 15. As mentioneo earlier, the revised shutdown margin calculation has been reviewed by the staff and found therefore, to b6 consistent with currently accepted practice and was, acceptable.

The licensee has also proposed modifications to the Bases of the above sections to reflect the revision to the shutdown margin analyses. The licensee has also deleted from the Bases section of Specifications 3.5 3.7the to references Cycle 15 to Cycle 1 rod worth data because it is no longer applicable core. [

The staff has reviewed the proposed changes described above and has concluded that the proposed ecdifications are consistent with the design analyses submitted in support of Cycle 15 operation which have been approved by the staff. Therefore, the staff concludes that the proposed technical specifications are acceptable.

B.5 Specification 3.9. "Operational Safety Instrumentation and Control Systems" Proposed revisions to Specification 3.9 consist of changes to the logic requirements of existing instrumentation identified in Table 3.9-1 to assure that assumptions made in the Cycle 15 safety analyses are preserved.

More specifically, the current specification provided a 1/1 logic require-ment for operation during start-up for the intermediate range start-up rate reactor trip.

The proposed requirement for the start-up rate trip requires that both start-up rate channels'(1/2 logic) be cperable due to single-failure considerations.

The present design of the variable low-pressure reactor trip 1cgic includes

- three channels and requires a total.of two trip signals from the available channels to initiate a reactor trip. Table 3.9-1 which specifies the minimum requirements for operability permits one channel to be inoperable (due to maintenance or surveillance). Should a channel be declared inoper-able, the inoperable channel is placed in the "tripped position and any trip signal (that is halffrom the remaining two channels would result in a reactor trip logic). During this 1987 outage additional channel for variable low-pressure re. CYAPCO actor trip,has installed however thean current logic requirement of needing two separate channel trips to initiate

reactor trip has remained unchanged. CYAPC0 has requested a change to Table 3.9-1 to reflect the change in the minimum functional requirements to keep the existing trip logic unchanged. As mentioneo earlier, should a trip channel be declared inoperable, that channel is placed in the "tripped" trip signalposition from oneandofthe theresulting requirement three operable for a (i.e.,1.3 channels reactor trip )is. any logic Because the existing requirements for reactor trip frcm the variable low-pressure trip channels remains unchanged by the addition of a fourth channel and the proposed revision to Table 3.9-1, the staff concludes that the proposed change is acceptable.

Lastly, the maximum pswer level for operation for the low coolant flow .*eactor trip with 2/4 logic is being revised from 84 percent power to 7a percent power. l This revision prevents the potential for operating above the ellowable power level for three-loop operation if inadvertent three-loop operation occurs after the loss of a single channel of the low coolant flow reactor trip.

The staff has reviewed the proposed revisions to Specification 3.9 and concludes that the proposed revisions are consistent with the design analyses for Cycle 15 and are, there. fore, acceptable.

B.6 Soecification 3.10. "Reactivity Centrol" Proposed Specification 3.10. "Reactivity Control Systems," replaces in-entirety current Specification 3.10. "Reactivity Control"; Specification 3.15, "Reactivity Anomalies"; and Specification 3.16 "Isothereal Coefficient of Reactivity."

As stated earlier, Table 1 contains a ene for one comparison of tne current requirements and the proposed reformatted and renumbered sections in the proposed TS for Cycle 15. Review of this table has assured the staff that all existing specifications, surveillance requirements, and action statements have been preserved in the proposed TS with three exceptions. These are l Specifications 3.10.B, 3.10.C and 3.10.F.3. The bases for their deletion are discussed below.

Current Specifications 3.10.5 and 3.10.C define the maximum worth of any control rod at rated power and at zero power, respectively. The original TS provided a formal fuel cycle design requirement for the early generatien

! of Westinghouse plants. The development of the Westinghouse relead methodology and pro:edures, reduced the need for a TS requirement for ejected rod worth. The reload safety anal includes all of the physics requirements (ysis e.g., checklist reactivity worth, forpower Haddam Neck peaking, etc.) needed to evaluate the impact of the fuel cycle design on the design basis rod ejection accident.

Based upon the above, the staff concludes that the deletion of the two TS identified above is acceptable.

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- 14 Current Specification 3.10.F.3 defines an action required if more than one red position indicator (RPI) and/or its group position indicator are inoperable. The bases for operation in the above configuration (more than one red position unknown) cannot be supported by the existing Cycle 15 safety analyses. Therefore, the licensee has requested that the current specification be deleted. In addition, proposed Specification 3.10.2.2 would restrict operation to a situation where no more than one RPI was inoperable as analyzed in the Cycle 15 design bases analyses. On the bases of the above discussion, the staff concludes that deletion of current j Snecification 3.10.F.3 is acceptable.

)

, In addition t5 the proposed deletions described above, proposed Section 3.10 contains revised requirements for the three-loop control rea insertion limits and four- and three-loop shutcown margin.

} The shutdcwn margin requirements for MODES 1 and 2 for three- and fcur-loop operation are based en the results of the revised steamline break accident submitted in support of Cycle 15 operation. The proposed shutdown margins cf 1800 pcm (four-loop) and 2600 pcm (three-loop) come Cirectly frem the revised analyses and assure that the assumptions used in the safety analys%

will be preserved fer Cycle 15. The requirements for MCCES 3, 4, anc 5

, are established by the boron dilution accident. Both accident analyses have been revised to reflect Cycle 15 parameters and have been reviewed by

the staff and found to be acceptable fer 180 EFFD.

' The current requirements for fcur-loop control red insertion, control rod alignment, RPI and step counter operability, centrol rod drop time, and shutdown bank withdrawal requirements are unchanged, and the current

' moderator temperature coef ficient (Specification 3.16) anc reactivity ancmalies (Specification 3.15) requirements have been trantferred to proposed Specification 3.10.

A new specification has been added to establish the minimum temperature for criticality at a value of 525'F in order to assure that (1) the moderator temperature coefficient is within the analy:ed range, (2) the reactor trip Instrumentation is within its norr.a1 cperating range, (3) the pressuri:er t

is operable with a steam bubble, and (4) that shutdown cargin requirements

) are met. '

Further, the proposed reformatted technical specification contains new applicability stataments, action statements and surveillance requirements.

New limiting conditions of operation have also been proposed to assure that the assumptiens used in the Cycle 15 safaty analysis are preserved.

These new requirerents reflect the current guidance contained in the Westinghouse STS and have been mcdified to reflect the Cycle 15 design safety analyses or the Haddam Neck Plant equipment, as appropriate.

The staff has reviewed proposed Specification 3.10 and concludes that (1) a comparison of the current specifications and the preposed limitirg conditiens of operation have verified that all previous requirements, with the exception of the three items discussed earlier, have been retained in

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the proposed refomatted specification, and (2) that the proposed speci-fication is supported by ,the plant design safety analyses for Cycle 15 operation which has been reviewed and found acceptable by the staff.

Therefore, the staff concludes that the proposed revisions to Specification 3.10 are acceptable.

B.7 Specification 3.17, "pewer Distribution Limits" Proposed Specification 3.17. "Power Distribution Limits," replaces in entirety current Specification 3.17. "Limiting Linear Heat Generation Rate (LNGR)"; Specification 3.18. "Pewer Distribution Fonitoring and Control";

and Specification 3.20 "Reactor Coolant System Flew, Temperature and Pressure." As centioned earlier, preposed Specification 3.17 also includes the specification on quadrant power tilt frem the footnote in current Specif t:ation 2.4 and adds a new specification for nuclear enthalpy rise hot

- channel factors (i.e., peaking factors) for three- and four-loop operation.

As stated earlier, Table 1 contains a one for one comparison of the current requirements and the proposed reformatted and renumbered sections in the proposed technical specifications for Cycle 15. Review of this table has assured the staff that all existing specifications, surveillance require-ments, and action statements have been preserved in the preposed technical specifications with two exceptions. These are current Specification 3.18.B.1.2 which describes the use of the thirble correlation. method for core power distribution monitoring and Specification 3.18.C.2 which required monthly average position of Control Bank B shall be at least 280 steps withdrawn when above 20 percent of rated thermal power when weighted by the daily average thermal energy generation. The bases for deletion of the two specifications are discussed below.

The thimble correlation method of monitoring power distribution has not been used and is justifiably deleted based on the reliability of the axial offset monitoring capability. Deletion of Specification 3.18.B.1.2 is, therefore, acceptable.

The licensee also proposes to d(Tete Specification 3.18 C.2. This speci-ication was intended to provide a means for controlling Xenon induced axial power shapes (and, therefore, peak linear heat genera. tion rate) during nonnal operation and maneuvering prior to the implementation of axial offset monitoring. Since the implementation of axial offset monitoring, the purpose of tracking the Bank B position has been to provide additional assurance that the assumptions in the fuel cycle design axial shape analysis remain valid. As with all Westinghouse plants, steady state operating conditions such as power level, control rod position, and temperature are monitored as part of nonnal core monitoring activities.

Based on the use of approved axial offset monitoring curves, the requirement to monitor Bank B position is no longer necessary and, therefore, the staff concludes that deletion of the current specification is acceptable.

_ _ _ _ _ - - - - - - - - - - - . J

In addition to the deletion of the specifications described above, proposed Specification 3.17 will include revised axial offset limits based on the Cycle 15 design safety analyses. The revised limits will consist of two

( curves, one valid for the range of 0-250 effective full power days (EFPD) and the other valid from 250 EFPD to the end of Cycle 15. The proposed specification also includes revisions to the RCS flowrate requirements for three and four-loop operation as well as a revision to the four-loop inlet tesiperature. The core flow rates of 233,870 and 184,730 gpm for four and three loop operation, respectively, were used in the design basis reanalysis and safety limit curves. These core flow rates correspond to vessel flow rates of 246,000 and 194,000 gpm, respectively, based on a conservative and previously approved core bypass flow fraction of 4.5%. The Cycle 14 measured vessel flow rates for four and three loop operation (including uncertainties) were 259.000 and 203.000 gpe, respectively. These data indicate the operating margin to the values used in the safety analysis.

New specifications for radial peaking for four- and three-loop operation have been proposed and are supported by the fuel cycle design and safety analysis for Cycle 15. The current TS total peaking value of 1.78 corre-spondstoanFfHvalueof 1.656 (1.78 divided by the engineering factor of 1.075). The proposed full-power values of 1.60 (four-loop operation) and 1.64 (three-lcep operation) are more restrictive and are bounded by the original design basis value (1.656). The proposed limits include an allowance for a peaking increase with reduced power (increased red insert-icn).

In addition, the proposed reformatted TS has' added new applicability, action statements and surveillance requirements. New limiting conditions of operation have been prr posed to assure that the assumptions used in the Cycle 15 safety analysis are preserved. These new recuirements reflect the current guidance contained in the 'destinghouse STS and have been modified to reflect the Cycle 15 design safety analyses or the Haddam Neck Plant equipment, as appropriate.

The staff has reviewed the proposed Specification 3.1F and concludes that (1) a comparison of the current specifications and the preposed limiting conditions of operation have verified that all previous requirements, with the exception of the two items discussed earlier, have been retained in the proposed refomatted specification, and (2) that the proposed specification is supported by the plant design safety analyses for Cycle 15 operation which has been reviewed and found acceptable by the staff. Therefore, the staff concludes that the proposed revisions to Specification 3.17 are acceptable.

B.8 Specificatien 3.24, "Seecial Test Exceotions" Specification 3.24 is a new technical specification that for:ralizes excep-tions to ree.yirements in order to perform various low-power start-up physics tests.

The new specification includes formal test exceptions currently identified in Specificatiens 3.10, A, 3.10.0 and 3.18.C for shutdown margin, moderater temperature coefficient, minimum temperature for criticality, control rod m . - - - _ - - - - l

alignment, control rod insertion (shutdown and control banks), and control rod position indication. The staff has reviewed the new specification and concludes that the proposed specificatien is consistent with the current guidance contained in the Westinghouse STS and has been modified to reflect the Cycle 15 design tafety analyses or the Haddam Neck Plant ecuipment, as appropriata. The staff concludes that the proposed specification is acceptable.

B.9 Specification 4.9. "Main Steam Is'olatien Valves" The current specification requires that a closure time of 10 seconds for the MS!Ys be verified each cold shutdcwn if it has not been tested in the previous three months. The proposed scecificatien clarifies the closure time requirement by specifying simultanecus closure of all four valves within 10 seconds, which is censistent witn tne assumptions made in the design bases analysis.

The staff has reviewed the prcposed change and concludes that it is acceptable.

C. Surmary The staff has reviewed the licensee's submittal in support of Cycle 15 cperation and the preposed TS changes for the Haddam Neck Plant. The staff concludes that the nuclear design of the Haddam Neck Plant for Cycle 15 is acceptable.

The staff also reviewed the reanalysis of the six design basis accidents which sre affected by the Cycle 15 reload and the proposed TS changes for consistency, adequacy arid completeness. Our review found that the reanalyses of thase six accidents was performed conservatively and in a manner consistent with the proposed TS. The set of accident analyses submitted by the licensee bounds other accidents or transients which would be impactec by changes described above, and, therefore, constitutes a ecmplete set of all accidents which are required to be reanalyzed. We also found that the required changes to the TS imposed by the new physics parameters of the reload core are based upon the results of the ac,gident analyses which were perfomed in a conservative manner.

The licensee stated that operation through 180 EFPD of Cycle 15 is beunded by the results and assumptions of the current design basis steamTherefore, line break analysis submitted in 1980 and previously approved by the NRC.

approval of the revised steam line break analysis is notHowever, required operatienfor Cycle 15 startup or the period between startup and ISO EFPD.

beyond that point requires an approved steam lir.e break analysis.

4.0 ENVIRONMENTAL CONSIDERATION

This amendment involves a change to a requirement with respect to the installation or use of facility ccmponents located within the restricted area as defined in 10 CFR Part 20 and changes to the surveillance requirements.

The staff has detamined that the amendment involves no significant increase

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in the amounts, and no significant change in the types, of any effluents that may be released offsite and that there is no significant increase in individual or cumulative occupational radiation exposure. The Cemission has previsculy issued a proposed finding that this a endment involves no significant ha:ards censideration and there has been no public cement on such finding. Accordingly, this amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environcental impact statement or environmental assessment need be prepared in connection with the issuance of this amendment.

5.0 CONCLUSION

The staff has concluded, based on the considerations (iscussed above, that (1) there is reasenable assurance that the health anc safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in ecmpliance with the Ccmission's regulatiens and the issuance of this amendment will not be inimical to the ecmon defense and security or to the health and safety of the public.

6.0 ACX.NCAEDGEMENT This Safety Evaluation has been prepared by F. Akstulewicz of POISA, NRR in coordination with C. Liang, G. Hsii and L. Lois, all of XRSB, hKR and XRSB technical centractor - International Technical Services, Inc.

7.0 REFERENCES

1. "NUSCO Thermal Hydraulic Medel Qualification Volume ! (RETRAN)," NUSCO 140-1, July 30, 1984

2. "Haddam Neck Plant Non-LOCA Transient Analysis," NUSCO 151,

3. "Haddam Neck Plant Cycle 15 Reload Technical Specification Change Requests and Reload Report," letter frem E. J. Mrce:ka (NUSCO) to U.S.

Nuclear Regulatory Cemission, June 1,1987.

4 "Haddam Neck Plant Revisions to Reanalysis of Nqn-LOCA Design Basis Accidents," letter from E. J. Mroczka (NUSCO) to U. S. Nuclear Regulatory Comission, May 8,1987.

5. "Haddam Neck Plant Additional Infomation - Rennalysis of Non-l.0CA Design Basis Accidents," letter frem E. J. Mroczka (NUSCO) to U.S.

Nuclear Regulatory Cemission, September 2,1987.

6. "Safety Evaluation Report on the RETRAN Ccmputer Code," U. S. Nuclear Regulatory Cemission, July 1984

7. "Safety Evaluation by the Office of Nuclear Reactor Regulation Regarding NUSCO Topical Report 140-2 VIPRE-01 Connecticut Yankee Atomic Pcwer Ccepany Docket No. 50-213 Haddam Neck Plant," October 1986.

1

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8. "Safety Evaluation by the Office of Nuclear Reactor Regulation  :

4 Regarding NUSCO Topical 140-2 VIPRE-01 W-3L CNBR Limit," U. S. Nuclear i

Regulatory Ccmmission August 1987.

! 9. "Physics Methodology for PWP. Reload Design," NLSCO-152, August 30, 1986.

10. "TAC 02, Fuel Pin Perforr.ance Analysis," BAW-10141 PA, Y. H. Hsii ..

et al., S&W Lynchburg, Virginia, dated June 1983.

] -

i; 11. "Chapter 10 - Incidents and Potential Hazards," Final Design Safety i

Analysis, Cennecticut Yankee Atemic Power Company, May 1966, i 12. S'n ' 0155, "Connecticut Yankee Atomic Power Company Haddam Neck l

' Nst c' Technical Report Supporting Cycle 15 Operation, dated 1 v+ 387. ,

13. Letter from J. H. Taylor, Babcock and Wilcox, to J. S. Berggren, NRC, "B&W's Response to TACO 2 Questions," dated April 8, 1982.

i l

, i Dated: November 12, 1987 .

j Corrected: April 28,1988 1

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TABLE 1 Comparison of Current Haddam Neck Technical Specifications to the Proposed Reformatted Standard Technical Specification Sections for Cycle 15 Operation at the Haddam Neck Plant Current Technical Seecification Proposed Cycle 15 Technical Specification 3.3 - Section -

3.3.A 3.3.2.1 3.3.2.2 3.3.B 3.3.1.2 Action Statement C 3 . 3 .~ 1. 3 Action Statement B 3.3.1.4.1 Action Statement B 3.3.1.4.2 3.3.C (1) 3.3.2.2 3.3.2 BASES 3.3.C (2) '(deleted) 3.3.C (3) (deleted) 3.3.C (4) 3.3.1.1 Action Stat: ment A 3.3.C (5) (daleted) 3.3.C(6) 3.3.4.1 3.3.C (6)(a) 3.3.4.1 Action Statw.ent A 3.3.4.1 Action 3tatement B 3.3.C (6)(b) 3.3.4.1 Action Statement A .

3.3.4.1 Action Statement B 3.3.C (7) 3.3.3 Action Statement A 3.3.3 Action Statement B 3.3.C(7)(a) 3.3.3 Action Statement B 3.3.C(7)(b) 3.3.3 3.3.0 3.3.1.3 3.3.1.4.1

3.3.E 3.3.4.2 3.3.E (1) 3.3.4.2 Action Statement A i 3.3.E (2) 3.3.4.2 Action Statement B i 3.3.E (3) 3.3.4.1 Action Statement C -

l 3.3.F(1) 3.3.1.2 3.3.7 (2) 3.3.1.2 3.3.F 2)(a) 3.3.1.2 Action Statement A 3.3.F 2)(b) 3.3.1.2 Action Statement C l

3.3.6 1) 3.3.1.3 3.3.G 2) 3.3.1.4.1 l 3.3.1.4.2 3.3.G (2)(a) 3.3.1.3 Action Statement A 3.3.1.4.1 Action Statement A 3.3.1.4.2 Action Statement A 3.3.G(2)(b) 3.3.1.3 Action Statement B 3.3.1.4.1 Action Statement B 3.3.1.4.2 Action Statement B 3.3.H (deleted)

s s esv TABLE- 1 CONTINUE 0 Current Technical Soecification Precosed Cycle 15 Technical Soecification 3.3 - Section Section 3.10 3.10 A 3.10.2.6 3.10.2.7 3.24.1

'3.10.B (deleted) 3.10.C (celeted) 3.10.0(1) 3.10.1.2 3.10.1.3 3.10.1.2 Action Statement 3.10.1.3 Action Statement 3.10.0(ii) 3.10.2.6 3.10.2.7 3.10.0 (11)(1) 3.10.2.6 Action Statement A 3.10.2.7 Action Statement A 3.10.0 (11)(2) 3.10.2.6 Action Statement B

' 3.10.2.7 Action Statement B 3.10.0 (11)(3) 3.10.2.6 Action Statement C 3.10.2.7 Action Statement C 3.10.E 3.10.2.1 3.10.2.1 Action Statement 3.10.F 3.10.2.2 3.10.2.2 Action Statement 3.10.G 3.10.2.3 3.10.H 3.10.2.4 3.10.2.4 Action Statement 3.10.2.4 Surveillance Req.

3.10.I 3.10.2.5 3.10.2.5 Action Statement 3.10.2.5 Surveillance Req.

Section 3.15 3.10.1.1 -

3.10.1.4 3.10.1.1 Surveillance Req.

3.10.1.4 Surveillance Req.

Section 3.16 3.10.1.5 Section 3.17 3.17.A 3.17.2.1 3.17.C 3.17.2.1 3.17.2.1 Surve111 ante Req.

3.17.2.2 3.17.2.2 Surveillance Raq.

3.17.0 3.17.2.2

_