L-14-137, License Amendment Request for Adoption of TSTF-535, Revision 0, Revise Shutdown Margin Definition to Address Advanced Fuel Designs

License Amendment Request for Adoption of TSTF-535, Revision 0, Revise Shutdown Margin Definition to Address Advanced Fuel Designs

ML14255A150
Person / Time
Site:	Perry
Issue date:	09/12/2014
From:	Harkness E FirstEnergy Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-14-137
Download: ML14255A150 (14)

Text

FENOC' Perry Nuclear Power Plant P.O. Box97 10 Center Road RrstEnergy Nuclear Operating Company Perry. Ohio 44081 Ernest J. Hsrlal8SS 440-280-5382 Vice President Fax: 440-280-8029 September 12, 2014 l-14-137 10 CFR 50.90 ATIN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

SUBJECT:

Perry Nuclear Power Plant, Unit No. 1 Docket No. 50-440, license No. NPF-58 license Amendment Request for Adoption of TSTF-535. Revision 0. "Revise Shutdown Margin Definition to Address Advanced Fuel Designs" Pursuant to 10 CFR 50.90, FirstEnergy Nuclear Operating Company (FENOC) hereby submits an amendment application for the Perry Nuclear Power Plant, Unit No. 1. The

• proposed amendment modifies the technical specification definition of "Shutdown Margin" (SOM) to require calculation of the SOM at a reactor moderator temperature of 68 degrees Fahrenheit (°F) or a higher temperature that represents the most reactive state throughout the operating cycle. This change. is needed to address new boiling water reactor fuel designs which may be more reactive at shutdown temperatures above 68°F.

To allow for normal NRC processing, FENOC requests approval of the proposed license amendment by September 15, 2015 with the amendment being implemented within 90 days.

There are no regulatory commitments contained in this submittal. If there are any questions or if additional information is required, please contact Mr. Thomas A. Lentz, Manager - Fleet licensing, at (330) 315-6810.

I declare under penalty of perjury that the foregoing is true and correct. Executed on September l'Z. , 2014.

Sincerel~y

~

c::::-.. t Ernest J. Harkness

Perry Nuclear Power Plant, Unit No. 1 L-14-137 Page 2

Enclosure:

Evaluation of Proposed License Amendment.

cc: NRC Region Ill Administrator NRC Resident Inspector NRC Project Manager State of Ohio (NRC Liaison)

Utility Radiological Safety Board

Enclosure L-14-137 Evaluation of Proposed License Amendment (11 Pages Follow}

EVALUATION OF PROPOSED LICENSE AMENDMENT

Subject:

Proposed Revision of Technical Specification 1.1 Definition, "Shutdown Margin (SOM)"

1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 4.2 Significant Hazards Consideration Analysis 4.3 Conclusions

6.0 REFERENCES

Attachments

1. Proposed changes to Technical Specifications (MARK-UP)

2. Proposed changes to Technical Specifications (RETYPED)

Perry Nuclear Power Plant Evaluation of Proposed License Amendment Page 2 of 7 1.0

SUMMARY

DESCRIPTION The proposed amendment would modify FirstEnergy Nuclear Operating Company's (FENOC) Perry Nuclear Power Plant, Unit No. 1 (PNPP) Technical Specifications (TS).

The proposed amendment modifies the TS definition of "Shutdown Margin" (SOM) to require calculation of the SOM at a reactor moderator temperature of 68°F or a higher temperature that represents the most reactive state throughout the operating cycle. This change is needed to address new Boiling Water Reactor (BWR) fuel designs, which may be more reactive at shutdown temperatures above 68°F.

The proposed changes are consistent with the Nuclear Regulatory Commission (NRC) approved Technical Specification Task Force (TSTF) Traveler TSTF-535, Revision 0.

The Federal Register notice published on February 26, 2013 (78 FR 13100) as part of the consolidated line item improvement process announced the availability of this TS improvement.

The TS 1.1 definition that would be revised by the proposed amendment is:

• "Shutdown Margin (SOM)"

The affected page of the current TS, annotated to show the proposed changes, is provided in Attachment 1. The re-typed TS page is provided in Attachment 2.

2.0 DETAILED DESCRIPTION Shutdown Margin (SOM) requirements are specified to ensure:

a. The reactor can be made subcritical from all operating conditions, transients, and Design Basis Events,

b. The reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and

c. The reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.

NUREG-1434 Revision 4.0, Standard Technical Specifications, General Electric BWR/6 Plants, Section 1.1, "Definitions," defines SOM.

SOM shall be the amount of reactivity by which the reactor is subcritical or would be subcrltical assuming that:

a. The reactor is xenon free,

b. The moderator temperature is 68°F, and

Perry Nuclear Power Plant Evaluation of Proposed License Amendment Page 3of7

c. All control rods are fully inserted except for the single control rod of highest reactivity worth, which is assumed to be fully withdrawn. With control rods not capable of being fully inserted, the reactivity worth of these control rods must be accounted for in the determination of SOM.

In a letter from Global Nuclear Fuel - Americas, LLC (GNF) to the U.S. Nuclear Regulatory Commission dated November 8, 2010 (Reference 1), GNF stated the Technical Specification definition of 68°F moderator temperature for the shutdown margin evaluation may not be the most reactive condition for some fuel designs. For fuel products through GE14, the maximum reactivity condition for SOM always occurs at a moderator temperature of 68°F, and the SOM is calculated at this temperature.

For cores with GNF2 fuel or other modem designs, it is possible that the most reactive moderator temperature may occur at a temperature above 68°F. For normal reload core designs, even those with 100% GNF2 fuel, it is expected that the maximum reactivity condition at beginning-of-cycle (BOC) will remain at 68°F. However, later in the cycle there is the possibility of a more limiting SOM at a temperature greater than 68°F.

The proposed change revises the definition of SOM to address this situation by

• changing the introductory sentence and item b of the definition to require calculating SOM at 68°F or a higher temperature corresponding to the most reactive state throughout the operating cycle. The revised definition will state:

SOM shall be the amount of reactivity by which the reactor is subcritical or would be subcritical throughout the operating cycle assuming that:

a. The reactor is xenon free;

b. The moderator temperature is ?. 68°F I corresponding to the most reactive state; and

c. All control rods are fully inserted except for the single control rod of highest reactivity worth, which is assumed to be fully withdrawn. With control rods not capable of being fully Inserted, the reactivity worth of these control rods must be accounted for in the determination of SOM.

(Changes are shown in bold.)

3.0 TECHNICAL EVALUATION

Title 1Oof the Code of Federal Regulations (CFR), Part 50, Appendix A. "General Design Criteria," (GOC), GOC 26, "Reactivity control system redundancy and capability," states:

Two independent reactivity control systems of different design principles shall be provided. One of the systems shall use control rods, preferably including a

Perry Nuclear Power Plant Evaluation of Proposed License Amendment Page 4of7 positive means for inserting the rods, and shall be capable of reliably controlling reactivity changes to assure that under conditions of normal operation, including anticipated operational occurrences, and with appropriate margin for malfunctions such as stuck rods, specified acceptable fuel design limits are not exceeded. The second reactivity control system shall be capable of reliably controlling the rate of reactivity changes resulting from planned, normal power changes (including xenon burnout) to assure acceptable fuel design limits are not exceeded. One of the systems shall be capable of holdjng the reactor core subcritjcal ynder cold conditions. (Emphasis added).

In BWR plants, the control rods are used to hold the reactor core subcritical under cold conditions. The control rod reactivity worth must be sufficient to ensure the core is subcritical by the amount of the SOM. The SOM is the additional amount of negative reactivity needed to offset the reactivity worth of changes in moderator and fuel temperature, the decay of fission product poisons, failure of a control rod to insert, and reactivity insertion accidents.

For cores licensed with GNF methods, the licensing basis requirements for SOM are specified in GESTAR II (NEDE-24011-P-A-19) (Reference 2), Section 3.2.4.1, 11 Shutdown Reactivity," which states:

The core must be capable of being made subcritical, with margin, in the most reactive condition throughout the operating cycle with the most reactive control rod fully withdrawn and all other rods fully inserted.

The GESTAR II requirement is consistent with the NRC Standard Review Plan (NUREG-0800) Chapter 4.3, "Nuclear Design," which states:

The adequacy of the control systems to assure that the reactor can be returned to and maintained in the cold shutdown condition at any time during operation. The applicant shall discuss shutdown margins (SOM). Shutdown margins need to be demonstrated by the applicant throughout the fuel cycle.

While the Standard Review Plan does not precisely prescribe that the temperature of minimum shutdown margin be determined, the requirement of shutting down the reactor and maintaining it in a shutdown condition suggests that considering a range of thermal and exposure conditions is appropriate in the determination of the minimum SOM.

For historical fuel products through GE14, the maximum reactivity condition for SOM always occurs at a moderator temperature of 68°F (the minimum expected reactor moderator temperature), and the SOM is calculated at this temperature. These fuel products are designed so that the core Is always under moderated when all control rods are inserted except for the single most reactive rod. In this under moderated condition, higher coolant temperatures result in a lower water density and less moderation. Therefore, specifying use of a minimum coolant temperature of 68°F in the SOM calculation results in the most limiting SOM value.

Perry Nuclear Power Plant Evaluation of Proposed License Amendment Page 5of7 In cores with GNF2 fuel, or other modern fuel designs with increased moderation. it is possible for the most reactive condition to exist at a moderator temperature greater than 68°F. For normal reload core designs, even those with all GNF2 fuel, it is expected that the maximum reactivity condition at BOC will remain at 68°F. The strong local absorption effects of gadolinia in fresh fuel make the core under moderated. Later in the cycle, as gadolinia is depleted, all cores become less under moderated, and there is the possibility that the maximum reactivity condition Is at a temperature greater than 68°F. Thus, late in the fuel cycle the most limiting SOM may occur at a temperature greater than 68°F.

The proposed change to the definition ensures that SOM Is calculated using the appropriate limiting conditions for all fuel types.

4.0 REGULATORY EVALUATION

FENOC has reviewed the model safety evaluation dated February 26, 2013

{78 FR 13100} as part of the Federal Register Notice of Availability. This review Included a review of the NRC staff's evaluation, as well as the information provided in TSTF-535. As described in the subsequent paragraphs, FENOC has concluded that the justifications presented in the TSTF-535 proposal and the model safety evaluation prepared by the NRC staff are applicable to PNPP and justify this amendment for the incorporation of the changes to the PNPP Technical Specifications {TS}.

4.1 Applicable Regulatory Requirements/Criteria GOC 26, "Reactivity control system redundancy and capability," states that one of the reactivity control systems shall be capable of holding the reactor core subcritical under cold conditions.

NUREG-0800, "Standard Review Plan," states that the reactivity control systems must be capable of maintaining the reactor in the cold shutdown condition at any time during operation.

The proposed change revises the definition of SOM to ensure that these regulatory criteria are met for all fuel types at any time in core life.

4.2 Significant Hazards Consideration Analysis The proposed change revises the definition of Shutdown Margin {SOM} to require calculation of the SOM at a reactor moderator temperature of 68°F or a higher temperature that represents the most reactive state throughout the operating cycle.

This change is needed to address new fuel designs, which may be more reactive at shutdown temperatures above 68°F.

FENOC has evaluated whether or not a significant hazards consideration is involved

Perry Nuclear Power Plant Evaluation of Proposed License Amendment Page 6of7 with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, Issuance of Amendment:

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No The proposed change revises the definition of Shutdown Margin (SOM). SOM is not an initiator to any accident previously evaluated. Accordingly, the proposed change to the definition of SOM has no effect on the probability of any accident previously evaluated. SOM is an assumption in the analysis of some previously evaluated accidents and inadequate SOM could lead to an increase in consequences for those accidents. However, the proposed change revises the SOM definition to ensure that the correct SOM is determined for all BWR fuel types at all times during the fuel cycle.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed amendment create the possibility of a new or different kind of accident from any previously evaluated?

Response: No The proposed change revises the definition of SOM. The change does not involve a physical alteration of the plant that is, no new or different type of equipment will be installed or a change in the methods governing normal plant operations. The change does not alter assumptions made in the safety analysis regarding SOM.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No The proposed change revises the definition of SOM. The proposed change does not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined. The proposed change ensures that the SOM assumed in determining safety limits, limiting safety system settings or limiting conditions for operation is correct for all BWR fuel types at all times during the fuel cycle.

Perry Nuclear Power Plant Evaluation of Proposed License Amendment Page 7of7 Therefore, the proposed change does not involve a significant reduction in a margin of safety.

4.3 Conclusions The proposed changes are intended and structured to maintain compliance with the applicable regulatory requirements and criteria identified in section 4.2 and with the guidance provided In NRC-approved TSTF-535, Revision 0.

In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

5.0 ENVIRONMENTAL CONSIDERATION

Evaluation of the proposed change has determined that the change does not involve (I) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in the individual or cumulative occupational radiation exposure. Accordingly, the proposed change meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), an environmental assessment of the proposed change is not required.

6.0 REFERENCES

1. Letter from A. Lingenfelter (Global Nuclear Fuel -Americas, LLC) to U.S.

Nuclear Regulatory Commission, Temperature Dependent Strong-Rod-Out Cold Shutdown Margin," dated November 8, 2010.

2. NEDE-24011-P-A-19, "General Electric Standard Application for Reactor Fuel (GESTAR II)," May 2012.

Attachment 1 PROPOSED CHANGES TO TECHNICAL SPECIFICATIONS CHANGES (MARK-UP)

(1 Page Follows)

Def!Diticma 1.1 1.1 Definitiou * (continued)

SllUTDOWH IWlGIH (SDI) SDI allall be the UOUDt of reactivity by which the reactor is subcdtical or would be sUbcritical throughout the OReratig cxsle assuming tbat r

a. fte reactor is xeaoa free1

b. fte moderator temperature is ~ 68°P, correspondt ll9 to the moat reactive state1 and

c. All control rods are fully iuerted except for the single control rod of highest reactivity worth, which ia assumed to be fully withdrawn.

With control roda not capable of being fully 1Daerted, the reactivity worth of these control rods must be accounted for in the determination of SDH.

STASGIRID UST BASIS A BTAGGBRBD UST BASIS shall couiat of the teatiq of one of the ayatema, aubayatema, cb.annels, or other designated c~nts during the i1terval specified by the Surveillance Prequeac:y, so tbat all ayatema, subsystems, chaliaela, or other designated COIDJODenta are teated during n Surveillance rrequeac:y intervals, where n is tie total number of *teu, aubayatllDI, chamlela, or other designated compoaeats iD the associated function.

DBRllAL PODI '1'llBRllAL POUR shall be the total reactor core beat tranafer rate to the reactor coolant.

TURBID mus SYSTBll The '!1JIBIBI mus SYSTBll USPOHSB !DIB couists RBSPORBB TDIB of two componenta 1

a. !be time from initial movement of the main turbine atop valve or control valve until 80' of the turbine bypass capacity is eatabli1bed1 ad

b. !be ti* from initial movement of the main turbine atop valve or control valve until ilitial IDVeUDt of the turbine bypass valve.

The na~e time *Y be measured by means of any aeries of sequential, overlapping, or total steps ao that the entire reaponae tlae is 'measured.

HUY

• UII! 1 1.0*6 .Amendment Ro. '9

Attachment 2 PROPOSED CHANGES TO TECHNICAL SPECIFICATIONS CHANGES (RETYPED)

(1 Page Follows)

(RETYPED, FOR INFORMATION ONLY Definitiollll 1.1 1.1 Definltiona *(continued)

SllUTDOWH llARQIH (SDM) SDM aball be the amount of reactivity by which the reactor ia aubcritical or would be aUbcritical throughout the operating cycle a11Ullling thats

a. The reactor ia xeDOD free1

b. The moderator temperature ia ~ 68°r, c:orreapondiag to tbe lllOlt reactive 1tate1 and

c. All control rods are fully inaerted except for the single control rod of higheat reactivity wrth, whleh la assumed to be fully withdrawn.

With control rods not capable of being fully

laaerted, the reactivity worth of these control rods must be accounted for ln the determination of SDM.

STAQQBRBD TBS! BASIS A STAGGBRBD TBS! BASIS shall conalat of the

==of one of the ~staa, aul>ayatems, a, or other 4eaipate4 components during the iDtenal ap_ecifled Dy the Burveillance rr~, ao that all ayataa, aubayatema, .

• cluuinela, or other designated components are teated during n surveillance Prequency intervals, where 11 la the total number of ayat8118, subayatema, cbaanela, or other d11ignated eompODeDts in the associated function.

nBRIAL POWBR shall be the total reactor core heat traufer rate to the reactor coolant, TURBID BYPASS SYS'ml The mun mus BYBTBll mtOHSB Tm conaiats RBSPOHSB TDIB of two components 1

a. The time from initial movaent of the main turbine atop valve or control valve until 8°'

of the turbine bypaaa capacity is eatablished1 ud .

b. The time from initial movement of the main turbine stop valve or control valve until initial movement of the turbine bypaH valve.

!be rea~e time may be measured by meana of ay aeries of sequential, overlapping, or total ateps so tbat the entire response tlme ia measured. .

HRRY

• URI! 1 1.0*6 Amendment Ho.