ML20245E592

Application for Amend to License DPR-61,changing Tech Specs to Support Cycle 16 Operation

ML20245E592
Person / Time
Site:	Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date:	07/28/1989
From:	Mroczka E CONNECTICUT YANKEE ATOMIC POWER CO., NORTHEAST UTILITIES
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20245E598	List: B13292, Proposed Tech Specs,Supporting Cycle 16 ML20245E592
References
B13292, GL-88-16, TAC-48019, TAC-66797, NUDOCS 8908110273
Download: ML20245E592 (10)

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", HARTFORD, CONNECTICUT o6141-0270 (203) 665-5000 July 28, 1989 Docket No. 50-213 B13292 Re: 10CFR50.90 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington,.DC 20555 Gentlemen:

Haddarn Neck Plant Cycle 16 Reload Proposed Chances to Technical Specifications f Pursuant to 10CFR50.90, Connecticut Yankee Atomic Power Company -(CYAPC0) hereby proposes to amend Operating License DPR-61 by incorporating the attached changes into the Technical Specifications for the Haddam Neck Plant.

These proposed changes are being submitted to support operation of the Haddam Neck Plant for Cycle 16. A description of the proposed changes is provided in Attachnent 1. The revised pages are provided in Attachment 2. These proposed changes are being nrovided in Standcrd Technical Specification (STS) format to support the conversion of the Haddam Neck Plant Technical Specifications to STS format. " Change bars" in the margin indicate changes related to this request on STS format pes previously October 26, subnigdMarchto the6,NRC Stag)in letters datedggust 29, 1988 June 2,1989,(5) June 23,1989,g8, 198 April 21, 1989, and July 28,1989.g)

(1) E. J. Mroczka letter to the U.S. Nuclear Regulatory Commission, "Haddam Neck Plant, Proposed Revision to Technical Specifications, Electrical Power Systems (TAC No. 66797)," dated August 29, 1988.

(2) E. J. Mroczka letter to the U.S. Nuclear Regulatory Commission, "Haddam Neck Plant, Amendment Request for Sections 1.0, 3/4.2, 3/4.9, 3/4.10, 3/4.11, 5.0 and 6.0 of the Revised Technical Specifications," dated October 26, 1988.

(3) E. J. Mroczka letter to the U.S. Nuclear Regulatory Commission, "Haddam Neck Plant, Amendment Request for Section 3/4.1 of the Revised Technical Specifications (TAC No. 48019)," dated March 6, 1989.

i (4) E. J. Mroczka letter to the U.S. Nuclear Regulatory Commission, "Haddam Neck Plant, Proposed Revision to Technical Specifications Section 3.6, Emergency Core Cooling System (ECCS)," dated April 21, 1989.

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i U.S. Nuclear Regulatory Conmission j B13292/Page 2 1 July 28, 1989  !

Backaround I

1. Generic Letter 88-16 In a letter dated October 4,1988,(8) the NRC Staff issued Generic letter 88-16, " Removal of Cycle-Specific Parameter Limits from Technical Speci-fications." This generic letter proposes that cycle specific parameters be removed from the technical specifications and be transferred to an

" operating limits report" that provides unit-specific operating limits for a particular operating cycle. The generic letter provided specific guidance as to how these changes could be implemented. The changes proposed by CYAPC0 herein are consistent with the guidance in the generic I letter. Specifically:

1) tt 's license auendment request proposes the addition of a definition k of a formal report that includes the values of cycle-specific parameter limits that are established using NRC-approved methodology and consistent with all applicable limits of the safety analysis.

P,r the Haddam Neck Plant, this report is called the " Technical

) Report Supporting Cycle Operation (TRSCO)" and is defined in pro-posed Technical Specification 1.36. This report has been routinely submitted to the NRC Staff with the reload license amendmen'. request to support operation for the next cycle. The references listed in this report provide the NRC Staff with the approved calculated methodologies.

2) This license amendment request provides an administrative reporting i requirement to submit the TRSCO to the NRC Staff. The Cycle 16 l TRSCO is provided as Attachment 3 to this letter. The reporting  !

requirement is provided as proposed Technical Specification 6.9.1.9. l I

(Footnote Continued)

(5) E. J. Mroczka letter to the U.S. Nuclear Regulatory Commission, "Haddam Neck Plant, Proposed Revision to Technical Specifications 3/4.4, 3/4.6, and 3/4.7 of the Revised Technical Specifications," dated June 2,1989.

(6) E. J. Hroczka letter to the U.S. Nuclear Regulatory Commission, "Haddam Neck Plant, Proposed Revision to Technical Specifications 2.0, 3/4.3 of the Revised Technical Specifications," dated June 23, 1989.

(7) E. J. Mroczka letter to the U.S. Nuclear Regulatory Comission, "Haddam Neck Plant, Proposed Technical Specifications, Sections 3.0/4.0 and 3.4.6.2 of the Revised Technical Specifications," dated July 28, 1989.

(8) D. M. Crutchfield letter to All Power Reactor Licensees and Applicants,

" Removal cf Cycle-Specific Parameter Limits from Technical Specifications (Generic Letter 88-16)." dated October 4, 1988.

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., a U.S. Nuclear Regulatory Commission B13292/Page 3 July 28, 1989

3) Proposed Technical Specification changes have been provided to specify that cycle specific parameters will be maintained within the limits provided in the TRSCO.

The cycle specific parameter limits that are being removed from the technical specifications are:

1) Moderator temperature coefficient

2) Control rod insertion limits

3) Axial offset alarm limits

4) Linear heat generation rate (LHGR) limits

5) Nuclear enthalpy rise hot channel factor limits The above limits have changed over the last several reloads. The axial offset alarm and LHGR limits are being revised for Cycle 16. With the approach recommended by the generic letter it will no longer be necessary for CYAPC0 and the NRC Staff to process a reload license amendment for the sole purpose of updating cycle-specific parameters. However, since

) the proposed technical specifications require submittal of the TRSCO to the NRC Staff, Staff review of the cycle-specific operating limits will continue.

II. MJji Offset /LHGR/DNB Parameg n In addition, this submittal involves proposed changes to Technical Specification Sections 3.2.1.1, 3.2.1.2, and 3.2.5 to provide clarifica-tions to existing surveillance requirements. These clarifications are necessary to assure compliance with surveillance requirements during plant start-up following refueling. e These changes to the technical specifications are proposed to clarify the surveillance requirements during plant start-up following a refueling outage. The applicability statement of Technical Specifications 3.2.1.1 and 3.2.1.2 requires monitoring the axial offset when operating above 40%

of rated power. However, proposed Surveillance Requirement 4.2.1.1.1 requires the excore/incore axial offset correlation to be determined and implemented upon reaching 80% power (50% for three loop operation) because this correlation cannot be accurately performed until a minimum of three days operation at 80% power (50% for three loop operation) after start-up. The normal operation surveillance requirement specifies continuous monitoring using the correlation above 40% power. The requirement of not exceeding 80% power (50% for three loop operation),

combined with the linear heat generation rate (LHGR) surveil 16nces of Specification 4.2.2.1.1 or 4.2.2.2.1 provide assurance that the LHGR will not exceed the initial conditions assumed for the loss of coolant acci-dent (LOCA) analyses prior to determining the correlation and performance of continuous axial offset monitoring. The proposed surveillance requirements acknowledge the need to perform the surveillance after the ,

correlation is implemented. j

! , a U.S. Nuclear Regulatory Commission B13292/Page 4 July 28,1989 The present Surveillance Requirement 4.2.5 requires verification of the reactor coolant system (RCS) total flowrate once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when operating in Mode 1. This surveillance cannot accurately be performed untjl achieving 100% power which allows the RCS flow test to be completed and the new flow coefficients to be implemented. The proposed change to i the Surveillance Requirement 4.2.5.1 transfers the requirement to verify RCS flow rate once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to Technical Specification 4.2.5.2, which now includes the requirement to verify RCS flow rate once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> following the determination using a heat balance within 7 effective full power days (EFPD) of achieving 100% power. The existing requirement to determine the RCS flowrate using a heat balance within 7 EFPDs of achiev-ing 100% power minimizes the time at power prior to determining the flow ,

coefficients for routine surveillance.

III. Control Rod Insertion Limits The proposed change to Technical Specification 3.1.3.5 redefines the fully withdrawn position to be 317 steps instead of 320 steps. The i revised fully withdrawn position has been defined based on the nominal ,

fuel rod / control rod interface. This change will allow greater opera-tional flexibility in the positioning of control rods to minimize future control rod wear concerns and provide additional margin to accommodate drift in the individual rod position indicators. The nuclear design analyses adjust the cold dimensions to hot conditions and accordingly model the control rod as fully withdrawn at 317 steps, and is consistent with the trip reactivity assumptions used in the safety analysis.

IV. RCS Heatuo/RCS Hydrostatic and leak Testino Related Chances Technical Specification 3.4.1.4.1 requires that at least one RHR loop be in operation in mode 5. This requirement severely limits the ability to heat up the reactor coolant system (RCS). The proposed change allows the RHR pump to be deenergized during heatup as long as 2 RCPs are operating and other conditions are met.

One of the recommendations which resulted from analysis of the thermal shield repair was that no more than two reactor coolant pumps be operated j at temperatures less than 350*F. Recent experience has demonstrated that the RCS heatup is very slow with two reactor coolant pumps and one RHR pump operating. The proposed change will not impact plant safety as heat removal capability will be maintained, since the pump will only be ,

deenergized, the other RHR loop will be operable .and reactor coolant 1 pumps are operating in at least two unisolated l' oops, with steam genera-tor secondary side water level greater than 25 percent. This will allow heatup to mode 4 to be achieved in a safe and timely manner.

1 The proposed changes to Technical Specification 3.4.9.1 allow the low l temperature overpressure protection system (LTOPS) to be isolated during )

performance of RCS hydrostatic and leak testing. Figure 3.4-3, providing i limits for hydrostatic and leak testing, specifies that heatup and l

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, 3 U.S. NucHar Regulatory Commission B13292/P&9e 5 July 28, 1989 {

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l cooldown rates are limited to 10 degrees F per hour for one hour prior to and during testing. With this limit, the LTOPS is ut required to protect the vessel and other RCS components during hydrostatic and leak j testing. The proposed change to Technical Specification 3.4.3.3 (LTOPS) again specifies that the LTOPS can be out of service during the perfor-mance of hydrostatic and leak testing.

l Bases Section 3/4.4.1 has been modified to reflect the fact that an RHR {

loop is not required during RCS heatup given certain conditions. 1 Significant Hazards Consideration In accordance with 10CFR50.92, CYAPC0 has reviewed the attached proposed 4 changes and has concluded that they do not involve a significant hazards '

consideration. The basis for this conclusion is that the three criteria of 10CFR50.92(c) are not compromised. The proposed changes do not involve a significant hazards consideration because the change would not:

1. Involve a significant inerene in the probability or consequences of  !

an accident orevicusly evaluated.

1. Generic letter 98-16 Related Chances t

There are no design basis accidents impacted by the format  !

change to relocate the cycle-specific parameter limits from the technical specifications to the TRSCO. The Cycle 16 parameter l limits are provided in the Core Operating Limits Section of the {

TRSCO (see Attachment 3). The Cycle 16 reload has affected some of the core physics parameters. These parameters were input to the design basis accident and transient analysis. The design basis LOCA and non-LOCA transients were evaluated to  ;

determine what impact resulted from the Cycle 16 reload core.

As discussed in the TRSCO, there is little if any impact on the consequences of any design basis transients. In addition, neither the proposed technical specification changes nor the Cycle 16 reload affect the probability of occurrence of any design basis accidents. Therefore, these proposed changes are concluded to not result in t significant increase in the probability or consequences of any accidents previously analyzed.

II. Arial Offset /LHGR/DNB Paremeters These proposed changes are clarifications of existing surveil-lance requirements. These changes have no impact on the l

operation of the Haddam Neck Plant. The axial offset surveil-i l

lance cannot be accurately performed until a minimum of three '

dtys operation at 80% and the RCS flow rate surveillance cannot be accurately performed until achieving 100% power. Therefore, j the proposed changes ensure that proper and accurate I l l

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4 s U.S. Nuclear Regulatory Commission B13292/Page C July 28 -1989 surveillance tests are performed. The linear heat generation rate (LHGR) surveillance as proposed ensures that the LHGR will not exceed the initial conditions assumed for the LOCA analyses prior to performing the axiel offset surveillance. As such, there is no impact on the probability or consequences of any accident previously evaluated.

III. Control Rod Insertion Limits Changing the 'all rods out" position from 320 steps to 317 steps does not impact the probability or consequences of any design basis accidents. 'The 317 step pos Lion is based on the interface between the fuel assemblies and the control rods.

All the physical models used in the cycle design and determina-tion of safety analysis input parameters assume that the "all rods out" position is 317 steps. No safety systems are affect-ed by this change nor are any design basis events affected.

This proposed change more precisely reflects the physical configuration in the core.

IV. RCS Heatuo/RCS Hydrostatic and Leak Testino The proposed changes allow the operating RHR pump to be deener-gized durmg RCS heatup and allows the LTOPS to be isolated during the performance of RCS hydrostatic or leak testing.

Clarifications to the testing, heatup, and cooldown curves are also included.

The ability to deenergize the operating RHR pump during an RCS heatup in MODE 5 is an operational consideration. Guidelines issued as a result of the thermal shield repair allow a maximum of 2 reactor coolant pumps operating in MODE 5. An RCS heatup with very low decay heat is difficult to accomplish with only 2 RCPs and a minimum RHR ficw with the RHR flow centrol valve fully closed. A footnote has been added to the LCO to allow the operating pump to be de-energized provided 'the following constraints are met:

1. Reactor coolant pumps are operating in at least two unisolated loops, with secondary side narrow range water level greater than 25%.

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2. No operations are permitted that would cause diletion of the reactor coolant system boron concentration.

3. The core outlet temperature is maintained at least 10*F below saturation temperature. .

t These constraints, combined with the current LCO requiring an RHR loop to be operable, provide an adequate heat sink for

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, I U.S. Nuclear Regulatory Commissior.

B13292/Page 7 July 28, 1989 operation in M3DE 5. There are no design basis accidents affected by this proposed change.

Isolating the LTOPS during an P.CS hydrostatic and/or leak test does nnt affect the design basis for a low temperature, over-pressurization event. Performance of a hydrostatic and/or leak test above 245'F and 235'F respectively, and maintaining a heatup rate s 10'F/ hour for one hour prior to and during the tests assure that the 10CFR50 Appendix b margin of safety is maintained.

The balance of the changes are to provide clarifications to current requirements. No design basis accidents are affected by these clarifications.

The proposed technical specification changes do not affect the probability of failure of the RHR or LTOP systems. The LTOPS is not required during the performance of a hydrostatic ind/or leak test prcvided they are performed above 245'F and 235'F respect hely, and a heatup ratt 1 10 F/ hour is maintained for one how. prior to and during the test.

The RHR system would purposely be taken out of service during an RCS heatup in MODE 5 with low decay heat by deenergizing the operating RHR pump. Shutting off the pump does not affect the -

probability of failure of that pump, nor does it affect the probability of failure of the remaining operable pump.

Overall, these proposed changes do not affect the probability or consequences of any design basis accidents nor do the changes increase the probability of a failure of a safety system or degrade the performance of a safety system below that assumed in the design basis analysis.

2. Create then' ossibility of a new or different kind of acqident from any previo'>s.ly evaluated I. Generic letter 88-16 Related Chancel There are no failure modes associated with the proposed techni-cal specification changes on the Cycle 16 reload. A review of

\ the affected non-LOCA and LOCA transients has demonstrated that 1 the plant response has not been modified to the point where a new accident has been identified. Accordingly, these changes are concluded to not present the possibility for a new, unana-lyzed accident.

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, b U.S. Nuclear' Regulatory Commission 813292/Page 8 July 28, 1989 II. Axial Offset /LHGR/DNB Parameters These changes provide clarifications and a correction of existing technical specifications to ensure that the surveil-lance requirements are effective and perform their intended function. There is no impact on plant operation or response.

Therefore, it is concluded that these proposed changes do not present the possibility for a new unanalyzed accident.

III. Control Rod Insertion limiti The proposed change redefines the "all rods out" position to more precisely reflect the physical relationship between the fuel and control rods. The plant response is not modified by this proposed change nor are there any new failure modes presented. The proposed change does not impact the probability of an accident to the point where it should be considered within the design basis.

IV. RCS Heatuo/RCS Hydrostatic and Leak Testina The plant response due to the proposed changes has not been modified to the point where it can be considered that a new accident has been defined. The hydrostatic and leak tests will continue to be performed above 245'F and 235'F respectively.

Taking the RHR pump out of service in MODE 5 with low decay heat will allow a normal RCS heatup.

The failure mode of a low temperature, overpressurization event occurring below 315'F while the LTOPS is isolated, has alreidy been analyzed. Limiting the heatup rate to less than or equal to 10 degrees F while the LTOPS is out of service addresses this potential. Therefore, these proposed changes do not create the potential for a new unanalyzed accident. There are no failure modes associated with taking the operating RHR pump out of service during an RCS heatup since the performance of  !

the RHR systen is not affected.

3. Involve a significant reduction in a marain of safety  ;

I. Generic letter 88-16 Related Chanaes The proposed changes to the technical specifications and the Cycle 16 reload have been evaluated for their impact on non-LOCA and LOCA design basis events. Since as previously stated there is no impact on the consequences of the design basis events, therefore, it follows that there is no impact on the protective boundaries. There are no failure modes associated with the proposed changes or the Cycle 16 reload. Therefore, -

there is no impact on the margin of safety.

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L U.S. Nuclear Re'gulatory Commission B13292/Page 9 July 28, 1989-II. 'Arial Offset /LHGR/DNB Par maters

. These proposed changes 'do . not ' involve any ' failure modes or

. changes in . plant ' operation or transient . response. These

. changes are proposed to better ensure.that operating limits are -

' maintained. Therefore, there is no impact ~ on the- margin of safety.

III. Control Rod Insertion Limits This' proposed change has no impact on t'he protective boundaries of the plant. There are no failure modes assoc ^ated with this change an.d there is no affect on the safety limits. This-proposed change simply reflects -the physical c:mfiguration in -

the core more precisely.

IV. RCS Heatun/RCS Hydrostatic and Leak Testina a

The proposed changes do not impact. the protective boundaries.

Performance of a hydrostatic and/or leak test above 245'F or 235'F respectively while maintaining a heatup rate i 10*F/ hour one hour prior to and during the test assures that the margin of safety required by 10CFR50 Appendix G is maintained.

Deenergizing the operating RHR pump in MODE 5' with low decay heat will result in a controlled RCS heatup without' affecting the protective boundaries.

~The Commission has provided guidance concerning the application of the stan-dards in 10CFR50.92 by providing certain examples (57 FR 7751, March 6, 1086) of amendments that are coasidered 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. ,

I This assumes that no significant changes are made to the acceptance criteria '

for the technical specifications, that the analytical methods used to demon-strate conformance with the technical specifications and regulations are not significantly changed and that the NRC Staff has previously found such methods acceptable. The attached proposed changes for Cycle 16 are similar to example j j

(iii) in that they involve changes resulting from a reactor core reloading.

The fuel assemblies are identical in design to those found previously accept- l able to the'NRC Staff for Cycle 15.

The proposed changes to Sections 3.2.1.1, 3.2.1.2, and 3.2.5 are not enveloped by a specific example. However, as statet earlier, th9se proposed changes ensure that surveillance requirements are paaformed at the most effective time and that protective limits are in place until the surveillance tests are performed. As described ebove, these proposed changes clearly do not consti-tute a significant hazards consideration.

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U.S. Nuclear Regulatary Commission -

B13292/Page 10 July 28,1989 Based upon the information contained in this submittal and the environmental assessment for the Haddam Neck Plant, there are no significant radiological or nonradiological impacts associated with the proposed action, and the proposed license amendment will not have a significant effect on the quality of the human environment.

The Haddam Neck Plant Nuclear Review Board has reviewed and approved the attached proposed revisions and concurs with the above determinations.

In accordance with 10CFR50.91(b), CYAPC0 is providing the State of Connecticut with a copy of this amendment.

CYAPC0 respectfully requests that this license amendment request be reviewed and issued by October 6,1989 ta support restart of the Haddam Neck Plant following the Cycle 16 refueling outga l

Very truly yours, CONNECTICUT YANKEE ATOMIC POWER COMPANY t f?n/

E. J (zka' ~ f/

Senio ice President cc: W. T. Russell, Region I Administrator A. B. Wang, NRC Project Manager, Haddam Neck Plant J. T. Shedlosky, Senior Resident Inspector, Haddam Neck Plant Mr. 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. J. 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 Licensee herein, and that the statements contained in said information are true and correct to the best of his knowledge and belief.

Mhtbni0- Y MotaryPuby MyCommission Ex$es Mad 3W3 1

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