ML20235B591

Undated Technical Evaluation:Haddam Neck Plant Reanalysis of Non-LOCA DBAs & Tech Spec Change Requests in Support of Reload Cycle 15

ML20235B591
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Site:	Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date:	09/24/1987
From:	INTERNATIONAL TECHNICAL SERVICES, INC.
To:
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NUDOCS 8709240157
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Technical Evaluation:

Haddam NeckLPlant Reanalysis of Non-LOCA' Design Basis Accidents and Technical Specification Change Requests in Support of Reload Cycle 15

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, International Technical Services, Inc.  !

420 Lexington Avenue.

New York, New York 10170  ;

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Evaluation o_t Haddam Neck Plant Reanalysis of Non-LOCA Desian Basis Accidents and Technical Specification Chanae Reauests in Succort of Reload Cycle _)5 1.0 Summary '

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( As part of the Cycle 15 reload for the Haddam Neck Plant, NUSCo plans to

! add 4 additional fuel assemblies to the core. As a result, several physics parameters will be changed. These parameters and their respective changes include (i) reduced shutdown margin, (ii) increased differential rod worth, i

(iii) revised reactivity (Doppler and Moderator Temperature) coefficients, and l (iv) revised axial peaking.

.iince certain of the current Technical Specifications (Tech Spec) are not met by the new reload, NUSCo requested approval of certain Tech Spec changes resulting from the reload. In addition, NUSCo took this opportunity (i) to conform to the Westinghouse Standard Format, (ii) to incorporate or add functions which would make the procedures ca:,ier for the operator, (iii) to conform to the Standard Review Plan requirements, and (IV) to reflect the more current plant conditions. All but five of the proposed Tech Specs resulted in no change or impose more restrictive requirements. The other five have been found to be acceptable in this review, provided that the Steam line Break accident reanalysis to be submitted in support of operation beyond mid-cycle

( supports the shutdown margin results.

The Cycle 15 reload and the Tech Spec changes required that four affected 1 l l

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accident analyses be revised. These analyses are the uncontrolled rod withdrawal, steam line break, boron dilution and rod cluster control assembly .

(RCCA) ejection. In addition, the dropped rod and loss of flow analyses were resubmitted because (1) NUSCo. plans to disable the turbine load runback feature and the submitted reanalysis doe not'take credit for this feature,-and-(2) NUSCo expects some amount of steam generator tube plugging.therefore needs to reanalyze LOF at lower flow conditions. The revised analyses were submitted to the NRC for approval.

Boron dilution governs the beginning of. cycle shutdown margins. 'NUSCo stated that analysis supports the conclusion that' operation through mid-cycle of Cycle 15 is bounded by the results and assumptions of.the current design.

basis steam line break analysis submitted in 1980 and previously approved by the NRC. Therefore, approval of. the revised steam line break analysis. is not required for' Cycle 15 start-up [4,5] and the period between the start-up and-mid-cycle. However, operation beyond that point requires acceptable SLB analysis. We note that the steam line break accident analysis governs three of the Tech Specs concerning shutdown margins; -Tech Spec Nos 3 .10 .1 '.1, 3.10.1.2 and 3.10.1.4. Review of the SLB reanalysis and affecting Tech Specs l

will be performed promptly following receipt of the NUSCo reanalysis promised to be submitted at a later date.

I ITS reviewed the reanalysis of the five design basis accidents and the l

proposed Tech Spec changes (see Attachment _1) for consistency, adequacy and completeness. Our review found that the reanalyses of these five transients l

was performed conservatively and in a manner consistent with the proposed Tech Specs. The set of transient analyses submitted by 'the applicant bounded the L other transients which would be impacted by changes described above, and therefore constitutes a complete set of all transients which.are required to be 're-analyzed. We also found that the required changes to' the Tech Specs imposed by the new physics parameters of the reload core are based upon the' results of analyses which_ were thoughtfully performed' in a conservative manner. With the exception of three Tech Specs mentioned earlier, these changes are properly reflected in the proposed Tech Specs and, if implemented, would provide adequate protection during these reanalyzed accidents.

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2.0 Introduction {

The Haddam Neck Plant is currently shut down for Cycle 15 reload which will add 56 fresh fuel assemblies to the core instead of the usual 52  !

As a result, several physics parameters are more limiting than l assemblies.  ;

Cycle 14 and the assumptions used in NUSCo 140-1 [1] and 151 [2]. These parameters include (i) reduced shutdown margin, (ii) increased differential rod worth, (iii) revised reactivity (Doppler and Moderator Temperature) coefficients, and (iv) revised axial peaking. This required that four affected accident analyses be revised: these analyses are; the uncontrolled ,

rod withdrawal, steam line break, boron dilution and rod cluster control assembly (RCCA) ejection.

In addition to the changes in the analysis assumptions due to reload, it became necessary to change the Technical Specifications because of the changes in physics parameters and an anticipated necnssity for a reduced minimum RCS fl ow. The applicant also took this opportunity to make additional Technical Specification changes: (i) to conform to the Westinghouse Standard Format, (ii) to incorporate or add functions which would make the procedures easier for the operator, (iii) to conform to the Standard Review Plan requirements, (iv) to reflect the more current plant conditions. NUSCo therefore concurrently requested approval of these Tech Spec changes [3]. Our assessment of the Tech Spec changes is found in Attachment 1 to this report.

In addition, reanalysis of the Loss of Flow accident was necessitated by the Tech Spec change to minimum RCS flows, and the Dropped Rod analysis was submitted for approval since the applicant wishes to disable the turbine runback feature to prevent spurious dropped rod signals from causing unnecessary plant transients.

l The revised analyses were submitted to the NRC for approval [4]. NUSCo stated that analysis supports the conclusion that operation through mid-cycle of Cycle 15 is bounded by the results and assumptions of the current design l l

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basis steam line break analysis submitted 'in 1980 and previously approv'ed by the NRC. Therefore, approval of the revised steam line break analysis is not required for Cycle 15 start-up [4,5] and the period between start-up 'and mid-cycle. However, . operation beyond that point requires acceptable stb analysis.

For the reload ' analyses, RETRAN, VIPRE and a NUSCO ' adaptation. of-I Westinghouse physics methodologies.. were employed. . RETRAN02/ Modo 2 has been approved .by the NRC, . with certain conditions [6]. The VIPRE-01. methodology.

has.been previously approved by the NRC [7,8]. _The physics methodology, which' has its base in the Westinghouse methodology, was used to compute physics parameters for the Cycle 15 reload core. This methodology has also been approved by the NRC [9,10].

Although RETRAN02/M003 has not yet been approved by the NRC, it is .

reportedly a corrected version of ' RETRAN02/M0D02 which was reviewed by the NRC. We find that analysis with RETRAN02/M0003 is acceptable as used by NUSCo for these transients, unless future review of that code - determines to 'the contrary.

3.0 Reload Analyses 3.1 Bod Withdrawal Transient Due to the design of Cycle 15 reload, the differential rod worth has increased substantially beyond the current design basis assumption.[9]. -

Therefore, Tech Specs are proposed to require the startup' rate trip to be operable whenever the reactor trip breakers are closed and the control . rod drive lift coils are energized, up to the P-7 interlock. -In addition, Tech Specs are proposed to require a different (more-restrictive) number of '

operating RCS loops during subcritical conditions.

Uncontrolled rod withdrawal transient analyses were performed by - -

NUSCo for 4-loop operation commencing at 100 % power and' 65 % power and commencing at 65 % power for 3-loop operation and at subtritical with and' 4

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without rod stop functioning. In all subcritical transieht cases, NUSCo i reasonably concluded that such transient would be terminated by the start-up  ;

rate trip or, by operator action before a- significant power level was reached and therefute that fuel thermal limits wouid not be challenged nor would DNBR limits be reached. ,

i Parametric studies were made for those transient cases started from power, 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 use: (1)leastnegativeDopp]erandmost positive MTC vs most negative Doppler and MTC; (ii) highest RCCA stuck out; and(iii),reactivityinsertionratesupto22.5pcm/sec(whichdasthemaximum .

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obtainable from any single or combination of two banks). Conservatism 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 DNBR computations were made.

No parametric computations were done to verify the accuracy of the RETRAN nodalization for this analysis. Nevertheless, since (1) this' is a particularly short transient in which only the primary pressure, core flow, )

and inlet temperature and power are important, and since (ii) comparison by l NUSCo of the pressure and temperature computed by RETRAN to actual plarit pressure and temperature data for a 30 % load rejection and a partial loss of feedwater event was good, and since (iii) as' stated in Section 3.5, we find the core flow computation acceptable for short transient such as this, we have reasonable assurances that NUSCo's computation of core fl ow., core inletI temperature and primary pressure are accurately predicted by the RETRAN model.

Furthermore, we have reasonable assurances that core power was conservatively computed in this analysis. Thus we find that we have reasonabit assbrances that the input parameters to VIPRE-01 which were derived from RETRAM are i adequate for the' purpose of the 4-RCP loss of flow analysis.

Since VIPRE-01 has been previously reviewed and found acceptable by-

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l the NRC, we have adequate assurances that the computation of DNBR by NUSCo for ']

the rod withdraval transient is acceptably accurate. In addition, NUSCo has :i

. imposed sufficient degree of conservatism to_ the; appropriate parameters such l that we have adequate assurances of conservative results.

NUSCo concluded that the minimum DNBR was always greater than ;1.3 and therefore that. fuel thermal limits would not be exceeded. Based up'on the ]

foregoing, we have reasonable assurances that these conclusion are correct.

3.2 Boron Dilution This '. reanalysis was submitted because the shutdown' margin is being reduced due to reload physics parameters.

c NUSco has stated thei the maximum possible dilution rate is 180 gpm and has therefore performed a simple computation of the reactivity insertion rate for such . dilution rate and ' reached the reasonable conclusion that'such rate is "well within the reactivity insertion rates of the uncontrolled RCCA withdrawal analysis." On that~ basis, no RETRAN and no VIPRE-01 DNBR g computations were. explicitly performed for the boron dilution accident. We concur with NUSCo's conclusion and have adequate assurances that the minimum DNBR will not be challenged by the boron dilution accident.

The boron dilution analyses.was used to established shutdown margins which would enable the operator to have at-least 15 minutes for the time from the first safety alarm 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 SRP 15.4.6. These new shutdown.. margins were them reflected . in new technical specifications and used as input for the other. 1 transients discussed in this review, all of which yielded results with respect to'which we have reasonable assurances are conservative. However, according.

to NUSCo, the Steam Line Break accident becomes the controlling accident for shutdown margin after approximately mid-cycle; and. therefore, we cannot' state, until we have reviewed the promised reanalysis of the Steam Line Break 6

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accident, that we- have: reasonable assurances that. the proposed shutdown-1 margins are conservative after approximately mid-cycle.

3.3; RCCA E.iection The' RCCA Ejection represents ~ the most. rapid potential reactivity insertion accident. . The NUSCO analytical. methodology for this accident is similar to that described above for the uncontrolled. rod withdrawal accidents, with NUSCO stating that they employed the following. conservatism: (i) the limiting; burnup parameters were combined to ' generateL the most severe system response, -and (ii) point kinetics was used with :no Doppler weighting.

multiplier. More specifically, burnup. parameters used were: no credit for ,

flux flattening effects of reacavity feedback; maximum bank-insertion at such I power level; adverse xenon; margins added to ejected rod worth to account for calculational uncertainties; fuel temperature feedback assumed to -be at its' minimum value over the entire burnup range; use'of the most positive MTC; use of the smallest delayed neutron fraction over. the entire burnup range (to minimize time to prompt criticality). In addition, trip reactivity was computed without considering the ejected rod and using trip and trip response delays .

NUSCo performed analyses for 4-loop HFP and HZP. and 3-loop full power and HZP operations, i

NUSCo computed that, assuming failure of all rods which reached DNBR of less than 1.3, 18% of the fuel rods' failed and.no rod had fuel melting at I the centerline in the 4-loop HZP case and none in the HFP cases. We did not

review the aspect of the submittal which describes the radiological consequences. We suggest that the NRC staff should review this section.

I 3.4 Drocoed Rod Accident j

1 The NUSCo analysis for this event used the same methodology as j s

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described for the RCCA ejection and' rod withdrawal accidents' described above. ]

This transient is a power reduction transient for which NUSCo used to have a {

turbine runback feature. However, because of a' large number of plant transients caused by spurious dropped rod signals, NUSCo. had - decided to

' disable the turbine runback feature and therefore needed to re-analyze this

transient. Parametric analyses were' conducted varying the. dropped rod worth from 0 to 180 pcm- (minimum to maximum exp'ected values) and with turbine load runback in manual and in automatic. The RCS was, in contrast to the reactivity insertion transient discussed above, assumed to be initially in j conditions with maximum core, inlet temperature, but with' minimum. primary.

pressure and core- flow, in each case' intended to produce the minimum computed t 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 l no trip in that case) to maximize the. power thus also tending to produce a- '

lower DNBR.

NUSCo concluded that the minimum DNBR was obtained for the full power 4-loop operation without taking credit-for.either turbine runback or rod stop protective features. In that case, NUSCo 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 rod accident for Haddam Neck is conservative.

3.5 Loss of Flow .i The applicant felt that the current Tech Spec limit for low RCS flow-  !

rate for 4-loop operation may not be met since there is very little available l margin now and some degree of steam generator tube. plugging is anticipated. I In order . to justify the lower proposed RCS . flow rate tech spec, NUSCo submitted the LOF analysis assuming a lower flow rate. In addition, the new i analysis included a change of the 3-loop low flow trip setpoint in the.

governing tech spec when going from 4-loop operation to three operation.

1 In 93.20 of the Tech Specs, NUSCo stated that their flow rate. j 1

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. v requirements were based upon a steam generator plugging level consistent with 500 equivalent plugged tubes per steam generator, and that an evaluation had been performed with the core physics by-pass flow fraction reduced to 4.5 % )

from 9 % which indicated that although the reactor vessel flow rate decreased l due to the plugging, the core flow increased due to the reduced bypass flow. l NVSCo had further advised usl that the net effect of these is that core flow I remains the same when the minimum RCS flow rate Tech Spec is reduced to f l

246,000 gpm (4-loop) and 194,000 gpm (3-loop) and that the core flow rate used l in the VIPRE calculations was the same as that used in the RETRAN calculations.

NUSCo analyzed the worst case loss of flow (LOF) event (complete loss of flow from full power), which results in the most severe power-to-flow i ratio and therefore the lowest DNB. In Section II.E of NUSCo 140-1, NUSCo presented LOF sensitivity studies of the impact of variation of RCP inertia, ,

junction inertia, rod insertion time, delay of scram signal and reduction of ramet s had a m nimal impact upon DNBR calculated using the correlation, and therefore that normal values could be used. Core flow was I compared to the FDSA computation for a 4 RCP coastdown in sIV.C of NUSCo 140-1, in which excellent agreement was obtained, thus lending credibility to l 1

NUSGo's choice of those parameters and to the overall core flows computed by i the RETRAN model for this transient. Although no experimental data were I presented to verify the flow coastdown curves, for the purpose of this I transient, we believe that good comparison to FDSA (11] (which we believe was l based upon manufacturers specifications) gives adequate assurances of accuracy l of the RETRAN model to permit its use in this event, l

NUSCo conservatively assumed that the transient commences from 102 % I power for the 4-loop transient and 67 % for the 3-loop transient and with the RCP inertia reduced by 10 % to cause the flow to reduce more rapidly. In addition, NUSCo assumed minimum initial RCS pressure and flow (although they did not re-analyze with the new lower initial tech spec RCS flow because core flow remained the same with the lower flow as with the higher one because of reduced bypass flow), and maximum core inlet temperature. Each of these 4

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i assumptions is conservative,-tending to lessen the computed minimum DNBR. In addition, the computed RCS pressure rise was: minimized by assuming minimum initial pressurizer level, maximum initial SG -level and maximum turbine stop valve closure time, pressurizer heaters off and maximum pressurizer. spray flow, and charging isolated with letdown flow available, Finally, reactivity insertion was maximized through the use of the .least negative Doppler coefficient and the most' positive MTC and reactor trip computation'. included instrument response and delay times. All of these conservatism contributed f to the acceptability of the analysis.

Also as in the Rod Withdrawal transient- analyses, this transient ,

depends only upon the accuracy of the RETRAN computation of core flow inlet-  !

temperature and flow and primary pressure and the accuracy of the VIPRE-01 ,

computation using them. As stated in Section 3.1, we have adequate assurance of the RETRAN computation of those parameters for short transient of this i type. NUSCO concluded that the minimum DNBR was -approximately 1.4 in' the worst case LOF transient.

Since the VIPRE-01 code has been approved by NRC, and since the  !

minimum DNBRs computed with VIPRE-01 using the RETRAN output described above are above 1.3 for both analyses, we have adequate assurances that the required DNBR limits will be met. _-

1 4.0 ' Conclusions l Five transient analyses have been re-analyzed and found to be conservative. The assumptions and conditions used in the analyses are found to be consistent with the proposed Tech Specs. The Steam Line Break accident is to- be . reanalyzed by NUSCo and will impact the acceptability of Tech Spec No.s 3.10.1.1, 3.10.1.2 and 3.10.1.4. In addition, use of RETRAN models in

.l these analyses is acceptable, i

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5.0 REFERENCES

1. "NUSCo Thermal Hydraulic Model Qualification Volume I (RETRAN)",

NUSCo 140-1, July 30, 1984.

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

3.  :"Haddam tieck Plant Cycle 15 Reload, Technical Specification Change Requests and Reload Report", letter from E.J. Mroczka (NUSCo) to U.S. Nuclear Regulatory Commission, June 1,1987..

4. "Haddam Neck Plant Revisions to-Reanalysis of Non-LOCA Design Basis Accidents, letter from E.J. Mroczka- (NUSCo) :to U.S. Nuclear '

Regulatory Commission, May 8, 1987.

5. "Haddam Neck Plant Additional Information - Reanalysis of' Non-LOCA Design Basis Accidents," Letter from E.J. Mroczka - (NUSCo) . to U.S.

Nuclear Regualtory Commission, September 2,1987.

6. " Safety. Evaluation Report on the RETRAN Computer Code", U.S. Nuclear -

Regulatory Commission, July 1984.

7. " Safety Evaluation by the Office of Nuclear Reactor Regulation Regarding NUSco Topical Report 140-2 VIPRE-Ol' Connecticut Yankee Atomic Power Company Docket No.50-213 Haddam Neck Plant", '0ctober -

1986.

8. " Safety Evaluation by the Office of Nuclear Reactor Regulation Regarding NUSCo Topical 140-2 VIPRE-01 W-3L DNBR .' Limit", U.S.

Nuclear Regulatory Commission, August 1987.

9. " Physics Methodology for PWR Reload Design", NUSCo 152, August 30, 1986.

10. " Safety Evaluation by the Office of Nuclear Reactor Regulation Relating to the Nuclear Design for the Cycle 15 Reload Connecticut Yankee Atomic Power Company Haddam Neck Nuclear Power-Station Docket-No. 50-213", July 1987

11. " Chapter 10 - Incidents and Potential Hazards", Final Design Safety Analysis, Connecticut Yankee _ Atomic Power Company, May 1966-11

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ATTACHMENT 1 1

Review of Technical Specification Haddam Neck Plant l

The following Tables 1 and 2 summarize our assessment of the proposed  !

Technical Specification Changes for Haddam Neck Plant, some resulting from the  !

Cycle 15 reload' and others to update the current Tech Specs to Westinghouse Standard Format or to conform to the Standard Review Plan. Some of the changes are to provide the plant operators with new procedures.

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l TABLE 1 Assessment of the Proposed Tech Spec Changes  !

Tech Spec No. Comment / Assessment 1.13 updated definition to be consistent with current SRP termin01ogy Table 1.1 Refueling Keff changed. The new physics caused a change j in the minimum Keff required to obtain a satisfactory j result from the Boron dilution event to a Keff .9479.  !

The current Tech Spec of .94 satisfies this requirement.

2.2.1 Vse of VIPRE and reduced RCS flow caused a need to gencrate new safety limit curves. The VIPRE-01 I methodologies were reviewed by the NRC and found to be acceptable for licensing calculations. (SER on NUSCO 140-2 dated 10/16/86). In the NRC's review contained in such 12 )

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l i SER, the- NRC' specifically reviewed 'the . new core ' safety 'l limit curves generated with VIPRE-Ol' and; found "very good agreement with the present safety limit curves". Since NUSCo has informed us that core flow rates remained the same between the current Tech Spec and the proposed, all parameters (except core physics'~which caused a more' restrictive peak) remained the same and only approved changes in basic methodology to the use of. VIPRE-01 was made. On that basis we conclude that the new safety limit curves are acceptable for licensing analysis.

2.4 Item (5) The 3 RCP operation low coolant. flow setpoint was reduced from its prior value of 90% of nominal, to 84% of nominal.

to enable NVSCO to avoid re-calibration. of the trip setpoint when running in the .3 pump ~ mode. This .is not related to the reload, but reqJired- reanalysis of'the LOF transient to assure that this setpoint adequately =

prevented DNB. The LOF analysis found that this was the case. Therefore we conclude that this proposed Tech Spec change is acceptable.

Item (8) A new trip is now credited in . analyses on high start-up rate at 5 decades / min. This is required to assure adequate limits .in the rod withdrawal . transients, and is therefore an acceptable change.

3.3.1.1 - 1.3 Changes.in format of Tech Spec.

The Tech Spec limits gives in these sections are identical to those in the current Tech Specs. The Tech Spec.for the number of loops operating in a given mode is determined by whether or not the rod withdrawal analyses:give acceptable results for that configurations. The Tech Specs in this section are consistent with the submitted re-analyses.

i 3.3.1.4.1 This represents a change in Tech Spec not required to the l reload. Whereas the current Tech Spec required 2, coolant  !

loops available . (" coolant loop" is defined as RHR or RC )

loop) unless the reactor. coolant loops are operable, the current Tech Spec permits operation in Mode 5 with (a) either the reactor trip system breakers shall be open or the CRD lift coils shall be de-energized and (b) either 2 RHR pumps operable or one RHR pump operable plus' 2 un-isolated SGs having narrow range water level above 25% i narrow range. This is a more restrictive requirement and  !

is therefore an acceptable change.

3.3.1.4.2 This also represents a more restrictive change, with the current Tech Spec as desired in 1.4.1 above while .the current requires 2 RHR loops operable, one in actual operation and either reactor trip system breakers open or CRD lift coils de-energized. This is an acceptable Tech' 13

Spec change.

3.3.2.1 2.2 Current Tech Spec required that all code safety valves are to be operable. if in Modes 1, 2, or 3. The proposed Tech Spec requires: i a) at least one operable in Mode 4, and b) all operable in Modes 1, 2, and 3.

These proposed Tech Specs are functionally same as current.

3.3.3 Unchanged.

3.3.4.1 Both the proposed Tech Spec and the current Tech Spec require both PORV's operable in Modes 1, 2, or 3.  ;

However, the proposed Tech Spec has specific setpoints j while the current Tech Spec references the FDSA for i values which were the same. Thus this does not represent any change.

3.3.4.2 New.

3.5 This Tech Spec has been cleaned up to reference the appropriate Tech Specs instead of incorporating them j explicitly.  !

l 3.7 This Tech Spec has been modified to reduce the boron concentration requirement from 2247 ppm to 2200' ppm.

Supporting analysis was performed to demonstrate the acceptability of this change.

Table 3.9-1 (2) Pressurizer variable low pressure reactor trip now  !

requires one of three redundant' signals, used to require one out of two. This is a more restrictive requirement and is therefore acceptable, j (5) low coolant flow reactor trip 4 loop operation - now l requires maintaining load below 74 % full power, used to require below 84% full power. This is a'more restrictive condition and is therefore acceptable. ,

SUR Intermediate Range SUR trip now requires 1/2 used to be l 1/1. Back-up was added because NUSCo is now crediting SUR  ;

trip in rod withdrawal analysis. This is a necessary Tech i Spec change.

3.10.1.1 This proposed Tech Spec provides a requirement-that there be a shutdown margin of at least 1800 pcm in Modes 1 and 2 in 4 loop operation. This margin requirement originates with the change in physics parameters caused by the reload, and impacts analyses of the SLB accident and the Baron dilution accident. The margin was adjusted to 14

obtain satisfactory results for. the Boron dilution event . N" and those: margins should ' yield adequate results ? forf the SLB analysis which is the limiting' transient. Thus this- .;

is an acceptable Tech Spec. change;.provided,- however, that .

operation beyond' mid-cycle is not , acceptable unless the i reanalysis of the SLB. accident supports it.

3.10.1.2 Shutdowr. margin requirement in ' Mode :3 with 4 " loop operation is_ also a proposed Tech Spec based upon 'the changes in physics with'the numerical value defined by-the 'l Boron dilution. event analyses. .=SLB- accident analyses

. should find that . the limits obtained by . Boron dilution event analyses yield acceptable SLB results. Thus.this is an acceptable Tech Spec change; .provided, however, that operation beyond mid-cycle- is' not acceptable unlessL the-reanalysis of the SLB accident- supports; it.

3.10.1.3' Shutdown margin for Modes 4 and 5 with 4. loop operation is also a- proposed Tech. Spec based upon the changes in  ;

physics with the numerical value defined by the Boron dilution event analyses.

3.10.1.4 Same as 3.10.1.1, but for 3 loop operation.

3.10.1.5 Same as current Tech Spec 3.16.

3.10.1.6 This_is a new Tech Spec which requires that'Tave in Modes 1 and 2 be greater than or equal .to 525 ~ F, and has no -

impact on the transients analyzed. Thus; this proposed Tech Spec was not reviewed by'ITS. l 3.10.2.1 This' Tech Spec is equivalent to current ~ Tech Spec 3.10.E.

3.10.2.2 This Tech Spec is equivalent to current Tech. Spec 3.10.F.

3.10.2.3 This Tech Spec is equivalent to current Tech Spec 3.10.G.

3.10.2.4 This Tech Spec is equivalent to current Tech Spec 3.10.H.

3.10.2.5 Current Tech Spec [3.10.I] said all shutdown rods shall be fully withdrawn in Modes 1 & 2, while proposed Tech Spec '

says all withdrawn at least 320 steps. Examination ' of 4 Figure 3.10-1 confirms that this- is' .an ' equivalent i requirement.

3.10.2.6 This Tech Spec is equivalent to current Tech Spec 3.10.D.ii.

3.10.2.7 This is a proposed Tech Spec which NUSCo says is less restrictive than the current Tech Spec. However, the i values for control group insertion limits .in this proposed Tech Spec were used in the analysis _and had .no impact.

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l Thus these changes are acceptable with - respect to the j analysis. ITS did not otherwise review these proposed '

changes.

3.11 This Tech Spec was modified to reflect new physics parameters. Analysis . supported this Tech Spec change, j

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3.13 This Tech Spec was modified to reflect new physics l

parameters. Analysis supported this Tech Spec change. t j

3.17.1.1 This Tech Spec is equivalent to current Tech Spec j 3.18.B.1.1.a with new figures based on the reload, which are very simil6r to the current figures and have been regrouped into figures use from 0 to 250 EFPD and 250 EFPD to E0L as compared to 1-125, 125-250 and 250-E0L.

3.17.1.2 This Tech Spec is equivalent to current Tech Spec  !

3.18.8.1.1.a j 3.17.2.1 This Tech Spec is identical to current Tech Spec 3.17.A.

3.17.2.2 This Tech Spec is identical to current Tech Spec 3.17.D.

3.17.3.1 - 3.2 These are proposed Tech Specs which define limits of the Nuclear Enthalpy Rise Hot Channel Factors. [The radial peaking factors delineated in Table 4.0-1 of Rev 2 [3]

were used in the computations and agree with these Tech 3 Specs]. '

3.17.4 This Tech Spec is the same as current Tech Spec 2.4 (4).

1 3.17.5 This Tech Spec is similar to current Tech Spec 3.20, but i Tc old is raised from 540.6. to 542 and this is less restrictive. The RCS flowrate is lowered from 257,000 gpm to 246,000 gpm as indicated by NUSCo in their reanalysis ,

l to reflect a corrected core by-pass flowrate and to provide for some SG tube plugging. This aspect of Tech Specs is more restrictive. In addition, current more restrictive Tech Specs are provided for-3-loop operation.

These Tech Specs are acceptable.

3.24 These Tech Specs relate to special exceptions to the Tech Specs for test conditions, and were not reviewed by ITS.

)

i Five of the proposed Tech Specs are less restrictive; they are (i) 3.10.1.2 - shutdown margin for Mode 3, (ii) 3.10.2.7 - control group insertion limit for three-loop operation, (iii, iv) 3.17.1.1 and 2 axial offset limits (0-125 EFPD) for both four and three-loop operations and (v) 3.17.5 - DNB parameters (Tcold). Our evaluation of these five is as follows:

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(i) This change is less restrictive only during the BOL when the maximum stuck rod worth is potentially less than 1200 pcm. The design basis accidents affected by this are the boron dilution accident and SLB with four-loops on. This is an acceptable Tech Spec change for the purposes of restart and for operation during the early portion of Cycle 15; provided, however, that operation beyond mid-cycle is not acceptable unless the reanalysis of the SLB accident supports it.

(ii) Since this value was used in the reanalysis and obtained acceptable results, this proposed Tech Spec. is acceptable.

(iii & iv) The revised axial offset limits in the 0-125 EFPD burn-up  ;

range is impacted for four-loop (-11 from -15) and three-loop (-7 from -12) {

and would not impact the centerline fuel temperature. Therefore, this i proposed Tech Spec change is acceptable. l (v) Use of 544*F is conservative, and is therefore acceptable.

4 d

i I

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TABLE 2 1

\

Review of Technical Specifications Haddam Neck Plant TS No. Description Nature of Change Category 1.11 A Channel Functional Test 0 1.12 Core Alteration 0 1 1.13 Shutdown Margin definition 2 1.14 Identified Leakage 0 Unidentified Leakage 1.15 0

~ Table 1.1 Operational Modes 2.0 Safety Limits and Maximum Safety Settings 2.1 Introduction 2.2 Safety Limits format change Figure 2.1-1 & 2 new methodology 4 reduced RCS flows  :

2.4 Maximum Safety Settings I (5) - LOF 3 (8) - Rod Withdrawal 4 1 3.3.1.1 Startup and Power Operation (Modes 1 & 2) 2  :

3.3.1.2 Hot Standby (Mode 3) Rod Withdrawal 2 3.3.1.3 Hot Shutdown (Mode 4) Rod Withdrawal 2 l 3.3.1.4.1 Cold Shutdown (Mode 5) Rod Withdrawal 2  !

3.3.1.4.2 Cold Shutdown (Mode 5) Rod Withdrawal 2 3.3.1.5 Isolated Loop (Modes 1,2,3,4,& 5) 311 ,

3.3.1.6 Isolated Loop Startup (Modes 3,4,5 & 6) 311 3.3.1.7 Idled Loop 3ii l 3.3.1.8 Idled Loop Startup 311 3.3.2.1 Safety Valves Shutdown 2 l 3.3.2.2. Safety Valves 2 3.3.3 Pressurizer added figure 3.3-1 1 3.3.4.1 Relief Valves new setpoint 2 l

3.3.4.2 LTOPS 2 1 3.5 CVCS format 5 3.7 Min. H2O Volume and Boron Concentration format 5 3.9 Operational Safety Instrumentation & Control Systems Table 3.9-1 Startup Requirement Rod Withdrawal 2 3.10 Reactivity Control System 18 1

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l 1

i l o  ;

l .q- 6r' l

3.10.1.1 Shutdown Margin (4-loop, Modes 1 & 2) SLB- .

2  ;

3.10.'1.2 ' Shutdown Margin (Mode 3) SLB/ Boron Dilution 2- l 3.10.1.3 Shutdown Margin (Modes 4 & 5) Boron Dilution 2

- 3.10.1.4 Shutdown Margin (3-loop, Modes 1 &~2) SLB 2-3.10.1.5 Moderator Temperature Coefficient 1 3.10.1.6 Minimum Temperature for Criticality 31~

3.10.2.1 Movable Control Assemblies -

1 ,

3.10.2.2 Position Indication _ Systems-Operating 1 l 3.10.2.3 Position Indication System-Shutdown 1

'l 3.10.2.4 Rod Drop Time 1

- 3.10.2.5 Shutdown Rod Insertion Limit 1 3.10.2.6' Control Grp Insertion Limit (4-Loop? 1 3.10.2.7 Control Grp Insertion Limit (3-loop) physics 2-3.11 Containment consistency 5 '

3.13 Refueling Boron Dilution 5 3.17 Power Distribution Limits I 3.17.1.1 Axial Offset Physics 2-3.17.1.2 Axial Offset .

Physics 2- l 3.17.2.1 Linear Heat Generation Rates 1 j 3.17.2.2 Linear Heat Generation Rates 1 i

- 3.17.3.1 Nuclear Enthalpy Rise Hot Channel Factor 2 3.17.3.2 Nuclear Enthalpy Rise Hot Channel Factor 2 3.17.4 Quadrant Power Tilt Ratio 1 1 3.17.5 DNB Parameters: LOF 5- .

3.24 S3ecial Test Exceptions  !

3.24.1 Slutdown Margin 311 l 3.24.2 Physics Tests 3ii i '

3.24.3 Position Indication System-Shutdown 31i 4.9 Main Steam Isolation Valves format - SLB 5

1. Those changes introduced in this submittal simply to conform to the 1 Westinghouse Standard Format and reflect' no. change in substance from the current technical specification in place. '

2. Those changes to make requirements more (or less, indicated by " ")

restrictive.

3. Those specifications which do not exist in the current Technical  !

Specification but either (i) required in order to conform- to -the i Westinghouse Standard Format or (ii) needed to provide operational I procedures.

4. Those required by the reload to reflect the change in the physics of the reloaded core and the change in the hydraulic conditions.

5. Other miscellaneous items, such as to be consistent with SRP, etc.

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