ML20214R608
| ML20214R608 | |
| Person / Time | |
|---|---|
| Site: | Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png |
| Issue date: | 06/01/1987 |
| From: | Mroczka E CONNECTICUT YANKEE ATOMIC POWER CO. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| Shared Package | |
| ML20214R609 | List: |
| References | |
| B12537, NUDOCS 8706080217 | |
| Download: ML20214R608 (21) | |
Text
i CONNECTICUT YANKEE ATOMIC POWFR COMPANY B E R L I N. CONNECTICUT P O. Box 270
- HARTFORD. CONNECTICUT 06141-0270 TELEPHONE l
us.. June 1,1987 i Docket No. 50-213 B12537 Re: 10CFR50.36 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555
References:
- 1. W. G. Counsil letter to 3. R. Miller and D. M. Crutchfield, NUSCO Thermal Hydraulic Model Qualification Volume I (RETRAN),
NUSCO 140-1, and NUSCO Thermal Hydraulic Model Qualification Volume 11(VIPRE), NUSCO 140-2, dated July 30,1984.
- 2. 3. F. Opeka letter to C.1. Grimes, Reanalysis of Non-LOCA Design Basis Accidents, NUSCO 151, dated June 30,1986.
- 3. E. 3. Mroczka letter to U.S. Nuclear Regulatory Commission, Revised Non-LOCA Design Basis Accident Analysis, Loss of Normal Feedwater Flow, dated March 10,1987.
- 4. E. 3. Mroczka letter to U.S. Nuclear Regulatory Commission, Revision to Reanalysis of Non-LOCA Design Basis Accidents, dated May 3,1987.
- 5. 3. F. Opeka letter to C. I. Grimes, Physics Methodology for PWR Reload Design, September 12,1986.
Gentlemen:
Haddam Neck Plant Cycle 15 Reload, Technical Specification Change Requests and Reload Report Pursuant to 10CFR50.90, Connecticut Yankee Atomic Power Company (CYAPCO),
hereby proposes to amend Operating License No. DPR-61 for the Haddam Neck Plant by incorporating the attached changes into the plant Technical Specifications. A description of the proposed changes is provided in Attachment 1. The revised pages are provided in Attachment 2.
CYAPCO has reviewed the attached proposed changes pursuant to the requirements of 10CFR50.59 and has determined that they do not constitute an unreviewed safety question. The basis for this determination is discussed in Attachment 3. CYAPCO has also reviewed the proposed changes in accordance with 10CFR50.92 and has concluded that they do not involve a significant hazards consideration. The basis for this 00 8706080217 870601 I PDR ADOCK 05000213 !
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U'.S'. Nuclear Regulatory Commission B12537/Page 2 June 1,1987 conclusion is that the three criteria of 10CFR50.92(c) are not compromised, a conclusion which is supported by our determination made pursuant to 10CFR50.59.
The Commission has provided guidance concerning the application of the standards in 10 CFR 50.92 by providing certain examples (51 FR 7751, March 6, 1986) of amendments that are considered not likely to involve significant hazards consideration. Example (iii) relates to a change resulting from a nuclear reactor core reloading, if no fuel assemblies significantly different from those found previously acceptable to the NRC for a previous core are involved. This assumes that no significant changes are made to the acceptance criteria for the Technical Specifications, that the analytical methods used to demonstrate conformance with the Technical Specifications and regulations are not significantly changed and that the NRC has previously found such methods acceptable. The attached proposed changes for Cycle 15 are similar to example (iii) in that they involve changes resulting from a reactor core reloading. The fuel assemblies are identical in design to those found previously acceptable to the NRC for Cycle 14. No significant changes have been made to the acceptance criteria for the Technical Specifications beyond those identified as being required as a result of the new design basis reanalyses of non-LOCA events. Although new analytical methods have been used to demonstrate conformance between the Technical Specifications and regulation, these methods have been employed to upgrade existing design basis analyses and analyze additional accident scenarios. The methods employed to perform the reanalyses of non-LOCA events were not done as a direct result of the requirements of Cycle 15 but rather as a voluntary upgrade to the design basis accident analyses. In summary, CYAPCO concludes that these charges do not involve a significant hazards consideration.
The Technical Report Supporting Cycle 15 Operation is provided in Attachment 4.
The Cycle 15 reload represents the first cycle to be licensed using analyses performed by Northeast Utilities Service Company (NUSCO) personnel on behalf of CYAPCO.
The non-LOCA transient analysis methodology employed for the Cycle 15 reload has been described in the Reference I submittal. The reanalysis of the non-LOCA design basis accidents using this methodology has been provided in References 2-4. The physics methodology used for the Cycle 15 design was provided in Reference 5.
As discussed during a telephone conversation on April 28, 1987, between NRC and NUSCO representatives, the Cycle 15 required Technical Specification changes ensure the assumptions made in the Cycle 15 design and the reanalyses of the non-LOCA design basis transients. Based on analyses performed and the review of the proposed revisions to Technical Specifications, it is concluded that the Haddam Neck Plant can be operated safely at the rated power level of 1825 MWt for Cycle 15. It should be noted that the attached proposed revisions to Technical Specifications assume the approval of another proposed revision to Techniqal Specificacions on the Low Temperature Overpressurization Protection system.(l> Should tl'at proposed revision not be approved, that submittal identifies the location of affected pages in this submittal which would require modification.
(1) See the E. 3. Mroczka letter to the U. S. NRC on the setpoint modification for the Low Temperature Overpressurization System dated June 1,1987.
__ _ _ _ _ _ A
U.S. Nuclear Regulatory Commission B12537/Page 3 June 1,1987 In accordance with 10CFR50.91(b), CYAPCO is providing the state of Connecticut with a copy of this proposed amendment.
Pursuant to the requirements of 10CFR170.12(c), enclosed with this amendment request is the application fee of $150.
We trust you find this information satisfactory and request review and approval of this amendment request by September 15,1987, in order to support the start of Cycle 15.
Very truly yours, CONNECTICUT YANKEE ATOMIC POWER COMPANY
+2u' E. 3.1)fodzka (/
Senio(Vice President cc: W. T. Russell, Region I Administrator F. M. Akstulewicz, NRC Project Manager, Haddam Neck Plant P. D. Swetland, Resident Inspector, Haddam Neck Plant Kevin McCarthy Director, Radiation Control Unit Department of Environmental Protection Hartford, Connecticut 06116 STATE OF CONNECTICUT )
) ss. Berlin COUNTY OF HARTFORD )
Then personally appeared before me E. 3. Mroczka, who being duly sworn, did state that he is Senior Vice President of Connecticut Yankee Atomic Power Company, a Licensee herein, that he is authorized to execute and file the foregoing information in the name and on behalf of the Licensees herein and that the statements contained in said information are true and correct to the best of his knowledge and belief.
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N/W JNotary Publi "
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Docket No. 50-213 B12537 Attachment 1 Description of Proposed Technical Specification Changes Haddam Neck Plant Cycle 15 June 1987 A
The proposed Technical Specification changes have been prepared to support the Cycle 15 reload. These changes accommodate the revised physics reload methodology, the Chapter 10 non-LOCA transient reanalysis, and the integration of reload-related Administrative Technical Specifications into the Safety Technical Specifications.
A major reorganization of the requirements of current Technical Specifications 3.3,
" Reactor Coolant System Operational Components," 3.10, " Reactivity Control," and 3.17, " Limiting Linear Heat Generation Rate," was required. This reorganization was based on the content of the Westingi ;se Standard format. Several of the current Technical Specifications (e.g., 3.15, 3.18, and 3.20) will be deleted since the require-ments have been transferred to the reorganized Specifications. One new Specification will be added to formalize the special test exceptions in order to perform various start-up physics tests.
The proposed changes are dicussed below:
1.0 DEFINITIONS The Shutdown Margin definition (1.13) is being revised to be consistent with the Standard Review Plan; i.e., all rods in, minus the maximum worth stuck rod.
Table 1.1 - Operational Modes - The reactivity condition (Keff) for the refueling mode is being revised from Keff f 0.92 to Keff .<_0.94. The current requirement is based on an "all rods in" core configuration and established the initial core load boron concentration of 2470 ppm (FDSA Section 4.2.2.1 and 10.2.3.2). This requirement has been interpreted as a shutdown margin requirement of 8 percent 4k/k in current Technical Specifications 3.5 and 3.11.
l l
The new requirement of Keff 10.94 is based on the boron dilution accident in MODE 6, which requires a shutdown margin of at least 5500 pcm. This corresponds to a Keff 6 0.9479. A Technical Specification requirement of Keff 5 0.94 was conservatively established which is more restrictive than Standard Westinghouse requirements (Keff f 0.95) and the requirement recognized by NUREG-0612 (Heavy Loads). The core configuration used to determine a required boron concentration for refueling is conservatively assumed to be "all rods out."
NUREG-0612 requires a Keff penalty of 0.05 to account for a heavy load crushing the core into a more reactive configuration. The heavy load for core crushing at Haddam Neck has been identified as the upper internals package. The upper internals package is only moved over the vessel before and after refueling; i.e., when the core is assembled (fuel assemblies, sources, control rods, and flow mixers in place). There-1 fore, considering the all rods in core configuration and the required 0.05 Keff penalty, aKeff requirement of Keff f 0.89 assures that af ter crushing the core will remain in a reactivity condition of Keff 10.94. The refueling boron concentration will be established by the more restrictive conditions of:
All rods out, Keff 10.94 or All rods in, Keff f 0.89
2.2 SAFETY LIMITS - REACTOR CORE The reactor core safety limit curves for four- and three-loop operation are being revised to account for a reduction in the RCS flow rate and a change in the calculational methodology.
The current safety limit curves are based on COBRA III-C methodology used by the Yankee Atomic Electric Company, submitted for the Cycle 6 reload and approved by the NRC as Amendment 3 to the license. The revised methodology is based on the VIPRE thermal hydraulic analysis code. This methodology was submitted and approved by the NRC.
The proposed specification includes Applicability and Action Requirements and a revised Basis. The current Specification does not include Applicability or Action Requirements. The proposed Specification requirements are consistent with the Westinghouse STS. The Basis is being revised to be consistent with the design FfH peaking factors for three- and four-loop operation. The MDNBR limit (1.30) and core exit void fraction limit (0.32) remain unchanged.
2.4 MAXIMUM SAFETY SETTINGS - PROTECTIVE INSTRUMENTATION The Nuclear Overpower and Low Coolant Flow trip settings will be revised. A new requiremer.+ for the High Start-Up Rate trip is being added.
The Quadrant Power Tilt Ratio (QPTR) requirement that appears as a footnote to the Nuclear Overpower trip is being deleted from this Specification. The QPTR requirement will be included in revised Technical Specification 3.17. The QPTR limit of 1.02 will remain unchanged.
4-The Low Coolant Flow requirement for three-loop operation is being revised to
> 34 percent of nominal three-loop flow. This change is being made so that the four-loop set point (2 90 percent nominal four-loop flow, in psid) will be valid for three-loop operation.
Currently, the indicated steam generator ap is higher during three-loop operation than four-loop operation. The milliamps (psid) set point must be adjusted to the
>90 percent nominal three-loop value upon entering the three-loop mode of operation.
Each channel must be taken out of service and power must be reduced below the P7 permissive value to perform the recalibration.
I 1 The proposed three-loop trip set point of ?_84 percent nominal three-loop flow was used in the reanalysis of the loss of flow accident.
The low-flow set point based on nominal four-loop flow is conservative for three-loop operation. Therefore, a power reduction below P7 for recalibration is not required when making a transition from four- to three-loop operation at power. Conversely, if the low-flow set point was initially established while in three-loop operation based on i 34 percent of the nominal three-loop flow, recalibration is required prior. to four-loop i operation since all loop set points would be nonconservative.
i A Specification for a High Start-Up Rate trip is being added due to the requirements of the rod withdrawal from subcritical accident. The reactivity insertion rate has increased from 90 pcm/ inch to 135 pcm/ inch, coupled with the RCS flow rate reduction for Cycle 15. The larger reactivity insertion rate and lower RCS flow rate requires that a reactor trip occur earlier in the transient due to the potential' power overshoot if the only protection is the Nuclear Overpower trip. The set point' of 4
5 decades per minute with a 1/2 logic will terminate- the transient sooner than the
- . - ~
5-Nuclear overpower trip, minimize the power overshoot, and assure acceptable DNB results. This trip will not be required if a rod withdrawal accident can be prevented by either opening the reactor trip breakers or de-energizing the control rod drive lift coils. The trip may be blocked above the 10 percent power interlock.
3.3 REACTOR COOLANT SYSTEM The current Specification provides multiple requirements for reactor coolant system operational components. These requirements are being reorganized into Westinghouse Standard format due to revised heat removal loop operability requirements as a result of the rod withdrawal from subcritical accident. Additionally, this reorganized Specification provides formal Applicability, Action, and Surveillance Requirements.
The integration of current requirements into the revised requirements is discussed below.
o Revised Specification 3.3.1.1 - This Specification provides requirements for the number of reactor coolant loops needed for operation in MODES I and 2. A reactor coolant loop includes associated piping, reactor coolant pump, steam generator, and loop isolation valves. A footnote in this revised Specification states that the loop out of service for three-loop operation may be idled or isolated. Three-loop operation will have formal Technical Specification require-ments for MODES I and 2 only. Current Technical Specifications do not differentiate between idled and isolated loop operation and do not define MODE Applicability for the three-loop requirements.
Current Specifications 3.3.C.4 and 3.3.C.5 require at least four reactor coolant pumps operating above 65 percent power and two steam generators operating. A potentially allowed configuration of 100 percent power, four reactor coolant
pumps, and two steam generators is not consistent with the design basis safety
. analysis. Similarly, current Specifications 3.3.C.2 and 3.3.C.3 potentially would allow critical operation without any reactor coolant pump and only two steam generators, and 9 percent power with only one reactor coolaat pump and only two steam generators operating. Neither of these potentially allowed operating configurations is consistent with the design basis safety analysis. Therefore, the current requirements in Specifications 3.3.C.2, 3.3.C.3, and 3.3.C.5 will be deleted.
o Revised Specification 3.3.1.2 - This Specification provides a requirement for the number of reactor coolant loops needed for operation in MODE 3. The current Specification 3.3F requirement of two loops operable /one loop operating is preserved provided a rod withdrawal accident can be prevented by either opening the reactor t*!p breakers or de-energizing the control rod drive lift coils, if a rod withdrawal accident cannot be prevented, three loops operable /two loops operating is required.
o Revised Specification 3.3.1.3 - This Specification provides a requirement for the number of heat removal loops (RCS or RHR) needed for operation in MODE 4.
The current Specification 3.3.G requirement of two loops operable /one loop operating is preserved, providing that a rod withdrawal accident can be prevented by either opening the reactor trip system breakers or de-energizing the control rod drive lift coils. If the rod withdrawal accident cannot be prevented, three loops operable /two loops operating is required.
o Revised Specifications 3.3.1.4.1 and 3.3.1.4.2 - These Specifications provide the requirements for the number of heat removal loops needed in MODE 5 for loops filled and loops not filled. The current Specification 3.3.G requirement of two
RHR loops operable /one RHR loop operating can be preserved for cases with reactor coolant loops not filled. For the case with reactor coolant loops filled and one RHR loop operating, two unisolated steam generators may be substituted for the other RHR loop operability requirement.
Current Specification 3.3.G allows the normal or emergency power source to be inoperable. The revised requirements would allow the emergency power source to be inoperable in MODE 5 with only one RHR loop operable and in operation if two unisolated reactor coolant loops were filled and the steam generators had a secondary side water level greater than 25 percent.
Revised Specification 3.3.1.ff.1 also requires that a reactor coolant pump in an unisolated loop shall not be started with one or more of the RCS cold leg temperatures _(3150F unless the secondary water temperature of each steam generator is less than 200F above each of the RCS cold leg temperatures. This requirement is equivalent to current Specification 3.3.D.
o Revised Specifications 3.3.1.5, 3.3.1.6, 3.3.1.7, and 3.3.1.8 - These new Specifi-cations provide requirements for idled and isolated loop operation. Current Technical Specifications do not provide any requirements for idled / isolated loop definitions or start-up requirements. The results of the isolated loop start-up accidrM require the reactor to be subcritical by at least 1000 pcm prior to an isolated loop start-up. Additionally,- the boron concentration in the idled / isolated loop must be greater than or equal to the boron concentration in the operating loops and the cold leg temperature of the idled / isolated loop must be within 200F of the highest cold leg temperature of the operating loops prior to idled / isolated loop start-up.
o Revised Specification 3.3.2.1 - This Specificatioli requires that a minimum of one pressurizer safety valve be operable with the appropriate set point in
MODE 4 unless the Low-Temperature' Overpressure Protection (LTOP) System is in service. The current Specification 3.3.A requires at least one pressurizer safety valve be in service if the RCS is above 3750F or 350 psig, except during.
hydrostatic tests.
o Revised Specification 3.3.2.2 - This Specification requires that all three pressur-izer safety valves ,be operable - in MODES 1, 2,' and 3. Current Specifica-tion 3.3.C.1 requires that three safety valves be operable in order to be critical (MODE I and critical portion of MODE 2). The. revised requirement is more restrictive than the current requirement.
o Revised Specification 3.3.3 - This Specification requires that pressurizer level remain within 1 5 percent of the programmed level and at least two operable groups of pressurizer heaters capable of. being powered ' from an emergency power source and each having a capacity of at least 150 kW. This requiremen is equivalent to current Specification 3.3.C.7.
o Revised Specification 3.3.4.1 - This Specification requires that all PORVs and associated block valves be opera
- ale in MODES 1, 2,' and 3, with a set point-between 2325 and 2350 psig. The emergency control air shall have a minimum pressure of 118 psig. Current Specification 3.3.C.6 requires that two PORVs and their associated block valves be operable, but does not specify a set point or control air requirement.
o Revised Specification 3.3.4.2 - This Specification requires that the LTOP system be operable with MODE Applicability Requirements. Current Specification 3.3.E provides the requirements of the RCS Overpressure Protection System.
Two of the current Specification requirements are not directly transferrable to the f
revised Specification. Current Specification 3.3.B requires that one or more reactor coolant pumps or an RHR pump be in operation when RCS boron concentration changes are being made. The revised Specifications for MODES 1-5 (3.3.1.1, 3.3.1.2, 3.3.1.3.,
3.3.1.4.1, and 3.3.1.4.2) all require at least one reactor coolant pump or RHR pump be in operation. All associated Action Statements require the suspension of all operations involving a reduction in the RCS boron concentration if no heat removal loop is in operation. Current Specification 3.3.H is a one-time requirement to allow demonstra-tion of a natural circulation cooldown and operator training. This requirement is no longer needed.
3.5 CHEMICAL AND VOLUME CONTROL SYSTEM The current Technical Specification requires a shutdown margin of 8 percent 4k/k prior to maintenance that requires draining the boric acid mix tank and requires that Valve BA-V-399 shall not be closed except when the reactor is shutdown and the RCS is borated to a Keff 10.92. The proposed revision changes the 8 percent ak/k shutdown margin and Keff 10.92 boration requirements to the refueling boron concentration. This revision will be consistent with the revised MODE 6 definition in Table 1.1. The Basis is being revised to delete a reference to Cycle 1 boron worth data.
3.7 MINIMUM WATER VOLUME AND BORON CONCENTRATION IN THE REFUELING WATER STCRAGE TANK
-I j
The current Technical Specification requires tnat the boron concentration in the RWST be not less than the refueling boron concentration. The revised transient analysis and LOCA analysis assume an RWST concentration of 2200 ppm and, therefore, establishes
the basis for the boron concentration requirement. The references to the Cycle 1 requirements will be deleted'.
3.9 OPERATIONAL SAFETY INSTRUMENTATION AND CONTROL SYSTEMS (TABLE 3.9-1)
The current Specification provides a 1/1 logic requirement for operation during start-up for the Intermediate Range SUR reactor trip. The new requirement for the Start-Up Rate trip in Specification 2.4 requires that both Start-Up Rate channels (1/2 logic) be operable due to single-failure considerations.
The current full-power operation logic requirement for the Pressurizer Variable Low-Pressure Reactor trip is 1/2. The revised logic requirement of 1/3 is due to the addition of a fourth channel of the Pressurizer Variable Low-Pressure Reactor trip.
This upgrade increases the functional operability of the Reactor Protection and Control System.
The current required action, if the logic required for full-power operation is not met for the Low Coolant Flow Reactor trip, is to maintain load below 84 percent full power. The revised requirement is to maintain load below 74 percent full power. This revision prevents the potential for operating above the allowable power level for three-loop operation if inadvertent three-loop operation -occurs af ter the loss of a single channel of the Low Coolant Flow Reactor Trip.
3.10 REACTIVITY CONTROL The current Specification includes requirements for the following:
o Power-dependent control rod insertion limit.
o Ejected rod worth.
o Control rod alignment.
o RPI and step counter operability.
o Control rod drop time.
- o. Shutdown bank withdrawal.
The revised Specification reorganizes the above requirements according to the Westinghouse Standard format for Section 3/4.1.1-Reactivity Control System and Section 3/4.1.3-Movable Control Assemblies. Revised requirements are provided for the three-loop control rod insertion limits and four- and three-loop shutdown margin.
The Cycle 15 desgn has been performed and safety analysis input parameters were developed based on these revised requirements.
The shutdown margin requirements for MODES I and 2 for three- and four-loop operation are based on the results of the steam-line break accident. The proposed shutdown margins of 1800 pcm (four-loop) and 2600 pcm (three-loop) have been shown to yield acceptable results. The requirements for MODES 3, 4, and 5 are established by the boron dilution accident. This accident has been reanalyzed for Cycle 15 with acceptable results for all possible loop operating configurations.
The current ejected rod worth requirements (3.10.A and 3.10.B) will be deleted since they are not included in the Westinghouse Standard format. The ejected rod worth and power peaking are evaluated against the rod ejection basis every reload.
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 will be unchanged, but placed in the Westinghouse Standard format.
Additionally, the current moderator temperature coefficient requirements (Specifica-tion 3.16) have been transferred to this revised Specification 3.10 and rewritten in the Westinghouse Standard format. The reorganized Specifications will provide formal Applicability, Action, and Surveillance Requirements.
A new Specification will be added to this section, consistent with the Westinghouse Standard format. The minimum temperature for criticality requirement of 5250F has been established to assure that the moderator temperature coefficient is within the analyzed range, the reactor trip instrumentation is within its normal operating range,-
the pressurizer is OPERABLE with a steam bubble, and shutdown margin requirements are met.
3.11 CONTAINMENT The current Specification for containment integrity requires that a shutdown margin of 3 percent Ak/k (all rods in) be maintained when the containment is open and the RCS is above 300 psig and 2000F. The new shutdown margin requirement of 2600 pcm for MODES 4 and 5 is more restrictive than the current requirement since a stuck rod must now be assumed.
The current Specification also requires that containment integrity not be violated with the reactor vessel head removed unless a shutdown margin greater than 8 percent 4k/k l
is maintained. This shutdown margin requirement is being revised to specify the refueling boron concentration for MODE 6.
l The shutdown margin basis in this Specification will be deleted since the shutdown margin basis is provided in the revised Specification 3.10.
1 3.13 REFUELING l
The current Specification requires that a boron concentration be maintained not less than the concentration required to shut down the core to a Keff of 0.92 (all rods in).
This Specification will be revised to require a boron concentration sufficient to maintain a Keff f 0.94 in all filled, unisolated portions of the RCS and the refueling canal. This revised Specification is more restrictive since the required boron concentration will be established by the more limiting of an all rods out core configuration (Keff f 0.94) or an all rods in core configuration (Keff 6 0.89). The Basis is being revised to reflect this change.
3.15 REACTIVITY ANOMALIES This Specification requires that the measured and predicted boron concentrations be monitored to assess the initial core reactivity and reactivity depletion rate. A discrepancy of I percent Ak/k was established as a limit that required a report to the NRC. This Specification will be deleted since the boron concentration tracking requirements will be included as an equivalent Surveillance Requirement in the revised Specification 3.10.
3.16 ISOTHERMAL COEFFICIENT OF REACTIVITY The current Specification provides requirements for the moderator temperature coefficient for the beginning and end of core life. This Specification will be deleted, but the equivalent requirements will be transferred to the revised Specification 3.10.
3.17 LIMITING LINEAR HEAT GENERATION RATE (LHGR)
The current Specification provides the LHGR limits for three- and four-loop operation.
These LHGR limits will be unchanged for Cycle 15. This Specification, however, will be expanded to include power distribution requirements provided in Section 3/4.2 for the Westinghouse Standard format.
The revised Specification will now include requirements for axial offset limits (currently in Specification 3.18), Quadrant Power Tilt Ratio (QPTR) (currently a footnote in Specification 2.4) and DNB parameters (currently in Specification 3.20). A new Specification will provide requirements for the Nuclear Enthalpy Rise Hot Channel factor (FAHN) for four- and three-loop operation.
The axial offset limits will be revised based on the results of the Cycle 15 design analysis. The revised limits will be reduced from a set of three curves to a set of two curves. The current set of curves for the 125-250 EFPD burn-up window will now be valid for the 0-250 EFPD burn-up range. This allows deletion of the current set of curves for 0-125 EFPD. The current set of curves for 250 EFPD to the end of cycle will be unchanged. The allowable axial power shapes that are provided as inputs to the safety analysis have been reviewed and yield acceptable results.
The Specification for QPTR is unchanged. The conversion to the Westinghouse Standard format has expanded the requirement to include more extensive Applica-bility, Action, and Surveillance Requirements.
The DNB parameters will include a revision to the RCS flow rate requirement for three- and four-loop operation and a revision to the four-loop inlet temperature. The
limit.for pressurizer pressure is unchanged. The analysis values of 1960 psia and 544.10F for pressure and temperature reflect changes in measurement uncertainty.
The proposed three- and four-loop RCS flow rates assume a 4.5 percent bypass flow fraction that has been reviewed and approved by the NRC. The Chapter 10 reanalysis and safety limit curves have assumed core flow rates of 233,870 and 184,730 gpm for four- and three-loop operation, respectively, for all accidents. . The 4.5 percent bypass flow fraction supports the RCS' flow rates of 246,000 and 194,000 gpm for four- and three-loop operation, respectively.
The new FAHNSpecifications for four- and three-loop operation are supported by the
~
fuel cycle design and safety analysis. These new Specifications are required by the Westinghouse Standard format. The current Technical Specifications refer to a total FAH value of 1.78 in the safety limit and reactivity control bases. However, 1
there is no current formal Surveillance Requirement. This 1.78 value corresponds to an FAH N value of 1.656 (1.78 divided by the engineering factor of 1.075; see FDSA Section 4.2.3). The proposed full-power values of 1.60 (four-loop) and 1.64 (three-loop) are bounded by the original design basis value and are therefore more restrictive. The proposed limits include an allowance for a peaking increase with reduced power (increased rod insertion). The Applicability, Action, and Surveillance Requirements '
provided are based on the Westinghouse Standard format.
3.18 POWER DISTRIBUTION MONITORING AND CONTROL The curreat Specification provides the axial offset limits, control rod insertion limits, and Surveillance Requirements for measuring the core power distribution to assure operation within the LHGRs. This Specification will be deleted, but the requirements are being transferred to the revised Specifications 3.10 and 3.17.
The power distribution measurement requirements using the movable in-core detector system will be included in the Surveillance Requirements to the LHGR Limiting Conditions for Operation (LCO) in the revised Specification 3.17. The two-thimble ;
method of LHGR surveillance will be deleted since it has never been used. The axial offset limits will now be specified as LCOs for four- and three-loop operation. The axial offset monitoring requirements will be included as Surveillance Requirements to the axial offset LCOs in the revised Specification 3.17.
The control rod insertion limit will be specified in the revised Specification 3.10. The low-power physics test exception will be provided in the new Specification 3.24. The current requirement to determine the monthly power weighted average Bank B position will be deleted as a formal Technical Specification Surveillance Requirement since it is not required by the Westinghouse Standard format. The survelibnce, however, will continue to be performed as routine core follow. The design basis value of 280 steps has been increased to 288 steps due to the revised physics reload methodology.
3.20 REACTOR COOLANT SYSTEM FLOW, TEMPERATURE, AND PRESSURE The current Specification requirements of flow, temperature, and pressure will be included in the revised Specification 3.17. Therefore, current Specification 3.20 will be deleted.
3.24 SPECIAL TEST EXCEPTIONS This will be a new Technical Specification that formalizes exceptions to requirements in order to perform various low-power start-up physics tests. The following current Technical Specifications identify test exceptions:
7
- j. o 3.10.A and 3.18.C.1 - Power-Dependent Control Rod Insertion Limits o 3.10.D. - Shutdown Margin in MODES 3, 4, and 5 and the Subcritical Portion'of Mode 2 The new Specification includes formal test exceptions for shutdown margin, moderator temperature coefficient, minimum temperature for criticality,' control rod alignment, control rod insertion (shutdown and control banks) and control rod position indication.
J ( This new specification is consistent with the Westinghouse Standard format and i
4 provides formal Applicability, Action, and Surveillance Requirements.
i t.
I 4.9 MAIN STEAM ISOLATION VALVES i
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.
l The' revised Specification clarifies the closure time requirement by specifying .
simultaneous closure of all four valves within 10 seconds. The Basis will be revised to 4
t
, delete an obsolete reference to the moderator temperature coefficient assumed in the steam-line break analysis. The reference to a prevention of a return to critical will i
also be deleted since the prevention of a return to critical is no longer an acceptance criterion for the steam-line break accident.
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