ML20236Y347

Advises That Util Will Conduct Heatup of Both Primary & Secondary Plant,Utilizing Nonnuclear Heat Sources Before Completion of All Restart Activities.Addl Info Re Resolution of Three Technical Evaluation Programs & Risk Analysis Encl

ML20236Y347
Person / Time
Site:	Sequoyah
Issue date:	12/04/1987
From:	Gridley R TENNESSEE VALLEY AUTHORITY
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
TAC-R00253, TAC-R253, NUDOCS 8712110259
Download: ML20236Y347 (13)

Text

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TENNESSEE VALLEY AUTHORITY CH ATTANOOGA. TENNESSEE 3740*

SN 157B Lookout Place l DEC 041987 l

l U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Hashington, D.C. 20555 Gentlemen:

In the Matter of ) Docket No. 50-328 Tennessee Valley Authority )

SEQUOYAH NUCLEAR PLANT (SQN) UNIT 2 - NONNUCLEAR HEATUP OF SQN UNIT 2 As part of TVA's planned activities leading to the restart of SQN unit 2, TVA will conduct a heatup of both the primary and secondary plant, utilizing nonnuclear heat sources, before completion of all restart activities. These plans have been previously discussed with NRC management and were. subsequently documented by letter from S. A. White to J. G. Keppler dated September 28, 1987. This activity will involve heating up the Reactor Coolant System (RCS) using normal operating procedures to rated temperature and pressure, i.e.,

547 degrees F and 2,235 psig. It is anticipated that this heatup will be conducted over a duration of five to six weeks. During this period, TVA will perform a " shakedown" of plant equipment and systems to ensure system integrity and functionality following the 2-year shutdown of the unit.  !

Equipment and systems will be operated, tested, and inspected. Upon completion, the unit will be returned to the cold shutdown condition for completion of outstanding restart activities and resolution of any problems encountered during heatup and cooldown of the primary and secondary plant.

As a prerequisite to this activity, TVA is evaluating known restart deficiencies and issues and defining the work required to be completed before .

entry into modes 4 and 3 in order to satisfy technical specification (TS)  !

operability requirements. This work will be completed before heatup. Known deficiencies associated with restart activities that remain outstanding at the time cf heatup will therefore only be those that do not compromise TS operability requirements. Completion and closure of restart activities will continue throughout this heatup. Of special interest are three major technical evaluttion programs (Integrated Design Inspection, silicone rubber cables, and pipe support calculation regeneration) that have generated broad issues and may have associated efforts continuing beyond heatup. Technical resolution of these issues for heatup has been achieved. The primary purpose of this submittal is to provide status for TVA's resolution of these technical evaluation program issues and to provide bases for our conclusion that operability of TS equipment required for heatup will not be compromised by any issues associated with these programs.

8712110259 871204 PDR ADOCK 05000328 h'

$i P PDR t An Equal opportunity Employer

i U.S. Nuclear Regulatory. Commission 04 BN It is recognized that there is the potential for deficiencies .to be identified-as a result of continuing technical reviews or from other inspection and testing activities that will be taking place during the heatup. Any deficiencies identified during this heatup will be evaluated for impact on TS

( operability requirements, and the associated limiting condition for operations will be complied with as required.

i For this specific SQN unit 2 heatup, the potential safety significance of-any identified deficiencies is less than for a normal heatup/ restart, primarily because of the specific core burnup, reduced fission product inventory, reduced decay heat load resulting from the extended shutdown, and the

/ , limitations of a nonnuclear heatup. TVA is providing in enclosure 2 a radiological and , risk analysis performed by TVA for a large break loss of coolant accident while SQN unit 2 is in mode 3 to bound'the potential consequences of an accident at SQN during this heatup. To provide further conservatism, a dose analysis was performed assuming no containment isolation function (i.e., ground release to environment with no holdup or filtration) and credit was only taken for systems required to provide cooling to the core (i.e., emergency core cooling water, component cooling water, and essential raw cooling water systems). On the basis of this analysis, it is clearly shown that the planned heatup of SQN unit 2 does not present any undue risk to the health and safety of the public.

Enclosure 3 lists the commitments made in enclosure 1.

If you have any questions, please telephone M. R. Harding at 615/870-6422.

Very truly yours, TENNESSEE VALLEY AUTHORITY R.Gridley,D]ifect Nuclear Licensing and Regulatory Affairs Enclosures cc: See page 3

U.S. Nuclear Regulatory Commission D .0 041987 cc (Enclosures):

Mr. G. G. Zech, Assistant Director for Inspection Programs Office of Special Projects U.S. Nuclear Regulatory Commission 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323 Mr. J. A. Zwolinski, Assistant Director for Projects Division of TVA Projects Office of Special Projects U.S. Nuclear Regulatory Commission 4350 East-West Highway EWW 322 Bethesda, Maryland 20814 Sequoyab Resident Inspector Sequoyah Nuclear Plant 2600 Igou Ferry Road Soddy Daisy, Tennessee 37379

ENCLOSURE 1 RESOLUTION STATUS OF THREE TECHNICAL EVALUATION PROGRAMS AND BASES FOR DETERMINING TECHNICAL SPECIFICATION (TS) COMPLIANCE FOR HEATUP 0F SEQUOYAH NUCLEAR PLANT (SQN) UNIT 2 I. INTRODUCTION Before restart of SQN unit 2 and before completion of all outstanding restart activities, TVA will conduct a heatup of the primary and secondary plant utilizing nonnuclear heat sources. This will involve heating up the Reactor Coolant System (RCS) to rated temperature and pressure, i.e., 547 degrees F and 2,235 psig.

The anticipated duration of this heatup is five to six weeks. During this heatup period, TVA will inspect and operate equipment idled during the 2-year shutdown of the unit. This shakedown will serve to ensure system integrity and functionality before the return to power operation of SQN unit 2. Upon completion, SQN unit 2 will be returned to cold shutdown for the scheduled closeout of any outstanding restart activities and the resolution of any problems encountered during this heatup of the primary and secondary plant.

To support this mode transition, TVA is evaluating known restart deficiencies and issues and defining the work required to be completed before entry into modes 4 and 3 in order to satisfy TS operability requirements. This work will be completed before heatup. Known deficiencies associated with restart activities that remain outstanding at the time of heatup will only be those that do not compromise TS operability requirements. Included within this evaluation effort are three major technical evaluation programs: the Integrated Design Inspection (IDI) of the essential raw cooling. water (ERCW) system, the silicone rubber cable evaluation being performed by TVA, and the TVA pipe support calculation regeneration effort. The status of resolution of issues associated with these technical evaluation programs and the basis for SQN unit 2 TS compliance as related to these issues are as follows.

II. TECHNICAL EVALUATION PROGRAM STATUS AND BASIS FOR TS OPERABILITY FOR HEATUP OF SQN UNIT 2 A. Pipe Support Calculation Regeneration Program The Pipe Support Calculation Regeneration Program was initiated following a TVA review that identified that a significant number of the original design calculations were not retrievable. The program L _ _ _ _ . _ _ . _ _ _ _ _ _

ENCLOSURE 1 description was submitted for NRC review by letter from R. Gridley to NRC dated August 21,1987, "Sequoyah Nuclear Plant (SQN) - Unit 2 Regeneration Of Pipe Support Calculations On Rigorous Analysis Piping."

The scope of this effort involves approximately 5,790 pipe supports on the rigorous analysis category I piping. Before program initiation, TVA developed comprehensive design criteria (SQN-DC-V-24.2) that meet the requirements of the American National Standards Institute (ANSI) 831.1, 1967 edition, and accordingly comply with the SQN Final Safety Analysis Report (FSAR). These design criteria include supplemental allowance for friction and proper allowable stresses for the normal, upset, and faulted analysis conditions. In order to provide additional confidence in the end product, functional verification activities (walkdowns) confirmed the installed configurations. The functional verification results were then utilized for the calculatic,n regeneration program. Results of this program indicated that approximately 5,160 of the 5,790 supports fully complied with Design Criteria SQN-DC-V-24.2; accordingly 630 did not. The following provides a summary of the evaluation of those supports that di'i not meet SQN-DC-V-24.2. The support numbers cited reflect evaluations completed to date; finalization of reviews may alter the numbers slightly.

The 630 pipe supports that were not found to be in full compliance with SQN-DC-V-24.2 were further reviewed against the unit 2 pipe support modifications restart criteria (CEB-CI-21.89) for prioritization of support modifications. Compliance with these restart criteria would ensure that pipe supports will remain operable under all design conditions even though their design does not fully meet the requirements of SQN-DC-V-24.2. CEB-CI-21.89 was formally I submitted for NRC review by letter from R. Gridley to NRC dated August 31, 1987, "Sequoyah Nuclear Plant (SQN) - Uni t 2 Support Modification Restart Criteria For Rigorous Analysis Piping"; and TVA presented the criteria to NRC in a meeting on September 1, 1987.

Results of the meeting are provided in the NRC Meeting Summary dated September 4, 1987. As a result of this review, it was determined  !

that approximately 290 of the 630 support modifications could be deferred until postrestart based on compliance with the restart {

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criteria. Further discussions of these criteria were held between '

TVA and NRC on October 29-30, 1987; the agreements reached for clarification / revision to CEB-CI-21.89 were subsequently documented by letter from R. Gridley to NRC dated November 17, 1987, "Sequoyah i Nuclear Plant (SQN) - Unit 2 Support Modification Restart Criteria Supplemental Revision." Review of the remaining 340 supports (630-290) against the revised criteria determined that an additional 220 support modifications could be deferred to postrestart; it should be noted, however, that approximately 80 of these 220 postrestart support modifications had already been ccmpleted before the October 29-30, 1987 discussion. Although 120 additicinal supports l

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i ENCLOSURE 1 (340-220) were therefore identified as restart modification candidates, inorarability of individual supports (failure of design to meet prescriN:d criteria) does not necessarily affect piping system operability. However, to confirm the modes 4 and 3 TS system

, operability bases, the approximately 120 restart pipe support modification will be completed before heatup of SQN unit 2, resulting in operability of all supports. The modifications will bring the associated supports into compliance with SQN-DC-V-24.2.

In summary, as a result of the functional verification walkdown effort, the review of supports against approved criteria, and the implementation of modifications as described above, it is determined that pipe supports on rigorously analyzed piping systems are now or will be before heatup fully operable througt compliance with either the pipe support design criteria SQN-0C-V-24.2 or restart criteria CEE-CI-21-89. Accordingly, outstanding support modifications will not adversely affect the ability of associated systems to perform their required safety function. Compliance to SQN-DC-V-24.2 for the approximately 430 support modifications deferred to postrestart is targeted to be achieved by completion of unit 2 fuel cycle 4.

B. Silicone Rubber Cable Evaluation Program Silicone rubber-insulated cables are used in 10 CFR 50.49 circuits in SQN unit 2. The cables are used in 120-volt control and 480-volt power applications and are manufactured by American Insulated Wire (AIW), Anaconda, and Rockbestos. A total of approximately 980 conductors is used in 50.49 circuits.

In order to address a special concern regarding unsupported vertical cable drops, TVA performed in situ dc high potential (hipot) testing in the worst-case silicone rubber cable installation. The test voltage was 10,800 volts dc based on the nominal thickness of the insulation. The first test of 16 conductors resulted in three voltage breakdowns and one conductor that exhibited increasing current leakage. Upon isolation of the problem, it was determined that none of these problems were because of the vertical support concern. Further, TVA removed the suspect conductors, replaced them with new conductors, and contracted with the University of Connecticut (UCONN) to provide TVA with additional laboratory tests. l UCONN tests found no evidence of mechanical degradation, no evidence of contamination, and that the test failure was localized. The damaged areas were examined and found to have as little as 8 mils of insulation remaining intact. A third party consultant with extensive silicone rubber insulation experience observed that the damages resembled those for impact.

At this time, TVA continued the dc hipot testing of 75 additional conductors. The test voltage was 240 volts de times the minimum environmentally qualified thickness, i.e., 240 volts times 40 mils l i

i Enclosure 1 equaling 9,600 volts. Of the 75 condur.tcrs tested, four voltage breakdowns occurred; and two conductors had a polarization index of less than 1.0.

J The resultant concern from this test experience was whether a silicone rubber-insulated cable with a remaining wall thickness of as little es 8 mils could perform its safety function when exposed to '

accident conditions. To resolve this concern, TVA contracted Wyle Laboratories to determine if cables with lower wall thicknesses are environmentally qualified for accident application. The test specimens included AIW, Anaconda, and Rockbestos cables with remaining insulation thicknesses of 2, 1, and 4 mils respectively.

The application of these cables is such that no significant internal heating because of current flow occurs during normal operation. j Therefore, they remain essentially at amoient temperatures. By applying the Arrhenius methodology to the SON specific application and additionally considering the inherent prcperties of silicone rubber cables, no pretest thermal aging was therefore required. This is consistent with practices as allowed by industry standards. The testing was specified to use industry standards (Institute of Electrical & Electronic Engineers 383-1974) and the actual SQN worst-case accident profile with margin. All of the cables were irradiated to the equivalent of 10-year normal service plus full mode 1 accident dose of radiation before testing.

All tested cables performed their IE function, that is, carried their rated voltage and current throughout the accident portion of the test. Postaccident margin testing involved wet and dry 500-volt dc insulation resistance test. In addition, the cables were wrapped oh a 40-times-the-cable outside diameter mandrel and were subjected to a dc hicot withstand voltage test greatly in excess of industry standard for the thickness, both dry and immersed in tap water. The dc withstand test voltages included one 5-minute step at 240 volts per mil of the measured thickness plus 1- or 5-minute stops as required, at 500, 1,000, 2,000, 3,000, 4,000, and 4,700 volts. ,

The 4,700-volt test correlates to twice-rated service voltage plus j 1,000 times 2.4, which is similar to test voltages specified in l ANSI / National Electrical Manufacturers Association MG-1 and ANSI i C37.90.

One sample of Anaconda cable with a remaining wall thickness of 1 mil performed its design function throughout the loss of coolant accident i (LOCA) testing but showed no margin. This is not significant because '

this thickness is far below the minimum thickness discovered installed in SQN unit 2 and the test voltage applied is in excess of that required for service. This portion of the cable was removed; and the post-LOCA margin assessment dielectric test continued with the remainder of the sample, which had 2-mil thick insulation. This and all remaining samples passed the aforementioned voltage withstand  ;

test.

Enclosure 1

-S-TVA has evaluated the status of the silicone rubber cable testing issue and determined that operability of 50.49 equipment is not compromised by the evidence obtained. The in situ tests conducted to date subjected the cables to voltages far in excess of what they will experience in normal operation or accident conditions; failure of '

some cables to meet the acceptance criteria specified for the in situ tests does not correlate to cable failure as a result of postaccident duty and environment. Conservative high-voltage in situ testing was conducted without information regarding the effect of accident and postaccident conditions on reduced insulation thicknesses. Actual envirc wental testing was subsequently performed by Wyle to directly and deh nitively determine the effects of reduced insulation thickness. Initial review of the test results confirms that cables with reduced insulation thicknesses as low as 2 mils are capable of performing the intended postaccident safety function, and reduced thicknesses of 2 mils are capable of withstanding margin tests. On this basis, we conclude that operability of associated SQN 50.49 equipment will not be compromised.

C. IDI of the ERCW NRC, in reviewing the work that TVA had performed and was performing to ensure the safety of SON before restart, conducted an IDI of the ERCH to provide additional assurance that major problems have been identified and resolved before restart. As a result of this inspection, NRC identified a number of findings, some being specific 1 deficiencies and others being issues requiring further evaluation and 1 analysis by TVA. TVA evaluated all open findings against the SON restart critoria to determine which issues must be resolved before restart. NRC issued under letter from James G. Keppler to S. A. White dated October 9,1987 . their preliminary assessment of l IDI findings requiring restart resolution; TVA responded in an October 29, 1987 letter from S. A. White to Ja;nes G. Keppler.

Subsequent to this correspondence, NRC conducted a reinspection over the period of November 2-10, 1987, on TVA's assessment of and .

corrective actions for the restart issues. Based on that  !

reinspection, TVA and NRC reached agreement on actions required to resolve the restart findings, with the exception of three that are still under evaluation by NRC. These restart actions will be documented in TVA's response to NRC's IDI report, 50-327 -328/87-48, transmitted by letter frc,m Stewart D. Ebneter to S. A. White dated November 6, 1987.

To support the planned heatup of SQN unit 2, TVA conducted an analysis of open (corrective actions not complete) restart IDI findings to determine if any of these findings affect modes 4 or 3 TS operability requirements. Documentation from these reviews are available in TVA's offices. Based on this review, TVA has determined that the following corrective actions must be completed before heatup.

Relief valves will be installed on ERCW lines serving the auxiliary control air compressors. I 1

Enclosure 1 l'

TVA will either verify to be acceptable or modify / replace, if required, existing ERCH flood-mode spoolpieces to ensure the spoolpieces can be installed.

TVA will replace the control power protection fuses on the traveling screen motor control center transformer and the

" isolation" fuses on the speed indication switch in order to ensure adequate isolation between IE and non-lE equipment.

If continuing analysis or review of any IDI finding results in potential TS operability concerns, TS limiting conditions for operation and action statements will be complied with as required.

Based on the above-described review, evaluation, and corrective actions, it is our conclusion that the IDI issues have been or are being resolved such that heatup of SQN unit 2 does not represent any undue risk to plant operation or the health and safety of the public. IDI issues have been thoroughly evaluated, and resolution actions are well defined. A limited number of issues could affect operability of modes 4 or 3 TS equipment; and for those issues, corrective actions have been or will be completed before heatup.

III. CONCLUSION Before restart of SQN unit 2, TVA will conduct a heatup of the primary and secondary plant utilizing nonnuclear heat sources. To support the heatup, TVA is evaluating known restart deficiencies and issues and defining the work required to be completed before entry into modes 4 and 3 in order to satisfy TS operability requirements. As part of this evaluation, TVA has reviewed the issues associated with three major technical evaluation programs--pipe supports, silicone rubber cables, and the IDI of ERCW--and concludes that resolution status for the programs ensures that operability of TS equipment is not compromised for the planned heatup. With regard to the pipe support calculation regeneration effort, calculations have been completed; and affected pipe supports are now or will be fully operable in accordance with prescribed criteria before heatup. Conservative in situ testing and subsequent Wyle environmental testing of silicone rubber cables have demonstrated that S0.49 silicone rubber cables with reduced insulation thicknesses are capable of performing their required safety function under postaccident conditions and are therefore fully operable. IDI issues have been thoroughly evaluated, and resolution actions are well defined. A limited number of issues could affect operability of modes 4 or 3 TS equipment:  ;

but for those few, prescribed corrective actions have been or will be '

completed before heatup.

Enclosure 1 From this summary and from the details provided previously in this enclosure, it is shown that the described resolution to the issues associated with these three evaluation programs ensures that operability of TS equipment will not be compromised during the planned heatup of  !

SQN. Activities associated with f'inal closure of these programs will I continue but do not affect current operability requirements. l l

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l ENCLOSURE 2 RADIOLOGICAL AND RISK ANALYSIS I, INTRODUCTION Before restart of SQN unit 2 to full power operation following a 2-year shutdown, TVA will conduct a heatup of the unit from mode 5 cold shutdown conditions to mode 3 hot standby conditions using nonnuclear heat sources. This will involve heating up the RCS to rated temperature and j pressure, i.e., 547 degrees F and 2,235 psig. Inspection and testing will be conducted during this heatup, and the unit will then be returned to cold shutdown conditions to complete restart activities required for nuclear restart of the unit. The expected duration of the heatup is five to six weeks.

To provide additional confidence in the safety of this special heatup, TVA has performed a conservative radiological and risk assessment. This quantifies the reduced risk for a design basis accident (large-break LOCA) postulated to occur during mode 3 conditions for the existing l

Unit 2 core. The radiological risk during this heatup following two years of core fission product decay is much less severe than that described in the SQN FSAR Chapter 15 accident analysis.

II. RADIOLOGICAL AND RISK ANALYSIS  ;

A. Radiological Dose Assessment

" analysis to determine the heatup of the fuel was performed by TVA assuming a large-break LOCA occurred while the plant was in mode 3.

It was conservatively assumed that,no cooling of the fuel would occur I

from the time of the LOCA until emergency core cooling water system (ECCS) water would completely recover the fuel. The results of the analysis showed that the fuel heatup rate was very low, and no damage to the fuel would be expected.

Even though no fuel damage would be expected, a conservative dose analysis was completed to bound the consequences of a mode 3 l

large-break LOCA. The analysis only takes credit for sys;3ms required to maintain core water inventory (i.e., ECCS, ERCW, and component cooling water system).

The analysis assumed core operation for 1,000 effective full-power days followed by 24 months of fission product decay. This assumption bounds the conditions of the unit 2 core at SQN.

A review of Chapter 15 events confirmed that a large break LOCA is the most limiting dose consequence event for the expected SQN unit 2 mode 3 condition. As such, a large-break LOCA was postulated to occur that resulted in the immediate release of 100 percent of the noble gases and 50 percent of the lodines from the core inventory to the containment; 50 percent of the lodines released to containment is

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ENCLOSURE 2 assumed available for leakage to the environment. These release l fractions are consistent with the SQN FSAR Chapter 15 assumptions for i the design basis accident. For additional conservatism in this analysis, a ground-level release with no holdup or filtration was assumed to occur from the containment to the environment. The results of this analysis case will therefore envelope the results of analysis assuming containment isolation occurs as designed.

The analysis for this postulated accident was performed using approved in-house computer codes previously utilized for calculating offsite doses for Chapter 15 of the SQN FSAR. Results for this postulated accident indicate that, even with these conservative assumptions, the 30-day dose at the outer boundary of the Low Population Zone (LPZ) would be less than 1 rem whole body and negligible for inhalation. In addition, the site boundary 2-hour dose is predicted to be less than 2 rem whole body with a negligible inhalation dose. Assuming containment isolation does occur, analysis yields LPZ boundary and site boundary doses of less than one millirem whole body with negligible inhalation dose. These doses are significantly less than those reported for the design bases for the LOCA in Chapter 15 of the SQN FSAR and considerably less than the limits in 10 CFR 100. It is therefore concluded that a large-break LOCA for SQN unit 2 during mode 3 does not constitute a significant radiological risk to the public.

B. Probabilistic Risk Assessment The probability of a safe shutdown earthquake (SSE) at SQN is estimated as 6 x 10-' per year. The probability of an SSE during this 5- to 6-week interval is 7 x 10-5 The probability of a ,

large-break LOCA at SQN taken from the Individual Plant Evaluation 1 (transmitted from S. A. White to S. D. Ebneter on March 18, 1987) is ,

estimated as 3 x 10** per year. The probability of LOCA during this 5- to 6-week interval is 4 x 10'5 The probability of a coincident SSE and LOCA during this 5- to 6-week interval is estimated at 2.8 x 10-5 .

III. CONCLUSION The probability of an SSE and/or a LOCA during the anticipated interval in question is extremely low, and the radiological consequences of the limiting accident--a large-break LOCA--occurring at SQN unit 2 while in mode 3 represent no undue risk to the health and safety of the public.

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ENCLOSURE 3 COMMITMENTS

1. Relief valves will be installed on ERCH lines serving the auxiliary control air compressors before heatup.

2. Before heatup, TVA will either verify to be acceptable or modify / replace, if required, existing ERCW flood-mode spoolpieces to er.sure all spoolpieces can be properly installed.

3. Before heatup, TVA will replace the control power protection fuses on the traveling screen motor control center transformer and the isolation fuses !

on the speed indication switch in order to ensure adequate isolation between IE and non-1E equipment.

4. Approximately 120 additional restart pipe support modifications will be completed before heatup of SQN unit 2 (80 'lready completed plus 120 additional, for a total of 200). The modi ications will bring the associated tupports into compliance with SwN-DC-V-24.2.

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