ML20204E821

Safety Evaluation Supporting Amends 243 & 233 to Licenses DPR-77 & DPR-79,respectively

ML20204E821
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Site:	Sequoyah
Issue date:	03/16/1999
From:	NRC (Affiliation Not Assigned)
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ML20204E811	List: ML20204E805 ML20204E821
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NUDOCS 9903250143
Download: ML20204E821 (6)

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1 UNITED STATES g

j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 30886-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 243 TO FACILITY OPERATING LICENSE NO.

DPR-77 AND AMENDMENT NO. 233 TO FACILITY OPERATING LICENSE NO. DPR-79 4

TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT. UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328

1.0 INTRODUCTION

The Tennessee Valley Authority (TVA, the licensee) requested amendments to Operating Licenses DPR-77 and DPR-79 for Sequoyah Nuclear Plant (SON), Units 1 and 2, respectively, in a letter dated April 6,1995, as supplemented on August 21,1995. The amendments would revise the SON licenses by deleting License Conditions 2.C.(23)F for Unit 1 and 2.C.(16)g for Unit 2 which authorize the TVA to operate the SQN postaccident sampling system (PASS), as described in several TVA letters to the NRC referenced in the license conditions. TVA also proposed to revise the detailed description of the PASS procedures currently contained in the SQN Final Safety Analysis Report (FSAR) to supesede the license conditions being deleted. A markup of the revised post accident sampling (PAS) program provisions in the FSAR was included with the TVA submittal. TVA proposed to make future changes to the PAS program as contained in the FSAR under the 10 CFR 50.59 process. Also, TVA proposed to incorporate into the amended PASS specifications the changes which were either approved by the U.S.

Nuclear Regulatory Commission (NRC) for the advanced reactors, or were included in the topical report submitted by the Combustion Engineering Owners Group (CEOG) and approved by the NRC. In addition, TVA intends to modify PASS operating procedures by abandoning use of the on-line instrumentation and relying on grab sample analyses. It also proposes modifying the required accuracy of boron analysis.

2.0 BACKGROUND

The following changes to the PASS specifications were proposed by TVA:

1. Change of time requirement for measurement of boron concentration in reactor coolant

2. Change of time requirement for activity measurement in reactor coolant and containment atmosphere

3. Change of time requirement for measurement of dissolved gases in reactor coolant

4. Elirnination of hydrogen analysis in containment atmosphere

5. Elimination of oxygen analysis in reactor coolant and in containment atmosphere

6. Elimination of the pH measurement in reactor coolant and sump water 9903250143 990316 PDR ADOCK 05000327 p

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TVA proposes to revise SQN PAS program to clarify the sampling and analysis capabilities.

TVA's revised PAS program for SQN is summarized in.the April 6,1995, application. For comparison purposes, a summary of SQN's current PAS Program is also provided in the April 6,1995, application. The proposed changes continue to satisfy the PAS objectives of NUREG-0737, Item II.B.3 and Regulatory Guide (RG) 1.97 Revision 2. The purpose of revising l

the PAS Program is to:

l (a) Improve operational reliability of SON's PASS facilities, l

(b) Reduce maintenance associated with PASS on-line chemistry fastrumentation, (c) Utilize more reliable laboratory analysis methods, (d) Reduce PASS operator radiation doses, and (e) incorporate practical methods for meeting the objectives of regulatory requirements.

Under the revised PAS program, SON's PASS facilities will be dedicated for grab sample l

acquisition. The associated on-line chemistry instrumentation will no longer be maintained and i

utilized. Sample analysis reliability will be maintained with the use of laboratory chemistry I

instrumentation. Eliminating operation of PASS on-line chemistry instrumentation will reduce PAS operator radiation doses. Moreover, the on-line PASS chemistry instrumentation requires l

extensive maintenance and is expensive to replace.

The proposed changes include relaxation of pcstaccident sampling / analysis response times, exemption of some PASS parameters and process changes in the PAS program. The justification for relaxation and exemption of PAS program requirements are based on NRC policy issues for advanced light-water reactors described in NRC memorandum SECY-93-087 from James M. Taylor, NRC Executive Director for Operations, to the NRC Commissioners dated April 2,1993. The following discussion sets forth the NRC staff's evaluation of these changes, including a summary of the PAS criteria from NUREG-0737, Item ll.B.3.

3.0 EVALUATION 3.1 Revision of the Detailed Descriotion of PASS Proaram in the FSAR The PASS requirements in the above stated SON license conditions comply with the criteria specified in item II.B.3 of NUREG-0737 to which the licensee committed in its letters sent to NRC on November 23 and December 21,1983, and January 9 and 10 and March 23,1984.

In the submittal, the licensee requested a modification to PAS program, which is administratively controlled and governed by Administrative Section 6.8.4.e in the plant's Technbal Specifications (TSs). A detailed description of the PASS and supporting systems is contained in Sections 9.4.10 and 9.5.10 of the plant's FSAR. The PAS program modification is justified by the specifications described in Generic Letter (GL) 83-37 which addresses this issue, it is also in accordance with the format provided by the Standard Technical Specifications for Westinghouse plants referenced in NUREG-1431, Revision 1. Elimination of the subject license conditions will facilitate the licensee's introduction of future changes to the PASS procedures under 10 CFR 50.59 and facilitate implementing system upgrades as new technologies develop. The staff finds that this modification does not diminish the capability of PASS to monitor plant's parameters in the post-accident environment and it is, therefore, acceptable.

3 3.2 Chances in PASS SDecifications The PASS program, specified in Section ll.B.3 of NUREG-0737, requires certain samples be taken at the specified times following the accident. In SECY-93-087, dated April 2,1993, and in the accompanying Staff Requirements Memorandum, dated July 21,1993, the Commission approved for the advanced reactors a certain number of relaxations of these requirements. The licensee requested that these relaxations also be approved for the PASS in the SQN plant.

Although the Commission approved all the above relaxations specifically for the advanced plant designs described in the Advanceo Light Water Reactor Utility Document issued by the Electric Power Research Institute (EPRI), they would equally apply to the operating reactors. There is a very close similarity between the specific control and diagnostic parameters which are monitored in the postaccident environment of the advanced and the operating reactors. The methods for monitoring them should be identical and the Commission approved relaxations could be extended to the operating reactors without affecting the quality of PASS results.

The CEOG submitted to the NRC Topical CEN-415, Revision 1-A, " Modification of Post-Accident Sampling System Requirements" which included several proposed modifications in PASS requirements. In 1993, several of these modifications were approved by the NRC.

Although they were requested specifically for Combustion Engineering plants, they are applicable to all pressurized water reactors (PWRs).

The following relaxations, which were either approved by the Commission for the advanced reactors or by NRC in the CEOG's topical report, were reviewed by the stal 3.2.1 Chance of Time Reauirement for Measurement of Boron Concentration in the Reactg.r Coolant NUREG-0737 requires PASS to have the capability to sample and analyze boron concentration in reactor and/or sump water within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> following the accident. This information is needed to provide insight for accident mitigation measures in order to prevent criticality during a degraded core accident. However, in the SQN plant this information is provided by instrumentation having fully qualified, redundant channels that have capability to monitor power in the range from 104 to 200 percent power. Therefore, similar to the advanced reactors, sampling for boron concentration measurements will not be needed until 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after the accident. The chango of i

time requirement for measurement of boron concentration from 3 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> following the accident is acceptable.

3.2.2 Chance of Time Reauirement for Measurement of Dissolved Gases in the Reactor Coolant NUREG-0737 requires PASS to have the capability to analyze reactor coolant for dissolved gases within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> after the accident. The Commission concluded that although dissolved gas analysis in PWRs is still needed, the time for analyzing these gases could be extended to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the accident. The information on the amount of dissolved gas in the reactor coolant is an important factor in evaluating postaccident conditions existing in the reactor vessel.

However, because of the existence of certain other monitoring parameters, 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the accident is an adequate time for obtaining this information. Extension of time requirement 4

4 for measurement of dissolved gases in the reactor coolant from 3 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the accident is acceptable.

3.2.3 Elimination of Hydrooen Analvsis in the Con'airimen1/3mosphere NUREG-0737 requires PASS to take hydrogen samples of the containment atmosphere.

However, hydrogen analysis of the containment atmosphere can be accomplished by the safety-grade hydrogen monitor required by 10 CFR 50.34(f)(2)(xvii) and item II.F.1 of NUREG-0737.

Since this safety-grade instrumentation provides adequate capability for monitoring postaccident hydrogen, there is no need for having this function in PASS. Elimination of the PASS measurement of hydrogen in the containment atmosphere is, therefore, acceptable.

3.2.4 Elimination of Oxvoen Analysis and in the Containment Atmosohere and in the Reactor Coolant There is no specific requirement in NUREG-0737 for oxygen concentration measurement in the containment inmosphere and the reactor coolant. It is only recommoded in NUREG-0737 and included in Regulatory Guide 1.97. The information on oxygen concentration in the containment atmosphere is needed for determining its potential for supporting deflagration of combustible gases. However, this concentration can be fairly accurately estimated by indirect means. The information on dissolved oxygen concentration is needed to ascertain the degree of long-term corrosion. This concentration can be estimated with a sufficient degree of accuracy from the amount of oxygen present in the containment atmosphere. Elimination of these measurements is, therefore, acceptable.

2.2.5 Elimination of oH in the Reactor Coolant and Sumo ' Water i

Measurement of pH of the sump water is recommended but not required by NUREG-0737 and it is included in Regulatory Guide 1.97. This information is needed for controlling alkalinity of the sump water so that iodinc reevolution and cerrosion of metallic components is minimized.

However, experience has indicated that this alkalinity can be controlled to a high degree of accuracy by using different additives. This is especially true for the SQN plant wnere ice in the ice condenser contains sodium tetraborate which by buffering action will stabilize pH of the i

containment surnp water. Elimination of the sump we'er pH measurement is, therefore, acceptable.

3.3 Removal of On-line Instrumentation The on-line inst'rumentation in PASS serves to perform direct measurement of the process variables. However, it is difficult to maintain this instrumentation in owabla condition and its replacement by grab sampling will simplify PASS operation. Since there are not any specific requb ements for using on-line instrumentation in PASS and grab sampling can provide comparable results, replacement of on-line instrumentation by grab sampling is acceptable.

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3.4 Modification of Accurac" of Boron Analysis l

The current TS specify *6 percent for the accuracy of boron concentration measurements 4ampling and analysis)in the reactor coolant and the sump water. Because of the uncertainty associated with the sample aliquot valves used in obtaining diluted samples, the licensee finds that this accuracy cannot be met and propou to change it to *10 percent for the range of

. boron concentration of 500 to 6000 ppm and *50 ppm for the range of 50 to 500 ppm. Although this represents some decrease in the accuracy of boron cancentration measurement, the change is not large enough to cause a significant change and the amount of boron in the reactor j

coolant or sump water could still be obtained with a sufficient degree of precision. This modification of TS is, therefore, acceptab'e.

4.0 CONCLUSION

S The staff has evaluated ti'e licensee's proposed modifications to the Operating Licenses for Sequoyah, Units 1 and 2. The proposed modifications are in three areas: (a) the PAS program will be administratively controlled by 10 CFR 50.59, (b) several relaxations will be introduced to the original PAS program which were either approved by the Commission for the advanced reactors or constituted generic changes previously approved bv the NRC for other operating reactors, and (c) a few changes in operating piotedures will be made. Based on its evaluation, as discussed above, the NRC staff finds five of the six changes preposed by Tt '.to be acceptable (Changes 1,3,4,5, and 6 listed on Page 1 of this evaluation). Mc ng the PAS program administratively controlled is in compliance with the requirement 3 specified in GL 83-37 and is in accordance with the format of the Standard Technical Specifications for Westinghouse-designed reactors (NUREG-1431. The proposed relaxations of PASS operations requirements are justified because tha staff found that their previous approval for other plants can be extended to SQN and the proposed changes to operating procedures will not affect the quality of the results obtained by PASS. The modified PAS program will meet the requirements of 10 CFR 50.34(f)(2)(viii) and of General Design Criterion Q as it applies to instruments for monitoring variables over their anticipated ranges for accident conditions. No changes to the SQN TSs are required.

The sixth proposed change, a change of the time reqsiirement for activity measurement in reactor coolant and containment atmosphere from 3 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (Item 2 above), is being evaluated on a generic basis by the NRC staff and a staff position has not yet been developed.

The NRC staff is currently reviewing a Wolf Creek application dated November 10,1998, and Westinghouse Owners Group Topical Report WCAP-14986, Revision 1. " Westinghouse Owners Group Postaccident Sampling System Requirements: A Technical Basis," which proposes to eliminate these sampling requirements. Upon c'evelopr ent of a staff position and issuance of a Wolf Creek amendment by the NRC, TVA should resubmit its request for this (Item 2) proposal.

4.0 STATE CONSULTATION

in accordance with the Commission's regulations, the Tennessee State official was notified of the proposed issuance of the amendments. The State official had no comments.

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5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to installation or use of a facility I

component located within the restricted area as defined in 10 CFR Part 20 and changes surveillance requirements. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding l

that the amendment involves no significant hazards consideration, and there has been no l

public comment on such finding (60 FR 20527, dated April 26,1995). The r' oust 21,1995,

. letter provided clarifying information that did not change the scope of the a,;,c. I application or the proposed no significant hazards consideration determination. Accordingly, the amendments

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meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connaction with the issuance of the amendments.

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6.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) there onable assurance that the health and safety of the public will not be endangered by

'ation 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 se::urity or to :he health and safety of the public.

l Principal Contributor: Krzysztof I. Parczewski l

Dated: March 16,1999 i

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