NL-10-0794, License Amendment Request for Incorporation of Previously NRC Approved Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler TSTF-5-A, Rev. 1

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License Amendment Request for Incorporation of Previously NRC Approved Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler TSTF-5-A, Rev. 1
ML103550234
Person / Time
Site: Hatch, Vogtle  Southern Nuclear icon.png
Issue date: 12/16/2010
From: Ajluni M
Southern Nuclear Operating Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-10-0794, TSTF-5-A, Rev 1
Download: ML103550234 (33)
v • d • e


Text

Mark J. Ajiluni, P.E. Southern Nuclear Nuclear Licensing Director Operating Company, Inc.

40 Inverness Center Parkway Post Office Box 1295 Birmingham, Alabama 35201 Tel 205.992.7673 Fax 205.992.7885 December 16, 2010 SOUTHERN Nr COMPANY Docket Nos.: 50-321 50-424 50-366 50-425 NL-10-0794 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Edwin I. Hatch Nuclear Plant - Units 1 & 2 Vogtle Electric Generating Plant - Units 1 & 2 License Amendment Request for Incorporation of Previously NRC Approved Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler TSTF-5-A, Rev. 1, "Delete Safety Limit Violation Notification Requirements" Ladies and Gentlemen:

In accordance with the provisions of 10 CFR 50.90 of Title 10 of the Code of Federal Regulations (10 CFR), Southern Nuclear Operating Company (SNC) is submitting a request for an amendment to the Technical Specifications (TS) for Edwin I. Hatch Nuclear Plant (HNP) and Vogtle Electric Generating Plant (VEGP). The proposed amendments affect Section 2.0 "Safety Limits (SLs)."

This amendment request proposes to delete requirements from the Technical Specifications that duplicate requirements found in the regulations (10 CFR 50.36).

The proposed changes to TS 2.0, "Safety Limits (SLs)" are consistent with Nuclear Regulatory Commission (NRC) approved Industry/Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler TSTF-5-A, Rev. 1, "Delete Safety Limit Violation Notification Requirements."

In addition to the strict implementation of TSTF-5-A, Rev. 1, SNC proposes to delete the first two paragraphs of TS Bases Section B 2.1.2 "Safety Limit Violations" and to delete labels 2.2.2 and 2.2.2.1 of this section. These changes are being made for consistency with the Improved Standard Technical Specifications and clarity of the TSTF-5-A, Rev. 1 implementation.

SNC requests approval of the proposed license amendments by September 24, 2011. Once approved, the amendment would be implemented within 90 days of issuance of the amendment.

Enclosure 1 provides the basis for the proposed changes. Enclosure 2 contains TS markup pages. Enclosure 3 provides clean-typed TS pages. Enclosure 4 includes TS Bases markups for reference only.

U. S. Nuclear Regulatory Commission Log: NL-10-0794 Page 2 SNC has evaluated this request under the standards set forth in 10 CFR 50.92(c) and determined that a finding of "no significant hazards consideration" is justified.

Mr. M. J. AjIuni states he is Nuclear Licensing Director of Southern Nuclear Operating Company, is authorized to execute this oath on behalf of Southern Nuclear Operating Company and to the best of his knowledge and belief, the facts set forth in this letter are true.

This letter contains no NRC commitments.

Respectfully submitted, M. J. Ajluni Nuclear Licensing Director MJA/GAL/emm Sworn to and subscribedbefore me thisJjj day of 0) e C...,8e* - , 2010.

Notary Public My commission expires: 149/.

Enclosures 1. Basis for Proposed Changes

2. Technical Specification Markup Pages
3. Clean Typed Technical Specification Pages
4. Technical Specification Bases Markup Pages cc: Southern Nuclear Operating Company Mr. J. T. Gasser, Executive Vice President Mr. D. R. Madison, Vice President - Hatch Mr. T. E. Tynan, Vice President - Vogtle Ms. P. M. Marino, Vice President - Engineering RType: Hatch=CHA02.004; Vogtle=CVC7000 U. S. Nuclear Regulatory Commission Mr. L. A. Reyes, Regional Administrator Mr. R. E. Martin, NRR Project Manager - Farley, Hatch and Vogtle Mr. P. Boyle, NRR Project Manager Mr. E.D Morris, Senior Resident Inspector - Hatch Mr. M. Cain, Senior Resident Inspector - Vogtle State of Georgia Mr. Allen Barnes, Director - Environmental Protection Division

Edwin I. Hatch Nuclear Plant - Units 1 & 2 Vogtle Electric Generating Plant - Units 1 & 2 License Amendment Request For Incorporation.Of Previously NRC Approved Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler TSTF-5-A, Rev. 1, "Delete Safety Limit Violation Notification Requirements" Enclosure 1 Basis for Proposed Change

Edwin 1. Hatch Nuclear Plant - Units 1 & 2 Vogtle Electric Generating Plant - Units 1 & 2 License Amendment Request For Incorporation Of Previously NRC Approved Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler TSTF-5-A, Rev. 1, "Delete Safety Limit Violation Notification Requirements" Enclosure 1 Basis for Proposed Change Table of Contents 1.0 Summary Description 2.0 Detailed Description 3.0 Technical Evaluation 4.0 Regulatory Evaluation 4.1 No Significant Hazards Consideration 4.2 Applicable Regulatory Requirements/Criteria 4.3 Precedent 5.0 Environmental Consideration 6.0 References

Enclosure 1 Basis for Proposed Change 1.0 Summary Description This amendment request proposes to delete requirements from the Technical Specifications (TS) that are duplicative or contained in other regulations or required to comply with requlations (10 CFR 50.36).

The proposed changes to TS 2.0, "Safety Limits (SLs)" are consistent with Nuclear Regulatory Commission (NRC) approved Industry/Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler TSTF-5-A, Rev. 1, "Delete Safety Limit Violation Notification Requirements."

2.0 Detailed Description The proposed change is to revise the Technical Specifications as follows:

Changes to VEGP Technical Specifications/Technical Specification Bases Affected Technical Specification Change Description 2.2 SL Violations Eliminate Actions 2.2.3, 2.2.4, 2.2.5, and 2.2.6 B 2.1.1 Reactor Core SLs Bases Eliminate discussion of Actions 2.2.3, 2.2.4, 2.2.5, 2.2.6 and the label of 2.2.1 B 2.1.1 Reactor Core SLs Bases Eliminate References 5 and 6 B 2.1.2 RCS Pressure SL Bases Eliminate discussion of Actions 2.2.3, 2.2.4, 2.2.5, 2.2.6, the labels of 2.2.2.1 and 2.2.2.2.

Eliminate fist two paragraphs of "Safety Limit Violations" and labels 2.2.2 and 2.2.2.1.

B 2.1.2 RCS Pressure SL Bases Eliminate References 6 and 7 Changes to HNP Unit 1 Technical Specification/Technical Specification Bases Affected Technical Specification Change Description 2.2 SL Violations Eliminate 2.2.1, 2.2.3, 2.2.4, and 2.2.5.

Change paragraph number from 2.2.2.1 to 2.2.1.

Change paragraph number from 2.2.2.2 to 2.2.2.

B 2.1.1 Reactor Core SLs Violations Eliminate 2.2.1, 2.2.3, 2.2.4, 2.2.5, and the label on 2.2.2.

B 2.1.1 Reactor Core SLs Eliminate References 3 and 5 B 2.1.2 RCS Pressure SLs Violations Eliminate 2.2.1, 2.2.3, 2.2.4, 2.2.5, and the label on 2.2.2 B 2.1.2 RCS Pressure SL Eliminate References 7 and 8 El -2

Enclosure 1 Basis for Proposed Change Changes to HNP Unit 2 Technical Specification/Technical Specification Bases Affected Technical Specification Change Description 2.2 SL Violations Eliminate 2.2.1, 2.2.3, 2.2.4, and 2.2.5.

Change paragraph number from 2.2.2.1 to 2.2.1.

Change paragraph number from 2.2.2.2 to 2.2.2.

B 2.1.1 Reactor Core SLs Violations Eliminate 2.2.1, 2.2.3, 2.2.4, 2.2.5, and the label on 2.2.2.

B 2.1.1 Reactor Core SLs Eliminate References 3 and 5 B 2.1.2 RCS Pressure SLs Violations Eliminate 2.2.1, 2.2.3, 2.2.4, 2.2.5, and the label on 2.2.2 B 2.1.2 RCS Pressure SL Eliminate References 7 and 8 3.0 Technical Evaluation SNC has reviewed the TSTF-5-A, Revision 1, and found it to be applicable as written. The proposed changes are to remove the duplicative requirements to report safety limit violations and requirements to preclude restart after a safety limit violation without NRC approval from the TS. These are considered an administrative action. These reporting and restart requirements are duplicative of what is already contained in the regulations (i.e., 10 CFR 50.36). The reporting requirements in 10 CFR 50.36 require that appropriate prompt notifications are made to the NRC and that Licensee Event Reports (LERs) are submitted to the NRC. 10 CFR 50.36 requires that these reports be performed in accordance with the requirements of 10 CFR 50.72 and 10 CFR 50.73. Therefore, if a TS safety limit is violated, appropriate reporting will be made to the NRC in accordance with the regulations. 10 CFR 50.36 also requires that operations must not be resumed until authorized by the Commission. Removal of duplicative reporting and restart requirements from the TS results in simplification of the TS and Bases and less administrative burden to track duplicative requirements. Adequate administrative controls exist in administrative programs at SNC for the identification and reporting of safety limit violations, and restart restrictions following safety limit violations, in accordance with 10 CFR 50.36, 10 CFR 50.72, and 10 CFR 50.73.

Labels associated with the duplicative TSs listed above are also proposed to be deleted and are reflected in the markups.

4.0 Regulatory Evaluation 4.1 No Significant Hazards Consideration The changes proposed by this license amendment application would revise the Safety Limits Section 2.0 of the VEGP and HNP TS to delete duplicative notification, reporting, and restart requirements from the TS. This proposed change is consistent with the Nuclear Regulatory Commission (NRC) approved Industry/Technical Specification Task Force (TSTF) Standard El -3

Enclosure 1 Basis for Proposed Change Technical Specification Change Traveler TSTF-5-A, Rev. 1, "Delete Safety Limit Violation Notification Requirements." This change facilitates improved content and presentation of Administrative Controls.

Southern Nuclear Operating Company (SNC) has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment', as discussed below:

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

Response: No.

The proposed change to remove the duplicative safety limit reporting, notification, and restart constraint requirements from the TSs does not affect the plant or operation of the plant. The change simply removes duplicative information from the TS that is covered in the NRC regulations. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed change to remove the duplicative safety limit reporting, notification, and restart constraint requirements from the TS does not introduce any new accident scenarios, failure mechanisms, or limiting single failures. All systems, structures, and components previously required for the mitigation of a transient remain capable of fulfilling their intended design functions. The proposed change has no adverse effect on any safety-related system or component and does not challenge the performance or integrity of any safety related system. This change is considered an administrative action to remove duplicative reporting, notification, and restart constraint requirements. Therefore, this proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

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

Response: No.

The proposed changes are administrative and do not involve any reduction in a margin of safety. All systems, structures, and components previously required for the mitigation of a transient remain capable of fulfilling their intended design functions. The proposed change has no adverse effect on any safety-related system or component and does not E1-4

Enclosure 1 Basis for Proposed Change 4.2 Applicable Regulatory Requirements/Criteria The proposed change to remove the notification, reporting, and restart requirements if a safety limit is violated from the TSs simply removes duplicative information from the TSs that is covered in the regulations (10 CFR 50.36). The reporting requirements in 10 CFR 50.36 require that appropriate prompt notifications are made to the NRC and that the Licensee Event Reports (LERs) are submitted to the NRC. 10 CFR 50.36 requires that theses reports be performed in accordance with the requirements of 10 CFR 50.72 and 10 CFR 50.73. Therefore, ifa TS safety limit is violated, appropriate reporting will be made to the NRC in accordance with the regulations. Adequate administrative controls exist in administrative programs at SNC for the identification and necessary reporting of safety limit violations in accordance with 10 CFR 50.36, 10 CFR 50.72, and 10 CFR 50.73. This change is consistent with Nuclear Regulatory Commission (NRC) approved Industry/Technical Specification Task Force (TSTF)

Standard Technical Specification Change Traveler TSTF-5-A, Rev 1, "Delete Safety Limit Violation Notification Requirements."

4.3 Precedent The proposed change to remove the duplicative notification, reporting, and restart requirements if a safety limit is violated from the TSs is consistent with Nuclear Regulatory Commission (NRC) approved Industry/Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler TSTF-5-A, Rev. 1, "Delete Notification, Reporting, and Restart Requirements ifa Safety Limit is Violated." This change is consistent with the license amendment application for Peach Bottom Atomic Power Station, Units 2 and 3. The NRC approved this license amendment request by letter dated May 10, 2006.

5.0 Environmental Consideration The scope of the proposed amendment is limited to the categorical exclusion provided by 10 CFR 51.21(c)(10)(ii) "Changes recordkeeping, reporting, or administrative procedures or requirements." Therefore, no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

6.0 References

1. Industry/Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler TSTF-5-A, Rev.1 "Delete Safety Limit Violation Notification Requirements."
2. May 10, 2006 letter from Richard V. Guzman (NRC),

Subject:

Peach Bottom Atomic Power Station, Units 2 and 3 - Issuance of Amendments Re: Incorporation of Previously NRC Approved Generic Technical Specification Changes (TAC Nos. MC3683, ... )

El-5

Enclosure 1 Basis for Proposed Change 6.0 References

1. Industry/Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler TSTF-5-A, Rev.1 "Delete Safety Limit Violation Notification Requirements."
2. May 10, 2006 letter from Richard V. Guzman (NRC),

Subject:

Peach Bottom Atomic Power Station, Units 2 and 3 - Issuance of Amendments Re: Incorporation of Previously NRC Approved Generic Technical Specification Changes (TAC Nos. MC3683, ... )

El -6

Edwin I. Hatch Nuclear Plant - Units 1 & 2 Vogtle Electric Generating Plant - Units 1 & 2 License Amendment Request For Incorporation Of Previously NRC Approved Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler TSTF-5-A, Rev. 1, "Delete Safety Limit Violation Notification Requirements" Enclosure 2 Technical Specification Markup Pages

SLs 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs 2.1.1.1 With the reactor steam dome pressure < 785 psig or core flow

< 10% rated core flow:

THERMAL POWER shall be < 24% RTP.

2.1.1.2 With the reactor steam dome pressure a 785 psig and core flow

> 10% rated core flow:

MCPR shall be > 1.07 for two recirculation loop operation or > 1.09 for single recirculation loop operation.

2.1.1.3 Reactor vessel water level shall be greater than the top of active irradiated fuel.

2.1.2 Reactor Coolant System (RCS) Pressure SL Reactor steam dome pressure shall be < 1325 psig.

/-- within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 2.2 SL Violations With any SL violation, the following actions shall be completed:

P o T- 1 V*^thp 1 heF...... the N,, G Oeratqons Q,f GR-tQF, in

.... G ... wih 14!0 CFR 50.72.11 S 2.2.2 Wi~thin 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />'.

2.2.M]1 Restore compliance with all SLs; and 2.2.R]2 Insert all insertable control rods.

2.2.3 W^i+hhR 24 h, .... notify the* Plo,,t Maa&T*

..... Presi-,*dent Hatch andte (continued)

HATCH UNIT 1 2.0-1 Amendment No. M

SLs 2.0 2.0 SAFETY LIMITS (SLs)

P.2 9c RS (GORtwRL4 T--TA Withon 30 days, a L gensee Event Repei4 (LF=R) shall be pFepared purs-W W, 10 GFR 50ý73. The LER shall be submitted to the NRG, tha ollsite -.- M

!--iftee, the Rl-;;nt MR-RaqeF, and the VgGe PresideRt HatGhj QPeFatiE)R ef the WRit Shall Rat be FesuFned until authoFized HATCH UNIT 1 2.0-2 Amendment No. P

SLs 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs 2.1.1.1 With the reactor steam dome pressure < 785 psig or core flow

< 10% rated core flow:

THERMAL POWER shall be <24% RTP.

2.1.1.2 With the reactor steam dome pressure > 785 psig and core flow

> 10% rated core flow:

MCPR shall be > 1.08 for two recirculation loop operation or > 1.10 for single recirculation loop operation.

2.1.1.3 Reactor vessel water level shall be greater than the top of active irradiated fuel.

2.1.2 Reactor Coolant System (RCS) Pressure SL Reactor steam dome pressure shall be < 1325 psig.

A withi~n2 h~ours -]

2.2 SL Violations"/

With any SL violation, the following actions shall be completed.

221 Within I heur, Rtn tM/he,,,,,Nr, Qf nvtiOR, G,*,;enter, in accerd~newt I 2l.2 ithi-2her 2.2.M1 Restore compliance with all SLs; and 2.2..]2 Insert all insertable control rods.

2. .3 Wthin 24 hourS, notify the Paot MaaqF ..... Vo, e Presid en;,t - Hat.h ,,A,- the*

(continued)

HATCH UNIT 2 2.0-1 Amendment No. g

SLs 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs In MODES 1 and 2, the combination of THERMAL POWER, Reactor Coolant System (RCS) highest loop average temperature, and pressurizer pressure shall not exceed the SLs specified in Figure 2.1.1-1.

2.1.2 RCS Pressure SL In MODES 1, 2, 3, 4, and 5, the RCS pressure shall be maintained < 2735 psig.

2.2 SL Violations 2.2.1 If SL 2.1.1 is violated, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

2.2.2 If SL 2.1.2 is violated:

2.2.2.1 In MODE 1 or 2, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

2.2.2.2 In MODE 3, 4, or 5, restore compliance within 5 minutes.

2.2.4- 10 CFR 50.72.

22 VVithiI 24 hurys noif tIUhe Pvulat Ma~a+ F anduuriiiueut ea

-4~ VA
W.qhrn au days a -414--4 A ACr;~

(4 P. 1-',,-

to the NSRC purSuant to 10 CFR 50.73.

2.2.6 Oporation of the unit shall not bo resumed until authorized by the NRC Vogtle Units 1 and 2 2.0-1 Amendment No. P (Unit 1)

Amendment No. (Unit 2)

Edwin I. Hatch Nuclear Plant - Units 1 & 2 Vogtle Electric Generating Plant - Units 1 & 2 License Amendment Request For Incorporation Of Previously NRC Approved Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler TSTF-5-A, Rev. 1, "Delete Safety Limit Violation Notification Requirements" Enclosure 3 Clean Typed Technical Specification Pages

SLs 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs 2.1.1.1 With the reactor steam dome pressure < 785 psig or core flow

< 10% rated core flow:

THERMAL POWER shall be <24% RTP.

2.1.1.2 With the reactor steam dome pressure > 785 psig and core flow

> 10% rated core flow:

MCPR shall be > 1.07 for two recirculation loop operation or > 1.09 for single recirculation loop operation.

2.1.1.3 Reactor vessel water level shall be greater than the top of active irradiated fuel.

2.1.2 Reactor Coolant System (RCS) Pressure SL Reactor steam dome pressure shall be < 1325 psig.

2.2 SL Violations With any SL violation, the following actions shall be completed within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s:

2.2.1 Restore compliance with all SLs; and 2.2.2 Insert all insertable control rods.

2.0-1 Amendment No.

HATCH UNIT 11 2.0-1 Amendment No.

SLs 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs 2.1.1.1 With the reactor steam dome pressure < 785 psig or core flow

< 10% rated core flow:

THERMAL POWER shall be < 24% RTP.

2.1.1.2 With the reactor steam dome pressure > 785 psig and core flow

> 10% rated core flow:

MCPR shall be > 1.08 for two recirculation loop operation or > 1.10 for single recirculation loop operation.

2.1.1.3 Reactor vessel water level shall be greater than the top of active irradiated fuel.

2.1.2 Reactor Coolant System (RCS) Pressure SL Reactor steam dome pressure shall be < 1325 psig.

2.2 SL Violations With any SL violation, the following actions shall be completed within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s:

2.2.1 Restore compliance with all SLs; and 2.2.2 Insert all insertable control rods.

HATCH UNIT 2 2.0-1 Amendment No.

SLs 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs In MODES 1 and 2, the combination of THERMAL POWER, Reactor Coolant System (RCS) highest loop average temperature, and pressurizer pressure shall not exceed the SLs specified in Figure 2.1.1-1.

2.1.2 RCS Pressure SL In MODES 1, 2, 3, 4, and 5, the RCS pressure shall be maintained *2735 psig.

2.2 SL Violations 2.2.1 If SL 2.1.1 is violated, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

2.2.2 If SL 2.1.2 is violated:

2.2.2.1 In MODE 1 or 2, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

2.2.2.2 In MODE 3, 4, or 5, restore compliance within 5 minutes.

Vogtle Units 1 and 2 2.0-1 Amendment No. (Unit 1)

Amendment No. (Unit 2)

Edwin I. Hatch Nuclear Plant - Units 1 & 2 Vogtle Electric Generating Plant - Units 1 & 2 License Amendment Request For Incorporation Of Previously NRC Approved Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler TSTF-5-A, Rev. 1, "Delete Safety Limit Violation Notification Requirements" Enclosure 4 Technical Specification Bases Markup Pages

Reactor Core SLs B 2.1.1 BASES APPLICABLE 2.1.1.3 Reactor Vessel Water Level (continued)

SAFETY ANALYSES active fuel must be adjusted for assemblies with a fuel length not 150 inches. For example, the top of the active fuel for GE13 fuel is 162.44 inches below instrument zero since the fuel length for this fuel type is 146 inches. The Core Operating Limits Report identifies fuel types and fuel lengths used in the current operating cycle.

SAFETY LIMITS The reactor core SLs are established to protect the integrity of the fuel clad barrier to the release of radioactive materials to the environs.

SL 2.1.1.1 and SL 2.1.1.2 ensure that the core operates within the fuel design criteria. SL 2.1.1.3 ensures that the reactor vessel water level is greater than the top of the active irradiated fuel in order to prevent elevated clad temperatures and resultant clad perforations.

APPLICABILITY SLs 2.1.1.1, 2.1.1.2, and 2.1.1.3 are applicable in all MODES.

SAFETY LIMIT VIOLATIONS VIOLATIONS Of any SL is vffo!ated, the NRC Operations Center must be notifed

'h lithin I hour,bR accordance With !0- (CFLR Fin 7) fd \

3' Exceeding an may causefuel damage and create a potential for radioactive r ases in excess of 10 CFR 100, "Reactor Site Criteria,"

limits (Ref. . Therefore, it is required to insert all insertable control rods and restore compliance with the SLs within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Completion Time ensures that the operators take prompt remedial action and also ensures that the probability of an accident occurring during this period is minimal.

if a y Satl 4ioJ*n l fatd th-f~ lager*

n- eRtll l ofhel RU ea F plntii* *a~t d the .. Review B S shall .Actinued (continued)

HATCH UNIT 1 B 2.0-4

Reactor Core SLs B 2.1.1 BASES I FE Y I I 2.2.341 E) submitte-within 30 dgy'-te.-he-NRG-irn-aoror4anee-wi.th--0-1GFl 57 Ref*.-- Aen- of the..re rt-shaIq.also-be feved.-

Isener-*,,,,*. ........... .uelea*- larat-afid4he-ut~iit dAa4the-SRg.

I !f ny SLis volated, resta ,,t of the u nit shall not co)m m e nce unti Ia", thrized by, the, NRG. This...... i.... t ensu~res the NRC that a!!,-1 h~i beqqR ka- itS ire~t~,-toI-H RF *mal e* a-*in REFERENCES 1. 10 CFR 50, Appendix A, GDC 10.

2. NEDE-2401 1-P-A, "General Electric Standard Application for 3 Reactor Fuels" (revision specified in the COLR).

P. 10 CFR 500.1 R. 10 CFR 100.

HATCH UNIT 1 B 2.0-5

RCS Pressure SL B 2.1.2 BASES APPLICABLE Addenda through the Winter of 1966 (Ref. 5), which permits a SAFETY ANALYSES maximum pressure transient of 110%, 1375 psig, of design pressure (continued) 1250 psig. The SL of 1325 psig, as measured in the reactor steam dome, is equivalent to 1375 psig at the lowest elevation of the RCS.

The RCS is designed to the USAS Nuclear Power Piping Code, Section B31.1, 1967 Edition, including Addenda A, C, and D (Ref. 6),

for the reactor recirculation piping, which permits a maximum pressure transient of 120% of design pressures of 1150 psig for suction piping and 1325 psig for discharge piping. The RCS pressure SL is selected to be the lowest transient overpressure allowed by the applicable codes.

SAFETY LIMITS The maximum transient pressure allowable in the RCS pressure vessel under the ASME Code,Section III, is 110% of design pressure.

The maximum transient pressure allowable in the RCS piping, valves, and fittings is 120% of design pressures of 1150 psig for suction piping and 1325 psig for discharge piping. The most limiting of these two allowances is the 110% of the reactor vessel design pressure; therefore, the SL on maximum allowable RCS pressure is established at 1325 psig as measured at the reactor steam dome.

APPLICABILITY SL 2.1.2 applies in all MODES.

SAFETY LIMIT VIOLATIONS Ifa L+-violated, he -NRG-Q ations Gente-mu 2M Exceeding the RCS pressure SL may cause immediate RCS failure and create a potential for radioactive releases in excess of 10 CFR 100, "Reactor Site Criteria," limits (Ref. 4). Therefore, it is 6] required to insert all insertable control rods and restore compliance with the SL within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Completion Time ensures that the operators take prompt remedial action.

(continued)

HATCH UNIT 1

RCS Pressure SL B 2.1.2 BASES VIQLAETV -t10 494 T the Hntilmt ae*;ld the R halb Retnibe wonthan 24 fheu n, , Th~ 2ln o

= ... 2" t -"ides. .......

I * .. . .I . .. . .. . . .. .. ..

4f- h L-s ie

-.-.-.-.-.-.--.. 4Gee see Event R~ef22 .sh4aJl-be- r-efAF ar-e4--an*

s'-bmitted w.AIithmR X) d12Y-s to the NRC mnaccordance With 10- C-FR 5_0_7A

!fayS s vielated, restat ef, thie unRit Shall net comeFAnce uRtl a*ih'ri7Hh the NlRG. This regwu~remen÷.1 l t~e*rS6 Fh*I hr *1 ie.... es a ,,e ,G9 ,w...-,*1 a*t,,)R REFERENCES 1. 10 CFR 50, Appendix A, GDC 14 and GDC 15.

2. ASME, Boiler and Pressure Vessel Code,Section III, Article NB-7000.
3. ASME, Boiler and Pressure Vessel Code,Section XI, Article IW-5000.
4. 10 CFR 100.
5. ASME, Boiler and Pressure Vessel Code,Section III, 1965 Edition, Addenda Winter of 1966.
6. ASME, USAS, Nuclear Power Piping Code, Section B31.1, 1967 Edition, Addenda A, C, and D.

HATCH UNIT 1

Reactor Core SLs B 2.1.1 BASES APPLICABLE 2.1.1.3 Reactor Vessel Water Level (continued)

SAFETY ANALYSES active fuel must be adjusted for assemblies with a fuel length not 150 inches. For example, the top of the active fuel for GEl3 fuel is 162.44 inches below instrument zero since the fuel length for this fuel type is 146 inches. The Core Operating Limits Report identifies fuel types and fuel lengths used in the current operating cycle.

SAFETY LIMITS The reactor core SLs are established to protect the integrity of the fuel clad barrier to the release of radioactive materials to the environs.

SL 2.1.1.1 and SL 2.1.1.2 ensure that the core operates within the fuel design criteria. SL 2.1.1.3 ensures that the reactor vessel water level is greater than the top of the active irradiated fuel in order to prevent elevated clad temperatures and resultant clad perforations.

APPLICABILITY SLs 2.1.1.1, 2.1.1.2, and 2.1.1.3 are applicable in all MODES.

SAFETY LIMIT VIOLATIONS 3

Exceeding an/may cause fuel damage and create a potential for radioactive re ases in excess of 10 CFR 100, "Reactor Site Criteria,"

limits (Ref. . Therefore, it is required to insert all insertable control rods and restore compliance with the SLs within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Completion Time ensures that the operators take prompt remedial action and also ensures that the probability of an accident occurring during this period is minimal.

24huFs-T-h -eu2A-y e -r-o (continued)

HATCH UNIT 2 B 2.0-4 I*'Tý*S'Ni 4

Reactor Core SLs B 2.1.1 BASES ISAFETY LIMIT 2.2" /.3 ;5;ý VIJI'l A TIUM7-IlI REFERENCE

1. OCf 50h AxAe nd tGDC 1 R....
2. anEDESL 41-Pvi-lateA, GaLicensee Evectric Saenrd,, App.licatio f IubkRetacto Fiuhiln s,"

devs t hec s*p in th*e)FCO2LR**.*

stafe te to ppF2ra tak I [f_ S,,__ -v_*.,,t, d, F*t~

,, -o, f the unit shall -not-Goommn~e-4*jq REFERENCES 1. 10 CFR 50, Appendix A, GDC 10.

2. NEDE-2401 1-P-A, "General Electric Standard Application for Reactor Fuels," (revision specified in the COLR).

. ...... 10 CFR5*72 10 CFR 100.

113 Q

HATCH UNIT 2 B 2.0-5

RCS Pressure SL B 2.1.2 B 2.0 SAFETY LIMITS (SLs)

B 2.1.2 Reactor Coolant System (RCS) Pressure SL BASES BACKGROUND The SL on reactor steam dome pressure protects the RCS against

-overpressurization. In the event of fuel cladding failure, fission products are released into the reactor coolant. The RCS then serves as the primary barrier in preventing the release of fission products into the atmosphere. Establishing an upper limit on reactor steam dome pressure ensures continued RCS integrity. Per 10 CFR 50, Appendix A, GDC 14, "Reactor Coolant Pressure Boundary," and GDC 15, "Reactor Coolant System Design" (Ref. 1), the reactor coolant pressure boundary (RCPB) shall be designed with sufficient margin to ensure that the design conditions are not exceeded during normal operation and anticipated operational occurrences (AOOs).

During normal operation and AOOs, RCS pressure is limited from exceeding the design pressure by more than 10%, in accordance with Section III of the ASME Code (Ref. 2). To ensure system integrity, all RCS components are hydrostatically tested at 125% of design pressure, in accordance with ASME Code requirements, prior to initial operation when there is no fuel in the core. Any further hydrostatic testing with fuel in the core may be done under LCO 3.10.1, "Inservice Leak and Hydrostatic Testing Operation."

Following inception of unit operation, RCS components shall be pressure tested in accordance with the requirements of ASME Code,Section XI (Ref. 3).

Overpressurization of the RCS could result in a breach of the RCPB, reducing the number of protective barriers designed to prevent radioactive releases from exceeding the limits specified in 10 CFR 100, "Reactor Site Criteria" (Ref. 4). If this occurred in conjunction with a fuel cladding failure, fission products could enter the containment atmosphere.

APPLICABLE The RCS safety/relief valves and the Reactor Protection System SAFETY ANALYSES Reactor Vessel Steam Dome Pressure - High Function have settings established to ensure that the RCS pressure SL will not be exceeded.

The RCS pressure SL has been selected such that it is at a pressure below which it can be shown that the integrity of the system is not endangered. The reactor pressure vessel is designed to Section III of the ASME, Boiler and Pressure Vessel Code, 1968 Edition, including (continued)

HATCH UNIT 2 B 2.0-6 REVISION 0

RCS Pressure SL B 2.1.2 BASES APPLICABLE Addenda through the Summer of 1970 (Ref. 5), which permits a SAFETY ANALYSES maximum pressure transient of 110%, 1375 psig, of design pressure (continued) 1250 psig. The SL of 1325 psig, as measured in the reactor steam dome, is equivalent to 1375 psig at the lowest elevation of the RCS.

The RCS is designed to Section III of the ASME, Boiler and Pressure Vessel Code, 1980 Edition, including addenda through Winter 1981 (Ref. 6), for the reactor recirculation piping, which permits a maximum pressure transient of 110% of design pressures of 1250 psig for suction piping and 1450 psig for discharge piping. The RCS pressure SL is selected to be the lowest transient overpressure allowed by the applicable codes.

SAFETY LIMITS The maximum transient pressure allowable in the RCS pressure vessel under the ASME Code,Section III, is 110% of design pressure.

The maximum transient pressure allowable in the RCS piping, valves, and fittings is 110% of design pressures of 1250 psig for suction piping and 1450 psig for discharge piping. The most limiting of these two allowances is the 110% of the reactor vessel and recirculation suction piping design pressure; therefore, the SLon maximum allowable RCS pressure is established at 1325 psig as measured at the reactor steam dome.

APPLICABILITY SL 2.1.2 applies in all MODES.

SAFETY LIMIT VIOLATIONS 114:- -

I" any , 0-81-4s-Aolated, 1

the ISIRgE22fnf-2tions GenteF must be notifie-WitNR 1 h-- ,, 6- ----- danGe with 10 CPR 50.72 ýRef4fl Exceeding the RCS pressure SL may cause immediate RCS failure and create a potential for radioactive releases in excess of 10 CFR 100, "Reactor Site Criteria," limits (Ref. 4). Therefore, it is required to insert all insertable control rods and restore compliance with the SL within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Completion Time ensures that the operators take prompt remedial action.

(continued)

HATCH UNIT 2 B 2.0-7

RCS Pressure SL B 2.1.2 BASES

  • and te SRBshah e notfiedj within 2ý4 hoers. The24hu immed"*atie ar-÷in and assesse the G,*,ntifi.*n o~f the uR hfere bne renpe444d If---a y-SL is-violaeq-+/-ensee-F=ent-BR 2a4-be-prepared I 1submL'ted within 30 days tc the R"- ...... ,4r....... Wih 10 FRI 50.73 (Ref. 8). A G-OnY of the rFbnert shall alsben phvie nnte~ar tnhe sepJ44G-I,,* * ,*...* is, violated , reesta irt. of. thie un it sha ll n ot co mn m e nce URNt!
  • ,,h,,P,*a y!he NRC, This requirement ........ the NR, hG .

!unit bpi-n- Us; rM;÷u te Re.*I er.*Ha REFERENCES 1. 10 CFR 50, Appendix A, GDC 14 and GDC 15.

2. ASME, Boiler and Pressure Vessel Code,Section III, Article NB-7000.
3. ASME, Boiler and Pressure Vessel Code,Section XI, Article IW-5000.
4. 10 CFR 100.
5. ASME, Boiler and Pressure Vessel Code,Section III, 1968 Edition, Addenda Summer of 1970.
6. ASME, Boiler and Pressure Vessel Code,Section III, 1980 Edition, Addenda Winter of 1981.

S7. 10 QCFR 5.0.72.1 HATCH UNIT 2 B 2.0-8

Reactor Core SLs B 2.1.1 BASES SAFETY LIMIT VIOLATIONS (continued) If the reactor core SL 2.1.1 is violated, the requirement to go to MODE 3 places the unit in a MODE in which this SL is not applicable.

The allowed Completion Time of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> recognizes the importance of bringing the unit to a MODE of operation where this SL is not applicable, and reduces the probability of fuel damage.

if the reactor core SL= 2.1.1 is violated, the NRC OperationS Center must be notified within 1 hour*, in aordanR e with 10 CFR 50.72 (Ref.

2.2.A' ifthe reactor Gore 9L 2.1.1 is vielatcd, the Plant Manager and the

,President Vogtlc shall be notified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. This 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Vee per*id p..v.des time for the plant operators and staff to take the appnropriate immodliate atonfin And aseSS the Gondtfion of the unit befre r-epo*inRg t the senPior managemenRt.

2.2.5 If the r..a.tPr.o.re 2L 2.1.1 is violated, a Licensee EveRt Report shall be prepared and submitted within 30 days to the NRC. This requi r. m t is in acor.dance With 10 CFR 50.73 (Ref. 6).

2.2.6 if the reaGcGtr core 9L 2.1.1 is violated, restart of the unit shall not commnRe u4ntl authorized by the NRC. This requirement ensures-t-heNRG that all neceS revli

- iews, analyses, and actions are comnpleted-befe e the U~itbegiRs its restart to normal opotin (continued)

Vogtle Units 1 and 2 B 2.1.1-5

Reactor Core SLs B 2.1.1 BASES (continued)

REFERENCES 1. 10 CFR 50, Appendix A, GDC 10.

2. FSAR, Section 7.2.
3. WCAP-8746-A, March 1977.
4. WCAP-9272-P-A, July 1985.
5. 10 CFR 50.72.
6. 10 CFR 50.73.

Vogtle Units 1 and 2 B 2.1.1-6

RCS Pressure SL B 2.1.2

,BASES SAFETY LIMITS Code,Section III, is 110% of design pressure. Therefore, the SL (continued) on maximum allowable RCS pressure is 2735 psig.

APPLICABILITY SL 2.1.2 applies in MODES 1,2, 3, 4, and 5 because this SL could be approached or exceeded in these MODES due to overpressurization events. The SL is not applicable in MODE 6 because the reactor vessel head closure bolts are not fully tightened, making it unlikely that the RCS can be pressurized.

SAFETY LIMIT Section 2.2, SL= Violations, previdcs, the Required ActioRn to be taken VIOLATIONS inrsose to a violation of a Safety Limit. The bases far the Required ActioRn* f SectioR 2.2 applicable to a ViOlation of thc RCS pressure S.L arc diSuss.d below.

2-22ý The Required Actions of this subsection state the specific status in whih4he-uit-mus- be placed if the RCS-pressure SL is vielated.

Sepa.atRequ4ird*A*tie*_-and Completion Times are provided_ fo MODES ! or 2 (Subsection 2.2.2.1) and for MODES 3, 4, or 5 (Subsection 2.2.2.2).

If the RCS pressure SL 2.2.2 is violated when the reactor is in MODE 1 or 2, the requirement is to restore compliance and be in MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

Exceeding the RCS pressure SL may cause immediate RCS failure and create a potential for radioactive releases in excess of 10 CFR 100, "Reactor Site Criteria," limits (Ref. 4).

The allowable Completion Time of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> recognizes the importance of reducing power level to a MODE of operation where the potential for challenges to safety systems is minimized.

(continued)

Vogtle Units 1 and 2 B 2.1.2-3

RCS Pressure SL B 2.1.2 BASES SAFETY LIMIT VIOLATIONS (continued) If the RCS pressure SL 2.2.2 is exceeded in MODE 3, 4, or 5, RCS pressure must be restored to within the SL value within 5 minutes.

Exceeding the RCS pressure SL in MODE 3, 4, or 5 is more severe than exceeding this SL in MODE 1 or 2, since the reactor vessel temperature may be lower and the vessel material, consequently, less ductile. As such, pressure must be reduced to less than the SL within 5 minutes. The action does not require reducing MODES, since this would require reducing temperature, which would compound the problem by adding thermal gradient stresses to the existing pressure stress.

if the RCS pressure SL is violated, the NR, Operations enteFr must be netified within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, in acGordaRne with 10 CFR 50.72 (Ref. 6).

2~4 g4S ki I* _ *&l I R -.. --!- --- ---/ I . .. .l ~ *._ _1 ._ *ll  !. t ~ I. . . .. T _ _ I^ A . . ..

wviee'-Presiaent Vogie, s,*fl,, -henut*e* w'fn-s4 -+ur.-

period provides time for the plant operators and staff to take the roite immediate n nassess the oditi of the unit before reporting to ei management.

ifthe RCS pressure SL= 2.2.2 is violated, a Licensee Event Report shall be prepared and submitted within 30 days to the NRC. This eir-men4t.4-4-aeoedaRGeth--40-GF:R-50.73 (Ref. 7).

if the RCS pressure SL2-2-.-is-velated esta4tefthe t shah net commence until authorized by the NRC. This requrent ensures the NRC that all necessary reviews, (continued)

Vogtle Units 1 and 2 B 2.1.2-4

RCS Pressure SL B 2.1.2 BASES SAFETY LIMIT 2-2-6 VIOLATIONS (continued) a-alyses --tieR a o omeetedGee4he-+e-u-gins-its-sr topEwmloeatioR.

REFERENCES 1. 10 CFR 50, Appendix A, GDC 14, GDC 15, and GDC 28.

2. ASME, Boiler and Pressure Vessel Code,Section III, Article NB-7000.
3. ASME, Boiler and Pressure Vessel Code,Section XI, Article IWB-5000.
4. 10 CFR 100.
5. FSAR, Section 7.2.

6, 10 CFR 5072.

7. 10 0FR 50.73.

Vogtle Units 1 and 2 B 2.1.2-5

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