ML20236S377
| ML20236S377 | |
| Person / Time | |
|---|---|
| Site: | Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png |
| Issue date: | 11/12/1987 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20236S372 | List: |
| References | |
| NUDOCS 8711250114 | |
| Download: ML20236S377 (22) | |
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UNITED STATES
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k NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20655 g
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT N0. 97 TO FACILITY OPERATING LICENSE N0. DPR-61 CONNECTICUT YANKEE ATOMIC POWER COMPANY HADDAM NECK PLANT DOCKET NO. 50-213 l
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1.0 INTRODUCTION
By letter dated June 1,1987, as modified by letter dated July 22, 1987, the Connecticut Yankee Atomic Power Company (CYAPC0, 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 i
15 and reflects major efforts in upgrading design basis analyses of accidents I
and in refonnatting the existing technical specifications as part of the planned conversion to the Westinghouse standard technical specifications.
More specifically, the license amendment deletes existing Technical Specifi-cations 3.15, " Reactivity Anomalies"; 3.16. " Isothermal Coefficient of Reactivity"; 3.18, " Power Distribution Monitoring and Control"; and 3.20,
" Reactor Coolant System Flow. Temperature and Pressure." The infonnation 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 3.17, leteness
" Limiting Linear Heat Generation Rate"; as required, to assure comp (3.24 with the existing technical specification. One new specification "Special Test Exceptions") has been added to formalize the special test j
exceptions required to perform various startup physics tests.
i 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 J
Refueling Water Storage Tank"; and 4.9, " Main Steam Isolation Valves" have I
been revised to account for the revised design basis analyses in support of I
the safe operation of the Haddam Neck Plant for Cycle 15.
2.0 BACKGROUND
By letter dated June 1,1987, as modified by letter dated July 22, 1987, I
Connecticut Yankee Atomic Power Company submitted revisions to the Haddam Neck Plant TS to reflect planned operation during Cycle 15.
f 8711250114 871112 PDR ADOCK 05000213 r
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The non-LOCA accident analysis methodology, including the use of RETRAN l
and VIPRE, was submitted for staff review by letter dated August 1, 1984.
J The physics methodology for Cycle 15 was provided for staff review by letter dated September 12, 1986. The plant design basis analyses using l
the referenced methodologies were submitted by letters dated June 30, 1986, 1
March 10, and May 8, 1987.
l The VIPRE methodology for thennal-hydraulic parameters was reviewed and a l
l Safety Evaluation forwarded to CYAPC0 by letter dated October 16, 1986, 1
l The core physics methodology has been reviewed and a Safety Evaluation
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l forwarded to the licensee by letter dated August 3,1987. -
1 The staff review and approval of the RETRAN transient methodology and the J
. limiting design bases transient analyses as they' apply to Cycle 15 operation I
at the Haddam Neck Plant is described in Section 3.A of this Safety 1
I Evaluation.
-3.0 EVALUATION A.
Cycle 15 Technical Review
1.0 INTRODUCTION
By letter dated June 1,1987, CYAPC0 submitted 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 nuclear design and accident analyses.
The licensee's objective for this submittal was to demonstrate that the limiting conditions resulting from the reanalyses of the non-LOCA design basis accidents are properly reflected in the proposed TS and to support its por.E'On that the Haddam Neck Plant can be operated safely at the rated power we<el of 1825 MWt throughout Cycle 15.
I 2.0 EVALUATION 2.1 Reload Description The Cycle 15 core consists of 157 15x15 fuel assemblies, each containing 204 fuel rods, 20 control rod guide tubes and 1 instrument guide tube.
All fuel pins have the same outside diameter of 0.422 inches. Four lead test assemblies contain zircaloy clad fuel rods (batch ISB) with a clad wall thickness of 0.027 inches compared to 0.0165 inches for the stainless steel 304 clad fuel rods of all other assemblies. The four assemblies containing zircaloy clad fuel rods were first loaded into the core for Cycle 13. A LOCA analysis for these assemblies was performed at that time.
l The minimum uranium theoretical density is 94.9% for all Cycle 15 fuel.
1 The initial uranium-235 enrichment for all new Cycle 15 assemblies is 4.0 l
percent. The assembly loading is of the out-in type, i.e., with the new I
assemblies loaded on the periphery of the core. This is a high neutron leakage loading.
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- 1 All 45 control rods are being replaced in Cycle 15 due to previously l
' observed cracking and wear problems. The new control rods are neutronically i
identical with the old except for. extended life features such as Inconel i
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, ISB and 16 and the 56 new assemblies of batch 17.
In addition, one once-burned assembly from batch 15C is to be reinserted in the core's central position. All fuel assemblies are mechanically interchangeable. The four zircaloy clad assemblies (batch ISB) have a 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 (fiscussed below for cladding collapse, stress, strain and fatigue.
Batch 15A is estimated to be the most limiting in terms of core exposure time and burnup to the end of Cycle 15. The power histories of all asserr.blies were analyzed to determine the worst power history for creep collapse.
All Cycle 15 assemblies were analyzed under the worst power history conditions.
For all five fuel batches, analyses predict collapse times and burnups exceeding the maximum expected residence time and burnup in Cycle 15.
For batch 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 15 will be 28,190 EFPH. The analysis was performed with TAC 02, which is an approved code, and the staff has determined that the cladding creep collapse analysis is acceptable.
All fuel rods were evaluated for stresses following the ASME guidelines for pressure vessels which require that the primary membrane stress intensity must be less than two-thirds of the minimum unirradiated yield strength.
In all cases, the calculated stress values are acceptable.
The fuel design criteria specify a limit of one percent or less on cladding plastic tensile circumferential strain. The licensee performed an evaluation of the cladding strain using the TAC 02 code and assumed the worst Cycle 15 heat generation rate and fuel burnup.
It was determined that the design strain limit is much higher than the predicted value for Cycle 15 and, therefore, it is acceptable.
The fatigue usage factor was calculated following the ASME pressure vessel code which defines a maximum allowable factor of 0.9.
Using conservative conditions for Cycle 15, the fatigue usage factor was found to be 0.2 for the stainless steel 304 cladding and 0.4 for the Zircaloy, which are acceptable.
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. i 2.3 Nuclear Design The Cycle 15 reload is the first reload which has been designed by Northeast l'
Utilities personnel for CYAPCO, one of their member utilities. The physics methodology is documented in the topical report NUSCO-152, " Physics Methodology for PWR Reload Design."
The major differences between the Cycle 15 design and the Cycle 14 design
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Il (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 (RTD) 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
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negligible effect on the fuel cycle design.
l 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 performed analyses to show that during Cycle 15, the limits of the moderator temperature coefficient, the Doppler temperature coefficient, the delayed neutron fraction, the prompt 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.
2.4 Transient and Accident Analyses Several physics parameters for Cycle 15 will be more limiting than for 1
Cycle 14. These parameters include reduced shutdown margin, increased differential rod worth, revised reactivity (Doppler and moderator tempera-
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ture) coefficients, and revised axial peaking factors. These changes i
required that six accident analyses be revised. These analyses are: the
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i uncontrolled rod withdrawal, dropped rod, main steam line break, boron dilution, loss of flow, and rod cluster control assembly (RCCA) ejection.
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The licensee stated that operation through 180 effective full power j
days (EFPD) of Cycle 15 is bounded by the results and assumptions of the j
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-_ 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 position and has concluded that the' existing SLB analyses would bound operation of the'Haddam Neck Plant for the first 180 EFPD of Cycle 15.
Operationbeyond180EFPDrequiresanapprovedsteamlinebreak(SLB) analysis.
2.4.1 Rod Withdrawal Transient 4
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 rod drive lift coils are energized, up to the P-7 interlock.
In addition, TS are proposed to reouire a different (more restrictive) number of operating RCS loops during subcritical conditions.
Uncontrolled rod withdrawal transient analyses were performed by NUSCO for four loop operation commencing at 100% power and 65% power, and commencing at 65% power for three loop operation and at subcritical 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 thennal 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:
(1) least negative Doppler and most positive moderator temperature coefficient (MTC) vs most negative Coppler and MTC; (ii) highest RCCA stuck out; and (iii) reactivity insertion rates up to 22.5 percent milli K per second (pcm/sec) (which was the maximum 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.
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No parametric computations were done to verify the accuracy of the RETRAN i
nodalization for this analysis. Nevertheless, since (1) this is a particularly short transient in which only the primary pressure, core flow, and inlet i
temperature and power are important, and since (ii) comparison by the i
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 f, hat the j
licensee's computation of core flow, core inlet temperature and primary j
pressure are accurately predicted by the RETRAN model.
Furthermore, the j
staff has reasonable assurance that core power was conservatively computed
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in this analysis.
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i The licensee concluded that the minimum DNBR is always greater than 1.3 and, therefore, tnat fuel thermal 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 rod withdrawal 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 performed a simple computation 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 conclusion that the minimum DNBR will not be challenged by the boron 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 from 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 becomes the controlling accident for shutdown margin af ter 180 EFPD. The staff has reviewed this issue and agrees with the licensee's assessment. The staff will evaluate the forth-coming reanalyses of the main steam line break accident to support plant operation after 180 EFPD.
However, the current analyses provide reasonable assurances that the proposed shutdown margins are conservative for the period between startup and 180 EFPD, 2.4.3 RCCA Ejection The RCCA ejection accident represents the potential accident with the most rapid reactivity insertion rate. The licensee's analytical methodology for this accident is similar to that described above for the uncontrolled rod withdrawal accidents, with the licensee stating that they employed the following conservatism:
(1) the limiting burnup dependent physics para-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 bank insertion at each power level; (2) assumed adverse xenon; (3) added margins to ejected rod worth to account for calculational uncertainties; t
(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 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.
The licensee performed analyses for four loop hot full power (HFP) and hot l
zero power (HZP) and three loop HFP and HZP operations. The licensee
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estimated that 18% of the fuel rods reached DNBR of less than 1.3, and no l
rod had fuel melting at the center line in the four loop HZP case and none i'
in the HFP cases.
The staff has reviewed the licensee's analyses and has concluded that the licensee's analyses of the RCCA ejection is acceptable.
2.4.4 Dropped Rod Accident The licensee's analysis for this event used the same methodology as described l
for the RCCA ejection and rod withdrawal accidents described above.
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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 rod signals, the licensee has decided to disable i
the turbine runback feature and, therefore, needed to reanalyze this transient.
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Parametric analyses were conducted varying the dropped rod worth from 0 to i
180 pcm (minimum to maximum expected 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 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.
I The licensee concluded that the minimum DNBR was obtained for the full power four loop operation crediting either the turbir.e runback or the rod 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 rod accident analyses for Haddam Neck are conservative and, therefore, acceptable.
2.4.5 Loss of Flow i
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.
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order to justify the lower proposed RCS flow rate TS, the licensee submitted the loss-of-flow (LOF) analysis assuming a lower core flow rate.
In addition, the new analysis included a change of the three loop flow trip i
setpoint in the governing TS.
The licensee stated that the flow rate requirements were based upon a steam generator plugging level consistent with 500 equivalent plugged tubes per steam generator, and that an evaluation has been performed with the core physics bypass flow fraction equal to 4.5 percent.
) 1 The licensee further indicated that the minimum RCS flow rate used is 246,000gpm(4-loop)and 194,000 gpm (3-loop) and that the core flow rate a
used in the VIPRE calculations was the same as that used in the RETRAN calculations.
The licensee analyzed the worst case LOF event (complete loss-of-flow from
'i full power), which results in the most severe power-to-flow ratio and,
" Thermal. Hydraulic Model Qualification Volume 1 (pical report NUSCO 140-1, therefore, the lowest DNB.
In Section II.E of to 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 I
of RCS flow loss coefficients. They concluded that variation parameters had a minimal impact upon DNBR calculated using the W-3L correlation and, therefore, that normal values could be used.
q Although 'no experimental data was presented to verify the flow coastdown curves for this transient, we believe that good comparison to the results j
in the Facility Description and Safety Analysis (FDSA) gives adequate q
assurances of accuracy of the RETRAN model to permit its use in this i
event.
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1 The licensee conservatively assumed that the transient commences from 102%
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power 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.
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l addition, the licensee assumed minimum initial RCS pressure and flow but assumed the maximum core inlet temperature.
Each of these assumptions is l
conservative, tending to lower the computed minimum DNBR.
In addition, J
the computed RCS pressure rise was minimized by assuming minimum initial j
pressurizer level, maximum initial SG level, maximum turbine stop valve I
i closure time, the pressurizer heaters off, maximum pressurizer spray flow, and the charging systems isolated but letdown flow available.
- Finally, reactivity insertion was maximized through the use of the least negative Doppler coefficient and the most positive MTC and the reactor trip comput-ation included instrument response and delay times.
CYAPC0 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, l
and since the minimum DNBRs computed with VIPRE-01 using the RETRAN output l
described above are above 1.3 for both analyses, the staff has adequate i
assurance that the required DNBR limits will be met. Therefore, the staff i
concludes that this analyses is acceptable.
In sumary, six transient and accident analyses have been reanalyr.ed by the l
licensee and found to be conservative and acceptable. The assumptions and l
conditions used in the analyses are found to be consistent with the proposed l-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 RETRAN models in these analyses is acceptable.
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Technical Specifications Given that extensive modifications have been proposed to the format of the existing technical specifications, the staff has reviewed the proposed i
Cycle 15 reload technical specifications to assure that the proposed
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technical specifications (1) retain all previous requirements, as appro-J riate, and (2) incorporate the appropriate parameters to reflect the revised i
safety analyses performed in support of Cycle 15 operation and which provide
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reasonable assurance of safe operation during Cycle 15.
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Table 1 contains a comparison of the existing technical specifications with
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the proposed new and reformatted technical specifications. This comparison l
assured that all existing requirements have been retained in the proposed j
reformatted technical specifications except for those which have been i
deleted as discussed below.
The proposed modifications to the technical specification sections are summarized below.
D.1 Section 1.0 " Definitions" Three changes have been proposed for Cycle 15 operation. The first is a changa to the definition of shutdown margin to reflect the Westinghouse metho,1 ology 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 MODES. The definition of the reactivity condition (Keff) for the refueling mode is being revised from the original technical specification value ofleTf 60.92 to a new value (Keff 6 0.94). The new Keff value is based upon a revised boron dilution accident analysis which was perfomed in support of Cycle 15 operation and is more restrictive than the current Westinghouse TS (Kefff0.95). As discussed earlier, this accident has been reviewed by the staff and was found to be acceptable.
In addition, this amendment proposes to revise the definition of Operational I
MODES 2-5.
Even though the numbers in Table 1-1 change, the reactivity conditions for MODES 2-5 in the current TS are identical to the reactivity conditions proposed for Cycle 15. The reactivity condition definitions have been intended to identify whether the core is critical (MODES 1 and 2) or subcritical (MODES 3-5) within the traditional 1% delta K/K reactivity uncertainty allowance. Since shutdown margin may be provided by control j
rods actually inserted, control rods available to insert and/or soluble i
boron, the core reactivity condition 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 consistent c'efinition of shutdown margin and shutdown margin requirements that were identified by l
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 support of Cycle 15 operation. The staff has found the submitted analyses to be acceptable, as described above, and therefore the staff concludes that the proposed modifications are acceptable.
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_ _. l B.2 Section 2.0, " Safety Limits and Maximum Safety Settings" The licensee has proposed two changes to Section 2.2, " Safety Limits -
Reactor Core." The first change is to provide Applicability and Action i
Statements to this specification. The existing specification currently does not include these iterus. The staff has reviewed these additions and l
found them to be consistent with plant operation and the Westinghouse 1
Standard Technical Specifications (STS) and, therefore, concludes that the propored additions are acceptable.
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 VIPRE code as well as a measured reduction in reactor coolant system (RCS) flow rate. The minimum departure from nucleate boiling ratio (MDNBR) limit and the core exit void fraction limit values remain unchanged. The Bases for Specification 2.2 have also been revised to reflect the use of the newer code (methodology) and includes a discussion of the design peaking factors for both three-and four-loop operation.
The staff has reviewed the proposed changes and concludes that the licensee has used previously approved methodology in calculating the revised figures.
Further, the licensee has assured that all previously approved design 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-4 cation is a lowering of the reactor trip setpoint for low coolant flow.
The lower setpoint is directly obtained from the reanalysis of the loss-of-flow accident. The staff has reviewed the revised loss-of-flow accident and found the analyses to be acceptable (see Section 3.A.2.4.5 above). The licensee has also added a new trip setpoint for a high start-up rate reactor trip. This trip is from safety-related equipment and is necessary to assure that assumptions made in the revised analysis of the control rod withdrawal from subcritical accident are preserved. The revised control rod withdrawal from subcritical accident for Cycle 15 has been reviewed and found to be acceptable (see Section 3.A.2.4.1 above).
The last modification to this section is the deletion of the requirement for determining the quadrant power tilt ratio (QPTR). The requirement to maintain a QPTR less than 1.02 has been deleted from Section 2.4 but has been preserved as Specification 3.17.4 in the revised format.
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 reactor trip.
The staff has reviewed the proposed modifications to Sections 2.2 and 2.4 and the Bases to Section 2.4.
The proposed modifications are consistent with the revised analyses which have been reviewed by the staff and found to be acceptable. Therefore, the staff concludes that the proposed modifica-tions to Section 2.0 are consistent with the analyses in support of Cycle 15 operation and are acceptable.
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1 B.3 Specification 3.3, " Reactor Coolant System" Proposed Specification 3.3, " Reactor Coolant System," replaces in entirety the current Specification 3.3, " Reactor Coolant System Operational Com-l ponents," with several exceptions as described below. As stated earlier.
l Table 1 contains a one for one listing of the existing requirements of the l
current TS and the reformatted and renumbered section in the proposed TS for Cycle 15. Review of this table has assured that all existing specifica-l tions, surveillance requirements and action statements, have been preserved l
in the proposed reformatted 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 l
power levels. These specifications are not consistent with the design reanalyses performed in support of Cycle 15.
Further, the number of reactor i
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 power levels greater than 1 percent.
Proposed Specification 3.3.1.1 would require a larger number of reactor coolant loops operating than previously required 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 S is acceptablea.
Current Specification 3.3 H was a one-time requirement during the 1981 1
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 demonstrated during the 1981 outage and, therefore, this specification is no longer l
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 discussed above, the proposed reformatted Specification 3.3 has added new applicability, action I
statements 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.
j 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 i
verified that all previous requirements, with the exception of four items l
discussed earlier, have been retained in the proposed reformatted specifi-cation, and (2) that the proposed specification is supported by the plant design safety analyses performed for operation during Cycle 15. Therefore, j
the staff concludes that the proposed revisions to Specification 3.3 are j
acceptable.
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B.4 Specification 3.5, " Chemical and Volume Control System" Specification 3.7, " Minimum Water Volume and Boron Concentration in the Refueling Water Storage Tank" Specification 3.11. " Containment" Specification 3.13, " Refueling" Consistent with the discussion in Section 3.B.1 above, the licensee calculated the required shutdown margin for Cycle 15 using assumptions which have previously been reviewed and approved by the staff for other l
Westinghouse designed pressurized water reactors. The shutdown margin requirements determine the minimum refueling boron concentration required to assure that the Keff values given in Table 1.1, OPERATIONAL MODES, can be satisfied. Modifications to the above specifications have been made I
to reflect the revised shutdown margin analyses for Cycle 15. As mentioned l
earlier, the revised shutdown margin calculation has been reviewed by the staff and found to be consistent with currently accepted practice and was, therefore, acceptable.
1 The licensee has also proposed modifications to the Bases of the above I
sections to reflect the revision to the shutdown margin analyses.
The licensee has also deleted from the Bases section of Specifications 3.5 and 3.7 references to Cycle 1 boron rod worth data because it is no longer applicable to the Cycle 15 core.
The staff has reviewed the proposed changes described above and has concluded that the proposed modifications 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 reactor trip.
The proposed requirement for the start-up rate trip requires that both start-up rate channels (1/2 logic) be operable due to single-failure considerations.
The present design of the variable low-pressure reactor trip logic 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 (duetomaintenanceorsurveillance). Should a channel be declared inoper-able, the inoperable channel is placed in the " tripped position and an) trip signal frca the remaining two channels would result in a reactor trip (that is half logic).
During this 1987 outage, CYApC0 has installed an additional channel for variable low-pressure reactor trip, however the current logic requirement of needing two separate channel trips to initiate
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reactor trip has remained unchanged. CYAPC0 has requested a change to l
Table 3.9-1 to reflect the change in the minimum functional requirements to keep the existing trip logic unchanged. As mentioned earlier, should a trip channel be declared inoperable, that channel is placed in the j
" tripped" position and the resulting requirement for a reactor trip is any 1
trip signal from one of the three operable channels (i.e.,1.3 logic).
Because the existing requirements for reactor trip from 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 power level for operation for the low coolant flow reactor trip is being revised from 84 percent power to 74 percent power.
This revision prevents the potential for operating above the allowable 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, therefore, acceptable.
B.6 Specification 3.10. " Reactivity Control" 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, " Isothermal Coefficient of Reactivity."
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 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 Specifications 3.10.B. 3.10.C and 3.10.F.3.
The bases for their deletion are discussed below.
Current Specifications 3.10.B 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 generation of Westinghouse plants. The development of the Westinghouse reload methodology and procedures, reduced the need for a TS requirement for includes all of the physics requirements (ysis checklist for Haddam Neck ejected rod worth. The reload safety anal e.g., reactivity worth, power 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.
{ i j
u 1
Current Specification-3.10.F.3 defines an action required if more than one j
rod position indicator (RPI) and/or its group' position indicator are inoperable. The bases for operation in the above configuration (more than one rod 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 Specification 3.10.F.3 is acceptable.
3 In addition to the proposed deletions described above, proposed Section 3.10 contains revised requirements for the three-loop control rod insertion limits and four-and three-loop shutdown margin.
The shutdown margin requirements for MODES 1 and 2 for three-and four-loop operation are based on the results of the revised steamline break accident of 1800 pcm (four-loop) and 2600 pcm (three-loop) proposed shutdown ma submitted in support of Cycle 15 operation. The come directly from the revised analyses'and assure that the assumptions used in the safety analyses will be preserved for Cycle 15. The requirements for MODES 3, 4, and 5 l
are established by the boron dilution accident. Both accident analyses l
have been revised to reflect Cycle 15 parameters and have been reviewed by the staff and found to be. acceptable for 180 EFPD.
The current requirements for four-loop control rod insertion, control rod alignment, RPI and step counter operability, control rod drop time, and shutdown bank withdrawal requirements are unchanged, and the current moderator temperature coefficient (Specification 3.16) and reactivity anomalies (Specification 3.15) requirements have been transferred 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 analyzed range, (2) the reactor trip instrumentation is within its normal operating range. (3) the pressurizer
(
is operable with a steam bubble, and (4) that shutdown margin requirements are met.
i Further, the proposed reformatted technical specification contains new applicability statements, action statements and surveillance requirements.
New limiting conditions of operation have also been proposed to assure that the assumptions used in the Cycle 15 safety analysis are preserved.
These new requirements reflect the current guidance contained in the Westinghouse 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 proposed Specification 3.10 and concludes that (1) a comparison of the current specifications and the proposed limiting conditions of operation have verified that all previous requirements, with the exception of the three items discussed earlier, have been retained in i
t-( the proposed reformatted specification, and (2) that the proposed speci-fication is supported by t1e 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.
i B.7 Specification 3.17, " Power Of stribution Limits" Proposed Specification 3.17. " Power Distribution Limits," replaces in entirety current Specification 3.17, " Limiting Linear Heat Generation Rate (LHGR)"; Specification 3.18, " Power Distribution Monitoring and Control";
'i and Specification 3.20, " Reactor Coolant System Flow, Temperature and Pressure." As mentioned earlier, proposed Specification 3.17 alto includes the specification on quadrant power tilt from the footnote in current Specification 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 16.
Review of this table has assured the staff that all existing specifications, surveillance require-ments, and action statements have been prei cd in the proposed technical specifications with two exceptions. These an current Specification 3.18.B.1.2 which describes the use of the thimble correlation method for core power distribution monitoring and Specification 3.18.C.2 which required control rod bank B to be completely withdrawn from the core above 20 percent power.
i Tte bases for deletion of the two specifications are discussed below.
The thimble correlation method of monitoring power distribution has not been used and has become obsolete in light of installation of the new plant process computer. Deletion of Specification 3.18.B.1.2 is, therefore.
acceptable.
The licensee also proposes to delete Specification 3.18.C.2.
This speci-ication was intended to provide a means for controlling Xenon induced axial i
power shapes (and, therefore, peak linear heat generation rate) during normal 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 normal core monitoring activities.
Based on the use of approved axial offset moaitoring 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.
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i i 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 ta the four-loop inlet temperature. The design basis accident reanalysis ano safety limit curves for three and four-loop RCS have assumed core flow rates of 233,870 and 184,730 gpm..for four and three-loop operation, respectively, for all accidents.
The revised technical specification restricts the analytical i
use of RCS flow rates to 246,000 and 194,000 gpm for four-and three-loop operation, respectively, thereby inswing reasonable margin to the values used in the safety analyses.
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-sponds to an F H value of 1.656(1.78 divided by the engineering factor and 1.64. Th proposed full-power values of 1.60 (four-loop operation) of1.075)(.three-loop 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 rod insert-ion).
l In addition, the proposed refonaatted TS has added new applicability, action statements and surveillance requirements. New limiting conditions of operation have been proposed to assure that the assumptions used in the Cycle 15 safety cralysis are preserved. These new requirements reflect the current guidarece contained in the Westinghouse STS and have been modified to reflect the Cycle 15 design safety analyses or the Haddam Neck Plant equipment, as appropriate.
3.17andconcludhskhui The staff has reviewed the proposed Specification 4 (1) a comparison of the current specifications and the proposed limfting conditions of operation have verified trat all previous requirements, with the exception of the two itelrs discussed earlier, have been retained in the proposed reformatted 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 acceptat'le by the staff.
Therefore, the staff concludes that the proposed revisions to Specification 3.17 ar.t:
acceptable.
B.8 Specification 3.24, "Special Test Exceptions" Specification 3.24 is a new technical specification that formalizes ent;p-tions to requirements in order to perform various low-power start-up physics tests.
f The new specification includes formal test exceptions currently identified
{
in Specifications 3.10. A 3.10.0 and 3.18.C for shutdown margin, moderator i
temperature coefficient, minimum temperature for criticality, control rod
. j alignment, control rod insertion (shutdown and contro~ banks), and control rod position indication. Tfk staff has reviewed the new specification and concludes that the proposed specification is consistent with the current L
i guidance contained in the Westinghouse STS and has been modified to reflect the Cycle 15 design safety analyses or the Haddam Neck Plant equipment, as appropriate. The staff concludes that the proposed specification is acceptable.
q
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B.9 Specification 4.9, " Main Sth w Isolation Valves" The current specification requires that a closure time of 10 seconds for the MSIVs be verified each cold shutdown if it has not been tested in the previous three months. The proposed specification clarifies the closure j
time requirement by specifying simultaneous closure of all four valves within 1
10 seconds, which is consistent with the assumptions made in the design bases analysis.
The staff has reviewed the proposed change and concludes that it is acceptable.
C.
Suninary The staff has reviewed the licensee's submittal in support of Cycle 15 operation and the proposed 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 are affected by the Cycle 15 reload and the proposed TS changes for consistency, adequacy and completeness. Our review found that the reanalyses of these six accidents was performed conservatively and in a manner consistert with the proposed TS. The set of accident analyses submitted by the licensee bounds other accidents or transients which would be impacted by changes described above, and, therefore, constitutes a complete set of all accidents which are required to be reanalyzed. We also found that f
the required changes to the TS imposed by the new physics parameters of
)
the reload core are based upon the results of the accident analyses which were performed in a conservative manner.
The licensee stated that operation through 180 EFPD of Cycle 15 is bounded by the results and assumpt h ns of the current design basis steam line break analysis submitted in 1980 snd previously approved by the NRC. Therefore, approval of the revised steam line break analysis is not required for Cycle 15 startup or the period between startup and 180 EFPD. However, operation beyond that point requires an approved steam line break analysis.
4.0 ENVIRONMENTAL CONSIDERATION
This amendment involves a change to a requirement with respect to the installation or use of facility components located within the restricted area as & fined in 10 CFR Part 20 and changes to the surveillance requirements.
The staff has determined that the amendment involves no significant increase 1
)
)
'l t
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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 Comission has previsouly issued a proposed finding that this amendment involves no sigrNficant hazards consideration and there has been no public coment on such finding.' Accordingly this amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to J
10 CFR 51.22(b) no environmental impact statement or environmental assessment need he prepared in connection with the issuance of this amendment.
5.0,(INCLUSION The staff has concluded, based on the considerations discussed above, that
.(1) there is reasonable assurance that the health and safety of the public will
^not be enhngrered by operation in the proposed manner, and (2) such activities
- ,11 be conducted in compliance with the Comission's regulations and the iuuance of tMs amendment will not be inimical to the comon defense and security or to the health and safety of the public.
6.0 ACKNOWLEDGEMENT This Safety EveJu6 tion has been prepared by F. Akstulewicz of PDISA, NRR in coordination with C. Liang, G. Hsii and L. Lois, all of XRSB, NRR and XRSB
-e technical contractor - International Technical Services, Inc.
1 7.0 FM ERENCES 1.
"WFC0 Tumal Hydraulic Model Qualification Volume I (RETRAN)," NUSCO 14M 1, July 30, 1984 2.
"6ddan,: Neck Plant Non-LOCA Transient Analysis," NUSCO 151.
3.s
'Haddam Neck Plant Cycle 15 Reload, Technical Specification Change Requests and Reload Report," letter from E. J. Mroczka (NUSCO) to U.S.
Nur.lcar Regulawry Comission, June 1,1987.
4r "Nddam Mck Plant Revisions to Reanalysis of Non-LOCA Design Basis fccidents," letter from E. J. Mroczka (NUSCO) to U. S. Nuclear Re';utatory Comission, May 8,1987.
5.
"rladdam Neck Plant Additional Information - Reanalysis of Non-LOCA Design Basis Accidents," letter fron E. J. Mroczka (NUSCO) to U.S.
Nuclear Regulatory Comission, September 2,1987.
6.
" Safety Evaluation Report on the RETRAN Computer Code," U. S. Nuclear Regulatory Comission, July 1984.
7.
" Safety Evaluation by the Office of Nuclear Reactor Regulation Regarding l
t4USC0 Topical Report 140-2 VIPRE-01 Connecticut Yankee Atomic Power 1
Company Docket No. 50-213 Haddam Neck Plant," October 1986.
1 1
1 1
e 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.
" TACO 2, Fuel Pin Performance Analysis," BAW-10141 PA, Y. H. Hsii, et al., B&W Lynchburg, Virginia, dated June 1983.
11.
" Chapter.10 - Incidents and Potential Hazards," Final Design Safety Analysis, Connecticut Yankee Atomic Power Company, May 1966.
12.
NUSCO-155, " Connecticut Yankee Atomic Power Company. Haddam Neck Plant," Technical Report Supporting Cycle 15 Operation, dated June 1987.
13.
Letter from J. H. Taylor, Babcock and Wilcox, to J. S. Berggren, NRC, "BaW's Response to TAC 02 Questions," dated April 8, 1982.
Dated: November 12, 1987 I
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 Specification 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 Statement A 3.3.C 5)
(deleted) 3.3.C 6) 3.3.4.1 3.3.C 6)(a) 3.3.4.1 Action Statement A 3.3.4.1 Action Statement B 3.3.C(6)(b) 3.3.4.1 Action Statement A'
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3.3.4.1 Action Statement B 3.3.C (7) 3.3.3 Action Statement A 3.3.3 Action Statement B 3.3.0(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 3.3.E(2) 3.3.4.2 Action Statement B 3.3.E (3 3.3.4.1 Action Statement C 3.3.F(1 3.3.1.2 3.3.F (2 3.3.1.2 3.3.F (a) 3.3.1.2 Action Statement A 3.3.F (b) 3.3.1.2 Action Statement C 3.3.G 3.3.1.3 3.3.G 2) 3.3.1.4.1 3.3.1.4.2 3.3.6(2)(a) 3.3.1.3 Action Statement A 3.3.1.4.1 Action Statement A l
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 J
3.3.H (deleted)
]
I l
I o
c TABLE 1 CONTINUED Current Technical Specification Proposed Cycle 15 Technical Specification 3.3 - Section Section 3.10 3.10.A 3.10.2.6 3.10.2.7 3.24.1 3.10.0 (deleted) 3.10..C (deleted) 3.10.0(i) 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.D (11)(3) 3.10.2.6 Action Statement C 3.10.2.7 Action Statenent 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 Reo.
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 Surveillance Req.
3.17.2.2 3.17.2.2 Surveillance Req.
3.17.0 3.17.2.2 l
TABLE 1 CONTINUED Current Technical Specification Proposed Cycle 15 Technical Specification Section 3.18 3.18.A 3.17.2.1 3.17.2.1 Surveillance Req.
3.18.B.1.1.a 3.17.1.1 3.17.1.2 3.17.1.1 Surveillance Req.
3.17.1.2 Surveillance Req.
3.18.B.1.1.b 3.17.1.1 Action Statement 3.17.1.2 Action Statement 3.18.B.1.1.c 3.17.1.1 3.17.1.2 3.17.1.2 Surveillance Req.
3.18.B.1.2 (deleted) 3.18.C.1 3.10.2.6 3.10.2.7 3.18.C.2 (deleted)
Section 3.20 3.20.A 3.17.5.a 3.20.B 3.17.5.b 3.20.C 3.17.5.c 3.20.0 3.17.5 Surveillance Req.
3.20.E 3.17.5 Action Statement 3.20.E 3.17.5 Surviellance Req.
_ _ _ _ _ _ _ _ _ _ _