NL-15-0217, Response to Request for Additional Information Regarding Multiple Technical Specifications Changes

Response to Request for Additional Information Regarding Multiple Technical Specifications Changes

ML15058A891
Person / Time
Site:	Vogtle
Issue date:	02/27/2015
From:	Pierce C Southern Co, Southern Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-15-0217
Download: ML15058A891 (20)

Text

Charles R. Pierce Regulatory Affairs Director Southern Nuclear Operating Company, Inc.

40 Inverness Center Parkway Post Office Box 1295 Birmingham, AL 35242 Tel 205.992.7872 Fax 205.992.7601 SOUTHERN A COMPANY February 27, 2015 Docket Nos.: 50-424 50-425 U.S. Nuclear Regulatory Commission ATIN: Document Control Desk Washington, D. C. 20555-0001 Vogtle Electric Generating Plant-Units 1 and 2 Response to Request for Additional Information Regarding Multiple Technical Specifications Changes Ladies and Gentlemen:

NL-15-0217 By letter dated July 18, 2014 (Agencywide Documents Access and Management System Accession No. [ACN] ML14203A124), Southern Nuclear Operating Company (SNC) submitted a license amendment request (LAR) to adopt various previously NRC-approved Technical Specifications Task Force (TSTF) Travelers.

By letter dated January 13, 2015, the Nuclear Regulatory Commission (NRC) sent SNC a request for additional information (RAI). Enclosure 1 to this letter contains SNC's responses to those NRC questions, which are transcribed prior to each SNC response. In some of the responses, SNC proposes to revise the Technical Specifications clean-typed pages and the Bases mark-up pages that were submitted as part of the LAR. Enclosure 2 of this letter contains those proposed revisions. Upon staff agreement with the proposed changes, SNC will promptly provide the updated package to the NRC.

This letter contains no NRC commitments. If you have any questions, please contact Ken McElroy at (205) 992-7369.

U.S. Nuclear Regulatory Commission NL-15-0217 Page2 Mr. C. R. Pierce states he is Regulatory Affairs 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.

Respectfully submitted,

c. r< ~

C. R. Pierce Regulatory Affairs Director CRP/EGA Sworn to and s bscribed before me this ~ 1 day of J=e. b,...oa...,., 2015.

My commission expires: /* Z.- 2.0 I I

References:

1. Letter from SNC to the NRC dated July 18, 2014, NL-14-0706, Vogtle Electric Generating Plant-Units 1 and 2, Application to Revise Technical Specifications to Adopt Previously NRC-Approved Generic Technical Specification Changes

2. Letter from the NRC to SNC dated January 13, 2015, Vogtle Electric Generating Plant-Request for Additional Information on Multiple Technical Specification Changes (TAG Nos. MF4515 and MF4516)

Enclosures:

1. SNC Response to Request for Additional Information on Multiple Technical Specification Changes

2. Revised Technical Specifications and Bases Pages

U.S. Nuclear Regulatory Commission NL-15-0217 Page3 cc:

Southern Nuclear Operating Company Mr. S. E. Kuczynski, Chairman, President & CEO Mr. D. G. Bost, Executive Vice President & Chief Nuclear Officer Mr. D. R. Madison, Vice President-Fleet Operations Mr. M.D. Meier, Vice President-Regulatory Affairs Mr. B. K. Taber, Vice President-Vogtle 1 & 2 Mr. B. J. Adams, Vice President-Engineering Mr. G.W. Gunn, Regulatory Affairs Manager-Vogtle 1 & 2 RType: CVC7000 U.S. Nuclear Regulatory Commission Mr. V. M. McCree, Regional Administrator Mr. R. E. Martin, NRR Senior Project Manager-Vogtle Mr. L. M. Cain, Senior Resident Inspector-Vogtle State of Georgia Mr. J. H. Turner, Environmental Director Protection Division

Vogtle Electric Generating Plant Request for Technical Specifications Amendment Adoption of Previously NRC-Approved Generic Technical Specification Changes SNC Response to Request for Additional Information on Multiple Technical Specification Changes to NL-15-0217 SNC Response to Request for Additional Information on Multiple Technical Specification Changes RAI3.1.4-1

{TSTF-110)

RAI 3.1.7-1

{TSTF-234)

The proposed changes include a change to the FREQUENCY of SR 3.1.4.1. The license amendment request (LAR) includes an accurate markup for SR 3.1.4.1 in the TS, but does not contain a mark-up for the Bases for SR 3.1.4.1, which is found on page B3.1.4-10 of the current Vogtle TS Bases. The NRC staff requests that a markup for the Bases be provided.

SNC Response The Frequency for Surveillance Requirement (SR) 3.1.4.1 in the current Vogtle Unit 1 and 2 (VEGP) Technical Specifications {TS) is consistent with those the Standard Technical Specifications {STS) 3.1.5, on which TSTF-11 0 is based, and is stated as:

In accordance with the Surveillance Frequency Control Program Once within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> thereafter when the rod position deviation monitor is inoperable.

The Bases for SR 3.1.4.1 provided in the STS includes a discussion of the basis for the second frequency, and TSTF-110 provides a markup of the Bases page for SR 3.1.4.1 to reflect elimination of the second surveillance frequency. The current VEGP Bases for SR 3.1.4.1 does not include a discussion of the second frequency, and therefore no markup was necessary to reflect its elimination.

The proposed change includes the insertion of Condition B in the TS and Bases, which is consistent with the Standard Technical Specification (STS). However, the third paragraph of the Condition B insertion ends with a reference to "(Ref.

4)." The current VEGP TS Bases for Section 3.1.7, as reflected on the mark-up page B3.1.7-6, only has 2 references in the References section. The NRC staff requests that this inconsistency be addressed and that References 3 and 4 be provided for evaluation if applicable.

SNC Response Inclusion of the text "(Ref. 4)" in the Bases markup page forTS 3.1.7, Condition B was an editorial oversight. A revised markup of the Bases Insert page that removes this text is provided in Enclosure 2.

E1-1 to NL-15-0217 SNC Response to Request for Additional Information on Multiple Technical Specification Changes RAI 3.4.11-1 (TSTF-247)

TS 3.4.11 Condition F and Required Actions are revised to allow a separate limiting condition for operability (LCO) entry for each inoperable block valve. The NRC staff noted the following editorial errors in the Bases discussion of the revised Action F.1 :

1. In the heading, "F.2 and F.3" are no longer applicable

2. In the first sentence, there is an extra "restore" The NRC staff requests that the licensee correct these errors.

SNC Response The text "F.2 and F.3" have been removed from the heading in the Bases discussion of Action F, and the extra "restore" has been deleted. A revised Bases markup page reflecting these changes is provided in Enclosure 2.

RAI 3.4.11-2 SR 3.4.11.1 requires block valve cycling to verify the valve can be closed. The (TSTF-284)

SR is modified by two notes, and the Bases discussion of the notes is different from the TSTF discussion in three places:

RAI 3.9.4-1 (TSTF-312)

1. The second Bases paragraph in the TSTF begins, "This SR modified by two notes," while the proposed sentence says, "This SR has two notes."

2. The first sentence of the TSTF Insert A is missing and needs to be included; it begins, "Opening the block valve in this condition... "

3. The next sentence of the TSTF reads, "Note 2 modifies this SR to... " The proposed sentence is missing the word "modifies" which needs to be included.

These changes need to made, consistent with the approved TSTF.

SNC Response The Bases markup insert for SR 3.4.11.1 has been revised to adopt the standard wording from TSTF-284. A revised Bases Insert page reflecting these changes is provided in Enclosure 2.

To explain the difference between TSTF-312-A requirements and the proposed changes to VEGP TS 3.9.4 and its associated Bases, the licensee states that LCO 3.9.4.b was previously amended to allow the emergency and personnel airlocks to remain open during CORE ALTERATIONS or during movement of irradiated fuel assemblies within the containment, and that the scope of this previous amendment (VEGP Amendments 92/70, dated November 30, 1995) overlaps the scope of TSTF-312-A, resulting in the statement of LCO 3.9.4 and its associated bases not being identical to those presented in TSTF 312 A.

E1-2 to NL-15-0217 SNC Response to Request for Additional Information on Multiple Technical Specification Changes RAI3.9.4-2 (TSTF-312)

The staff reviewed the referenced Amendments 92/70, and noted that the emergency air lock was not included in the scope of Amendments 92/70. The NRC staff further noted that only the requirements for the personnel airlock are considered in TSTF-312-A since the emergency air lock normally has not been used for personnel entry into or exit from inside of the containment during a plant refueling outage. In addition, the NRC staff noted that no discussion was provided for the open equipment hatch in the application.

The staff requests that the licensee identify the VEGP license amendments that approved allowing the equipment hatch and the emergency airlock to remain open during CORE ALTERATIONS and during movement of irradiated fuel assemblies within the containment.

SNC Response Amendment 92/70 approved changes toTS 3.9.4 that allow the personnel airlock to be open during core alterations or movement of irradiated fuel within the containment, provided a designated individual is available to close the personnel airlock door (Agencywide Documents Access and Management System Accession No. [ACN] ML012350007). NRC review of the Vogtle FHA confirmatory dose calculation was documented in the Safety Evaluation Report (SEA) that accompanied this license amendment.

Changes toTS 3.9.4 to allow the emergency airlock to be open during core alterations or movement of irradiated fuel within the containment, provided a designated individual is available to close the personnel airlock door were approved in Amendment 105/83 (ACN ML012390325). As stated in the SEA that accompanied this amendment, the bases for the change to allow open emergency airlock doors are the same as for the previously approved personnel airlock doors.

Changes toTS 3.9.4 to allow the equipment hatch to be open during core alterations or movement of irradiated fuel within the containment, provided that capability for closure is maintained was provided in Amendment 115/93 (ACN ML003749439). The prior NRC review of the Vogtle FHA confirmatory dose calculation that was performed for Amendment 92/70 was documented in the Safety Evaluation Report that accompanied this license amendment.

TSTF-312-A includes a Reviewer's Note that identifies the need for a confirmatory fuel handling accident (FHA) dose calculation that is accepted by the NRC staff, and that indicates acceptable radiological consequences. This Reviewer's note also states "the time to close such penetrations or combination of penetrations shall be included in the confirmatory dose calculations." In the application, the licensee only provides a commitment to incorporate the time needed to close open containment penetration(s) into the existing VEGP FHA dose calculations.

E1-3 to NL-15-0217 SNC Response to Request for Additional Information on Multiple Technical Specification Changes The NRC staff requests that the licensee submit the revised FHA dose calculations for its review as part of the application, or provide justification that a confirmatory FHA dose calculation is not needed for the open containment penetrations.

SNC Response The reference to Amendment 92/70 is corrected to indicate that the Vogtle confirmatory FHA dose calculation supporting the current licensing basis was reviewed and accepted by the NRC in the Safety Evaluation Report for license amendments 149/129. This acceptance is documented in a letter from Siva P.

Lingam (NRC) to Tom E. Tynan (SNC), dated February 27, 2008 (ACN ML080350347). Specific details of the FHA dose calculation (i.e., inputs, outputs, assumptions) were provided by SNC in letters dated August 28, 2007 (ACN ML072470691 ), October 9, 2007 (ACN ML072850108), and December 21, 2007 (ACN ML073580035).

The Vogtle FHA dose calculation provides a bounding analysis of offsite and control room doses for FHA events within the containment or fuel handling building. The containment building exhaust release rate is established based on an assumption that 99.9% of the activity that is released from the reactor cavity or refueling pool during the FHA event is released to the environment within the initial 0-2 hour period. This assumption is conservative since no pressurization of containment occurs as a result of the accident, and the containment building exhaust release rate is not dependent on the size or number of open containment penetrations. The release is modeled as ground level, and there is fundamentally no contribution to calculated doses after the first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of the event. It is assumed that all activity released from the reactor cavity or the refueling pool into the containment is released to the environment prior to isolation of any open containment penetrations.

Calculated doses for the FHA event, with either simultaneous or individual release though one or more open containment penetrations, an open equipment hatch, or open personnel airlock doors, is shown to be within the Standard Review Plan criteria of 6 REM to the whole body and 75 REM to the thyroid at the exclusion area and low population zone boundaries. Calculated doses are also within the control room dose acceptance criteria from 10 CFR 50, Appendix A, General Design Criterion 19, of 5 REM whole body, and the interpreted limit of 50 REM thyroid, as described in R.G. 1.195, "Methods and Assumptions for Evaluating Radiological Consequences of Design Basis Accidents at Light-Water Nuclear Power Reactors." Calculated doses are also within 10 CFR Part 100 limits.

Due to the fact that the FHA dose consequences are shown to be within acceptance limits without assuming that any containment penetration leak path(s) are isolated in response to the event, there is no need to establish a required time to close unisolated containment penetration(s), or incorporate these times into the FHA confirmatory dose calculation. Consistent with the changes identified in TSTF-312-A, TS 3.9.4, and the associated Bases, will be E1-4 to NL-15-0217 SNC Response to Request for Additional Information on Multiple Technical Specification Changes RAI 5.5.17-1 (TSTF-343) revised to include a Note that directs isolation of open containment penetrations by designated and available individuals in the event of an FHA. However, due to the fact that the commitment proposed in the LAR submittal to identify the time to close such penetrations or combination of penetrations in the confirmatory dose calculations is not necessary, it is withdrawn.

To explain the difference between TSTF-343-A changes in STS requirements and the proposed changes to VEGP TS 5.5.17 requirements, the licensee states that the changes identified for VEGP TS 5.5.6, "Prestressed Concrete Containment Tendon Surveillance Program," and conforming changes to the TS Bases for SR 3.6.1.2 and a reference to RG 1.35 are not adopted because those changes are already reflected in the current VEGP TS and Bases. The applicable VEGP license amendment numbers for those changes were not stated in the application. Similarly, conforming changes to the TS Bases for SR 3.6.1.1, due to the proposed change to inspection requirements for the steel liner plate, are not adopted in this license amendment request. Further, the NRC staff noted that the application does not discuss TSTF-343-A changes related to inspection requirements for the containment outside concrete surfaces.

The NRC staff requests that the licensee address all other proposed changes in TSTF-343-A that are not proposed for adoption by this license amendment request, and identify the applicable VEGP license amendments that approved those changes for incorporation into the VEGP TS.

SNC Response Changes corresponding those described in TSTF-343-A for VEGP TS 5.5.6, "Prestressed Concrete Containment Tendon Surveillance Program," the TS Bases for SRs 3.6.1.1 and 3.6.1.2, and the Bases References forTS 3.6.1 were approved by the NRC as license amendments 147/127 in a letter from Robert E.

Martin (NRC) to D. E. Grissette (SNC), dated December 12, 2006 (ACNs ML062970484 and ML063480167).

Changes involving requirements for the containment outside concrete surfaces, corresponding those described in TSTF-343-A for VEGP TS 5.5.17 (ISTS TS 5.5.16), "Containment Leakage Rate Testing Program," were approved by the NRC as license amendments 122/100 in a letter from Leonard Olshan (NRC) to J. B. Beasley, Jr. (SNC), dated June 6, 2001 (ACN ML011570674).

RAI 5.5.17-2 In Enclosure 4 of the application, the licensee provides the clean-typed TS (TSTF-343) pages. The NRC staff noted the following:

1. On Page 5.5-16, a paragraph break was inserted in error after "10 CFR 50, Appendix J," in the first paragraph of Specification 5.5.17, and E1-5 to NL-15-0217 SNC Response to Request for Additional Information on Multiple Technical Specification Changes RAI3.9.6-1

{TSTF-349)

2. On Page 5. 5-17, an existing Item 4 regarding an extension of Type A test frequency to "15 years", was replaced with the addition of the proposed change under TSTF-343-A. There is no discussion in the application for the removal of the existing Item 4. The staff requests that the licensee address the above two discrepancies.

SNC Response The discrepancy noted on page 5.5-16 has been corrected and a revised, clean-typed page reflecting the change is provided in Enclosure 2.

With respect to the discrepancy on page 5.5.17, the VEGP Containment Leakage Rate Testing Program, as described in TS 5.5.17, contains a one-time exception to NEI 94-01, Rev. 0, "Industry Guidelines for Implementing Performance-Based Option of 10 CFR 50, Appendix J," that is identified as Item 4. The proposed change from TSTF-343 is inserted into TS 5.5.17 as Item 4, and the one time exception is renumbered and retained in VEG P TS 5.5.17 as Item 5.

Renumbering and retention of Item 4 was intended to be included with the clean-typed and markup pages that were provided in the LAR submittal package.

Clean-typed pages reflecting renumbering and retention of TS 5.5.17, Item 4 and Item 5 are provided in Enclosure 2.

The LCO 3.9.6 Statement is revised to add a second Note permitting all residual heat removal (RHR) pumps to be de-energized for no more than 15 minutes when switching from one RHR train to another.

In the revised Bases discussion of the new LCO Note the NRC staff noted that in the second sentence of the added paragraph, parentheses are placed around "and the core outlet temperature is limited to > 10 degrees F below saturation temperature" for no apparent reason. In accordance with the Writer's Guide, parentheses are used to indicate clarifying details for the preceding text.

The NRC staff requests that the licenses explain the use of parentheses in this case.

SNC Response The use parentheses in the Bases discussion of LCO 3.9.6 was an editorial preference. A revised Bases markup page reflecting removal of the parentheses is provided in Enclosure 2.

E1-6

Vogtle Electric Generating Plant Request for Technical Specifications Amendment Adoption of Previously NRC-Approved Generic Technical Specification Changes Revised Technical Specifications and Bases Pages

BASES APPLICABILITY (continued)

ACTIONS Rod Position Indication B 3.1.7 in which power is generated, and the OPERABILITY and alignment of rods have the potential to affect the safety of the plant.

In the shutdown MODES, the OPERABILITY of the shutdown and control banks has the potential to affect the required SDM, but this effect can be compensated for by an increase in the boron concentration of the Reactor Coolant System.

The ACTIONS table is modified by a Note indicating that a separate Condition entry is allowed for each Gf~~~

AO ~

tAB!!~~ I inoperable rod position indicatorlll"' meffi&upfand ffu}iach t3ef-Mwithl I AO ffiOFe tl'leA oA§ inoperable demand position indicatorliA the bet'lij.

T

.... h_e_R_e-qu-i-re_d_A_c_t-io_n_m_a_y_ a_ls_o_b_e_...., This is acceptable because the Required Actions for each Condition satisfied by ensuring at least once provide appropriate compensatory actions for each inoperable oer 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> that Fa satisfies LCO position indicator.

3.2.1, FdeltaH satisfies LCO 3.2.2, and SHUTDOWN MARGIN is within the t-A-.1- -------------------.

ITSTF-234 1 imits provided in the COLR, indirectly orovided the non-indicating rods When one DRPI nnel per group fails, the position of th ad may 1_a_v_e_n_o_t_b_e_e_n_m_o_v_e_d_. --------~ still be determined by use of the movable incore detectors. Based on experience, normal power operation does not require excessive movement of banks. If a bank has been significantly moved, the Insert Bases 3.1.7-Condition B Vogtle Units 1 and 2 Required Action of.

. below is required. Therefore, verification of RCCA positio ithin the Completion Time of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> is adequate for allowing continu d full power operation, since the probability of simultaneously aving a rod significantly out of position and an event sensitive to that ad position is small.

C.1 or C.2 Reduction of THERMAL POWER to s 50% RTP puts the core into a condition where rod position is not significantly affecting core peaking factors.

The allowed Completion Time of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> is reasonable, based on operating experience, for reducing power to s 50% RTP from full power conditions without challenging plant systems and allowing for rod position determination by Required Action A.1 above.

(continued)

B3.1.7-4 Revision No.~

Insert Bases 3.1.7-Condition 8 8.1. 8.2. 8.3 and 8.4 When more than one DRPI per group fail, additional actions are necessary to ensure that acceptable power distribution limits are maintained, minimum SDM is maintained, and the potential effects of rod misalignment on associated accident analyses are limited. Placing the Rod Control System in manual assures unplanned rod motion will not occur. Together with the indirect position determination available via movable incore detectors will minimize the potential for rod misalignment.

,TSTF-234 1 The immediate Completion Time for placing the Rod Control System in manual reflects the urgency with which unplanned rod motion must be prevented while in this Condition. Monitoring and recording reactor coolant T av9 help assure that significant changes in power distribution and SDM are avoided. The once per hour Completion Time is acceptable because only minor fluctuations in RCS temperature are expected at steady state plant operating conditions.

The position of the rods may be determined indirectly by use of the movable incore detectors.

The Required Action may also be satisfied by ensuring at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> that Fa(Z) satisfies LCO 3.2.1, F:H satisfies LCO 3.2.2, and SHUTDOWN MARGIN is within the limits provided in the COLR, provided the non-indicating rods have not been moved. Verification of RCCA position once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> is adequate for allowing continued full power operation for a limited, 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period, since the probability of simultaneously having a rod significantly out of position and an event sensitive to that rod position is small. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Completion Time provides sufficient time to troubleshoot and restore the DRPI system to operation while avoiding the plant challenges associated with a shutdown without full rod position indication.

Based on operating experience, normal power operation does not require excessive rod movement. If one or more rods has been significantly moved, the Required Action of C.1 or C.2 below is required.

BASES ACTIONS (continued)

Vogtle Units 1 and 2 Pressurizer PORVs B 3.4.11 E.1. E.2. E.3. and E.4 If more than one PORV is inoperable and not capable of being manually cycled, it is necessary to either restore at least one valve within the Completion Time of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or isolate the flow path by closing and removing the power to the associated block valves. The Completion Time of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is reasonable, based on the small potential for challenges to the system during this time and provides the operator time to correct the situation. If one PORV is restored and one PORV remains inoperable, then the plant will be in Condition B with the time clock started at the original declaration of having two PORVs inoperable. If no PORVs are restored within the Completion Time, then the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 4 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. In MODES 4, 5, and 6, maintaining PORV OPERABILITY may be required. See LCO 3.4.12.

F.11. F.2. and F.31 two block valves are

~vitfiin 72 heur~. The Completion small potential for challenges to th'-:-e-:s:-:-:y-:;

stc:-e-::-:m~*n uring this time and provide the operator time to correct the situat1 G.1 and G.2 If the Required Actions of Condition F are not met, then the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 4 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant ITSTF-247 1 (continued)

B 3.4.11-6 Revision No. 0

BASES ACTIONS SURVEILLANCE REQUIREMENTS

'INSERT-Bases SR I 3.4.11.1

)

Pressurizer PORVs B 3.4.11 G.1 and G.2 (continued) conditions from full power conditions in an orderly manner and without challenging plant systems. In MODES 4, 5, and 6, maintaining PORV OPERABILITY may be required. See LCO 3.4.12.

SR 3.4.11.1 Block valve cycling verifies that the valve(s) can be closed if needed.

The Surveillance Frequency is controlled under the Surveillance 1:=-::s;:~w::::::s;rgthot l SR 3.4.11.2 SR 3.4.11.2 requires a complete cycle of each PORV. Operating a PORV through one complete cycle ensures that the PORV can be manually actuated for mitigation of an SGTR. The Surveillance Frequency is controlled under the Surveillance Frequency Control ITSTF-284 1 INSERT-Bases SR l~----~7rogram.

_3.4.11.2 REFERENCES INSERT-Bases 3.4.11 Reference Vogtle Units 1 and 2

1. Regulatory Guide 1.32, February 1977.

B 3.4.11-7 REVISION IH]

INSERT - Bases SR 3.4.11.1 This SR is modified by two Notes. Note 1 modifies this SR by stating that it is not required to be performed with the block valve closed in accordance with the Required Actions of this LCO. Opening the block valve in this condition increases the risk of an unisolable leak from the RCS since the PORV is already inoperable. Note 2 modifies this SR to allow entry into and operation in MODE 3 prior to performing the SR. This allows the test to be performed in MODE 3 under operating temperature and pressure conditions, prior to entering MODE 1 or 2. In accordance with Reference 2, administrative controls require this test be performed in MODE 3 or 4 to adequately simulate operating temperature and pressure effects on PORV operation.

INSERT - Bases SR 3.4.11.2 ITSTF-284 1 The Note modifies this SR to allow entry into and operation in MODE 3 prior to performing the SR. This allows the test to be performed in MODE 3 under operating temperature and pressure conditions, prior to entering MODE 1 or 2. In accordance with Reference 2, administrative controls require this test be performed in MODE 3 or 4 to adequately simulate operating temperature and pressure effects on PORV operation.

Insert-Bases 3.4.11 Reference

2. Generic Letter 90-06, "Resolution of Generic Issue 70, 'Power-Operated Relief Valve and Block Valve Reliability,' and Generic Issue 94, 'Additional Low-Temperature Overpressure for Light-Water Reactors,' Pursuant to 10 CFR 50.54(f),"June 25, 1990.

BASES LCO (continued)

RHR and Coolant Circulation -

Low Water Level B 3.9.6 Additionally, one loop of RHR must be in operation in order to provide:

a.

Removal of decay heat;

b.

Mixing of borated coolant to minimize the possibility of criticality; and

c.

Indication of reactor coolant temperr=-

a=

tu:.:...;re7

.:--:--==:----::---:-~~

, ~two Notes. The first Note I The second Note permits the This LCO is modified by a Note that l~lows one RHR loop to be RHR pumps to be de-inoperable for a period of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> provided the other loop is energized for </= 15 minutes OPERABLE and in operation. Prior to declaring the loop inoperable, when switching from one consideration should be given to the existing plant configuration. This train to another. The consideration should include that the core time to boil is short, there is==,...-.,....,......

circumstances for stopping no draining operation to further reduce RCS water level and that the ITSTF-349 both RHR pumps are to be capability exists to inject borated water into the reactor vessel. This limited to situations when the permits surveillance tests to be performed on the inoperable loop outage time is short and the ~during a time when these tests are safe and possible.

core outlet temperature is limited to > 10 degrees F An OPERABLE RHR loop consists of an RHR pump, a heat below saturation exchanger, valves, piping, instruments and controls to ensure an temperature. The Note OPERABLE flow path and to determine the low end temperature. The prohibits boron dilution or flow path starts in one of the RCS hot legs and is returned to the RCS draining operations when cold legs.

RHR forced flow is stopped.

APPLICABILITY ACTIONS Vogtle Units 1 and 2 Two RHR loops are required to be OPERABLE, and one RHR loop must be in operation in MODE 6, with the water level < 23 ft above the top of the reactor vessel flange, to provide decay heat removal and mixing of the borated coolant. Requirements for the RHR System in other MODES are covered by LCOs in Section 3.4, Reactor Coolant System (RCS), and Section 3.5, Emergency Core Cooling Systems (ECCS). RHR loop requirements in MODE 6 with the water level

23 ft are located in LCO 3.9.5, "Residual Heat Removal (RHR) and Coolant Circulation-High Water Level."

A.1 and A.2 If less than the required number of RHR loops are OPERABLE, action shall be immediately initiated and continued until the RHR loop is (continued)

B 3.9.6-2 REVISION @]

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.16 5.5.17 MS and FW Piping Inspection Program This program shall provide for the inspection of the four Main Steam and Feedwater lines from the containment penetration flued head outboard welds, up to the first five-way restraint. The extent of the inservice examinations completed during each inspection interval (ASME Code Section XI) shall provide 100%

volumetric examination of circumferential and longitudinal welds to the extent practical. This augmented inservice inspection is consistent with the requirements of NRC Branch Technical Position MEB 3-1, "Postulated Break and Leakage Locations in Fluid System Piping Outside Containment," November 1975 and Section 6.6 of the FSAR.

Containment Leakage Rate Testing Program A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Testing Program," dated September 1995, as modified by the following exceptions:

1.

Leakage rate testing for containment purge valves with resilient seals is performed once per 18 months in accordance with LCO 3.6.3, SR 3.6.3.6 and SR 3.0.2.

2.

Containment personnel air lock door seals will be tested prior to reestablishing containment integrity when the air lock has been used for containment entry. When containment integrity is required and the air lock has been used for containment entry, door seals will be tested at least once per 30 days during the period that containment entry(ies) is (are) being made.

3.

The visual examination of containment concrete surfaces intended to fulfill the requirements of 10 CFR 50, Appendix J, Option B testing, will be performed in accordance with the requirements of and frequency specified by ASME Section XI Code, Subsection IWL, except where relief or alternative has been authorized by the NRC. At the discretion of the licensee, the containment concrete visual examinations may be performed during either power operation, e.g., performed concurrently with other containment inspection-related activities such as tendon testing, or during a maintenance/refueling outage.

(continued)

Vogtle Units 1 and 2 5.5-16 Amendment No.

Amendment No.

(Unit 1)

(Unit 2)

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.17 Containment Leakage Rate Testing Program (continued)

4.

The visual examination of the steel liner plate inside containment intended to fulfill the requirements of 10 CFR 50, Appendix J, Option 8, will be performed in accordance with the requirements of and frequency specified by the ASME Section XI code, Subsection IWE, except where relief has been authorized by the NRC.

5.

A one time exception to NEI 94-01, Rev. 0, "Industry Guidelines for Implementing Performance-Based Option of 10 CFR 50, Appendix J":

Section 9.2.3:

The next Type A test, after the March 2002 test for Unit 1 and the March 1995 test for Unit 2, shall be performed within 15 years.

The peak calculated primary containment internal pressure for the design basis loss of coolant accident, Pa, is 37 psig.

The maximum allowable containment leakage rate, La, at Pa, is 0.2% of primary containment air weight per day.

Leakage rate acceptance criteria are:

a.

Containment overall leakage rate acceptance criteria are:::;; 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are':::;; 0.60 La for the combined Type 8 and Type C tests, and :::;; 0. 75 La for Type A tests;

b.

Air lock testing acceptance criteria are:

1)

Overall air lock leakage rate is:::;; 0.05 La when tested at~ Pa,

2)

For each door, the leakage rate is:::;; 0.01 La when pressurized to

~ Pa.

The provisions of SR 3.0.2 do not apply to the test frequencies specified in the Containment Leakage Rate Testing Program.

The provisions of SR 3.0.3 are applicable to the Containment Leakage Rate Testing Program.

(continued)

Vogtle Units 1 and 2 5.5-17 Amendment No.

Amendment No.

(Unit 1)

(Unit 2)

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.18 5.5.19 Configuration Risk Management Program The Configuration Risk Management Program (CRMP) provides a proceduralized risk-informed assessment to manage the risk associated with equipment inoperability. The program applies to technical specification structures, systems, or components for which a risk-informed allowed outage time has been granted. The program shall include the following elements:

a.

Provisions for the control and implementation of a Level 1 at power internal events PRA-informed methodology. The assessment shall be capable of evaluating the applicable plant configuration.

b.

Provisions for performing an assessment prior to entering the LCO Condition for preplanned activities.

c.

Provisions for performing an assessment after entering the LCO Condition for unplanned entry into the LCO Condition.

d.

Provisions for assessing the need for additional actions after the discovery of additional equipment out of service conditions while in the LCO Condition.

e.

Provisions for considering other applicable risk significant contributors such as Level 2 issues and external events, qualitatively or quantitatively.

Battery Monitoring and Maintenance Program This program provides for restoration and maintenance, based on the recommendations of IEEE Standard 450-1995, "IEEE Recommended Practice for Maintenance, Testing, and Replacement of Vented Lead-Acid Batteries for Stationary Applications," of the following:

a.

Actions to restore battery cells with float voltage < 2.13 V, and

b.

Actions to equalize and test battery cells that had been discovered with electrolyte level below the top of the plates.

(continued)

Vogtle Units 1 and 2 5.5-18 Amendment No.

Amendment No.

(Unit 1)

(Unit 2)