ML18018B426

Forwards Responses to Draft SER Open Items Re Use of Cracs Code,Min Yield Strength Specified for Austenitic Stainless Steel in RCPB & Underclad Cracking in carbon-manganese Steel

ML18018B426
Person / Time
Site:	Harris
Issue date:	10/27/1983
From:	Mcduffie M CAROLINA POWER & LIGHT CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
References
LAP-83-507, NUDOCS 8311020111
Download: ML18018B426 (29)

Text

REGUI.ATORY IORNATION DISTRIBUTION SYS'+ (RIBS)

ACCESSION NBR:8311020111 DOCB DATE: 83/10/?7 NOTARIZED: NO DOCKET FACIL;50-400 Shearon Harris Nuclear Power Pl anti Uni,t 50-001 Shearon Har ris Nuclear Power Planti Unit ii Car olina 2~ Carolina 05000400 05000001 AUTH, NAME AUTHOR AFFILIATION MCDUFFIEiM~ A, Carolina f ower 8 Light Co, RECIP ~ NAME RECIPIENT AFF II I ATION DENTONgH,RB Office. of Nuclear Reactor Regulationi Director

SUBJECT:

Forwards responses to draft SER open items re use of CRACS codeimin yield strength .specified for austenitic stainl s steel in RCPB 8 underclad cracking in Carbon"Manganese Steel.

DISTRIBUTION CODE: 8001S COPIES RECEIVED:LTR ENCL ~ZE:

TITLE: 'Licensing Submittal: PSAR/FSAR Amdts L e ate Correspondence' NOTES:

RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL NRA/DL/ADL 0 NRR LB3 BC 0 NRR LB3 LA 0 BUCKLEY'S 01 1 1 INTERNALS ELD/HDS1 1 0 IE FILE 1 1 IE/DEPER/EPB 36 3 3 IE/DEPER/IRB 35 1 1 IE/DEQA/QAB 21 1 1 NRR/DE/AEAB 1 0 NRR/DE/CEB 11 NRR/DE/EHEB 1 1 NRR/DE/EQB 15 2 2 NRR/DE/GB 28 2 2 NRA/DE/MEB 18 1 NRR/DE/MTEB 17 1 1 NRA/DE/SAB 29 1 1 NRR/DE/SGEB 25 1 1 NRA/DHFS/HFEBQO 1 NRR/DHFS/LQB 32 i NRR/DHFS/PSRB 1 1 NRR/DL/SSPB 1 0 NRA/DSI/AEB 26 1 1 NRR/DSI /ASB 1 1 NRA/DSI/CPB 10 NRR/DS I/CSB 09 1 i NRR/DSI/ICSB 16 1 1 NRR/DSI/METB 12 1 1 NRA/DSI/PSB 1 8 22 1 1 NRR/DS I/RSB 23 1 1 EG F L 00 1 1 RGN2 3 RM AM I/MI8 1 0 EXTERNAL; ACRS Q] 6 BNL(AMDTS ONLY) 1 1 DMB/DSS (AMDTS) 1 1 FEMA REP DIV S9 1 1 LPDR 03 i 1 NRC PDR 02 1 1 NSLC 05 1 1 NTIS 1 1 TOTAL NUMBER OF COPIES REQUIRED'TTR 53 ENCL

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ClPQP, SERIAL: LAP-83-507 Carolina Power 8 Light Company QQf 87 1983 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation United States Nuclear Regulatory Commission Washington, DC 20555 SHEARON HARRIS NUCLEAR POWER PLANT UNIT NOS. 1 AND 2 DOCKET NOS. 50-400 AND 50-401 RESPONSES TO REQUESTS FOR ADDITIONAL INFORMATION

Dear Mr. Denton:

Carolina Power & Light Company hereby transmits one original and forty copies of additional information requested by the NRC as part of the safety review of the Shearon Harris Nuclear Power Plant. The cover sheet of the attachment summarizes the related Open Items addressed in the attachment along with the corresponding review branch and reviewer for each response.

We will be providing responses to other requests for additional information shortly.

Yours very truly, M. A. McDuffie Senior Vice President Nuclear Generation JHE/mf (8332JHE)

Enclosures cc: Mr. B.C. Buckley (NRC) Mr. Wells Eddleman Mr. G.F. Maxwell (NRC-SHNPP) Dr. Phyllis Lotchin Mr. J. P. O'Reilly (NRC-RII) Mr. John D. Runkle Mr. Travis Payne (KUDZU) Dr. Richard D. Wilson Mr. Daniel F. Read (CHANGE/ELP) Mr. G. 0. Bright (ASLB)

Mr. R. P. Gruber (NCUC) Dr. J. H. Carpenter (ASLB)

Chapel Hill Public Library J. L. Kelley (ASLB)

Wake County Public Library 8311020111 831027

=PDR ADOCK 05000400 '

PDR 411 Fayetteville Street ~ P. O. Box 1551 o Raleigh, N. C. 27602

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ATTACHMENT LIST OF OPEN ITEMS/NEW ISSUES, REVIEW BRANCH AND REVIEWER Auxiliary Systems Branch/N. Wagner Open Items 370(3), 372(1), 372(9)

Materials Engineering Branch/D. Smith Open Items 324, 328 Power Systems Branch/E. Tomlinson Open Item 106 Radiological Assessment Branch/S. Block Open Item 173 Reactor Systems Branch/E. Marinos Open Items 212, 214 Structural Engineering Branch/S. Kim Open Item 335

Shearon Harris Nuclear Power Plant DSER Open Item 370(3)

Supplemental Information Show how the CRACS complies with the guidelines of positions C.4.a and C.4.d of Regulatory Guide 1.95.

Clarification The NRC has requested that CP&L provide additional information to demonstrate that the CRACS purge makeup and purge exhaust systems meet the single failure criteria.

Response

FSAR Figure 9.4. 1. 1 has been revised to reflect the current design of the CRACS. The revised figure shows the purge makeup and purge exhaust systems with isolation valves in series. Purge makeup valves 3CZ-B17SA-1, 3CZ-B18SB-1, and 3CZ-B17SA-2, 3CZ-B18SB-2, Units 1 and 2 respectively, are powered from the "A" and "B" trains respectively. Purge exhaust valves 3CZ-B13SA-1, 3CZ-B14SB-1 and 3CZ-B13SA-2, 3CZ-B14SB-2, Uni.ts 1 and 2 respectively, are powered from the "A" and "B" trains respectively. Thus, the failure of one isolation valve in the open position would not preclude isolation of the Control Room.

Shearon Harris Nuclear Power Plant SER Open Item 372, Part 1 (ASB Review Question 10.4.9(1))

Can the AFW motor-driven pumps deliver 475 gpm with their r'ecirculation line(s) open, at the setpoint pressure?

Response

With the AFW pump recirculation line open, one motor-driven AFW pump has the capacity to deliver approximately 440 gpm to the steam generators at the setpoint pressure. As described in FSAR Section 10.4.9.1, 475 gpm is required only during a loss of main feedwater. During this scenario, two motor-driven auxiliary feedwater pumps can provide more than the required 475 gpm with their recirculation lines open. If one of the two motor-driven auxiliary feedwater pumps is unavailable due to a loss of voltage, the motor-operated valve on the recirculation line of the other motor-driven auxiliary feedwater pump will automatically close. The recirculation lines automatically isolate upon loss of power to their redundant pumps as described in FSAR Section 10.4.9.2.3. In addition, the recirculation valves can be remote-manually closed from either the Control Room or the auxiliary control panel. With the recirculation line secured, the unaffected motor'-driven AFW pump has the capability to deliver approximately 490 gpm to the steam generators at the setpoint pressure.

Shearon Harris Nuclear Power Plant DSER Open Item No. 372, Part 9 (ASB Review Question 10.4.9(9)

If the single valve in the supply line from the CST fails and causes loss of water supply to the AFW pumps, at what setpoint will an alarm be provided for low pump suction pressure, and at what setpoint with there be an alarm and pump trip or Lo-Lo pump suction? If the pumps are reset to start after pump trip and switchover to SW, are the pumps provided with sufficient margin to prevent damage by cavitation?

Response

The postulation of a normally locked-open, admnistratively controlled, manual valve to fail and block flow to the AFW is not considered to be a credible event. If, however, the supply of the CST to the AFW pumps is blocked, pressure switches are provided on the AFW pumps (motor-, turbine-driven) suction side to provide Lo alarm at approximately 16 psig and Lo-Lo alarm and pump trip at approximately 6 psig. As shown in Figure 7. 3. 1-9 and 7.3. 1-10, alarm indication is provided for operator action.

The supply line is normally water solid and will provide static head to the pump suction. Loss of water (i.e., blocked flow from valve) will cause low pump suction pressure and the pump(s) will trip prior to loss of NPSH.

Switchover to ESW will provide water to the pump supply header and maintain solid supply water.

Shearon Harris Nuclear Power Plant DSER Open Item 324 (SRQ 252.1)

Is there a maximum yield strength specified for austenitic stainless steel used in the Reactor Coolant Pressure Boundary (RCPB)? (SRP 5.2.3)

Response

SRP 5.2.3;Section II. ACCEPTANCE CRITERIA, Part 4.b. (Fabrication...of Austenitic Stainless Steel), fourth paragraph on Page 5.2.3-9 states the following:

Laboratory stress corrosion tests and service experience provide the basis for the criterion that cold-worked austenitic stainless steels used in the reactor coolant pressure boundary should have an upper limit on the yield strength of 90,000 psi. (Applicable to material reviewed by MTEB in item I.4.b)

Note that the SRP refers to cold-worked austenitic stainless steel which is not used in the RCPB. The RCPB uses austenitic stainless or carbon steel with stainless cladding. Therefore an upper limit on yield strength is not applicable to these components.

Shearon Harris Nuclear Power Plant DSER 328 (SRQ 252.5)

Supplemental Information Provide additional information concerning underclad cracking in carbon-manganese steel pressure vessels.

Response

The Shearon Harris Nuclear Power Plant complies with the recommendations of Regulatory Guide 1.43 as discussed in Section 1.8 of the SHNPP FSAR. To ensure guide compliance, welding processes that induce underclad cracking by generating excessive heating and promoting grain coarsening in the base metal are not used.

The reactor vessel heads and shell courses were constructed of SA-533 Grade B Class 1 plate material made to a fine grain practice. The closure head and vessel flanges and the primary nozzles were constructed of SA-508 Class 2 forging material. This plate and forging material was clad utilizing the shielded metal are and the two-wire submerged arc processes which are con-sidered low heat input processes. Since the plate material and the low heat input clad processes used on forging material are not subject to restrictions by the guide, the vessel is in compliance with Regulatory Position C.l.

Regulatory Position C.2 is not applicable in this case. The reactor vessel fabricator monitored and recorded the weld parameters to verify compliance with the parameters established by the procedure qualifications of Regulatory Position C.3.

The steam generator and the pressurizer parts which are clad are constructed of SA-533 Grade A Class 2 and SA-508 Class 2a steels. These materials are made to fine grain practice and welding is done with low heat input techniques.

Shearon Harris Nuclear Power Plant DSER Open Item 106 Supplemental Response Provide seismically designed emergency lighting in areas required to ensure safe shutdown.

Response

SHNPP is designed such that the plant can be brought to, and maintained in, a safe shutdown condition from the Main Control Room (MCR). The normal emergency lighting in the MCR is seismically designed to assure sufficient lighting is available for manual operator actions following a SSE. In addition, adequate lighting is provided throughout the plant to meet the requirements of 10 CFR50, Appendix R.

The source of power for the Normal Emergency Lighting system is Class 1E and can be powered from the Station Diesel Generators. In addition, the portion of the Normal Emergency Lighting system that feeds the MCR is redundant and seismically designed. This includes the cables, conduits, and light fixtures.

Seismic adequacy is provided for the lighting panels by their similarity to seismically qualified panels utilized in the Class 1E MCCs. These panels are fabricated by the same vendor and from observation and inspection are the same as the panels in the 1E MCCs. The Control Room lighting panels are seismi-cally wall mounted and would experience less dynamic acceleration than. the similar panels within the lE MCCs.

Seismic adequacy will be provided for the Control Room lighting transformers by test and/or analysis. The transformers are seismically floor mounted.

The above measures are more than satisfactory to ensure the Control Room lighting remains functional following a SSE event.

Shearon Harris Nuclear Power Plant (SHNPP)

Draft Safety Evaluation Report (DSER)

Radiological Assessment Branch Open Item 173 (DSER Section 12.5.1, page 12-13)

The Harris plant organization chart shows the Environmental and Radiation Control Manager reporting to the Manager, Plant Operations. It is the staff's position that in matters relating to radiological health and safety, the individual responsible for the radiation protection support group's activities has direct responsibility to both employees and management. This responsi-if bility can best be fulfilled the Environmental and Radiation Control Manager not only has access to the General Plant Manager, but also is indepen-dent of the plant operations management, whose prime responsibility is conti-nuity or improvement of station operability. This apparent lack of indepen-dence is also an open item.

Response

When the first CP&L nuclear plant began operation, the plant organization consisted of a Plant Superintendent with an Operating Supervisor, Maintenance Supervisor, and an Engineering Supervisor reporting to him. As the individual roles of operations, maintenance, and engineering personnel became more complex and demanding, additional personnel were added to the plant staffs.

As the plant staff evolved, so too did the management structure, in order to maintain control of the many staff activities. In addition to the changes in management structure, the management positions have gradually been upgraded to provide the necessary authority and responsibility required to perform the respective management fuctions. The position of Plant General Manager has replaced the previous postion of Plant Superintendent. The positions of nager Maintenance, Manager Operations, and Manager Technical Support have replaced the previous positions of Maintenance Supervisor, Operating Supervisor, and Engineering Supervisor respectively. With the increased emphasis CP&L has placed on ALARA, the position directly responsible for Radiation Protection and Control has been upgraded from a Foreman to the Manager Environmental & Radiation Control. While the Plant General Manager has ultimate responsibility for overall plant operations, a new position, Manager Plant Operations, was created to meet the managerial demands of day-to-day plant operations. This position has the same authority'and responsi-bility for the safe operation of the plant as did the previous Plant Superintendent. The positions of Manager Environmental & Radiati.on Control, Manager Maintenance, and Manager Operations report to the Manager Plant Operations. The objective of having these three disciplines work as an integral unit is to develop a close coordination of operations and maintenance activities with a constant emphasis on the Health Physics and ALARA programs. This objective has been realized with great success.

.The Manager Environmental and Radiation Control is the station's Radiation Protection Manager and as such has the responsibility and authority to report to the General Manager on any aspect of the radiation potection program or its implementation as appropriate. As indicated in the draft Technical Specifications, Section 6, Figure 6.2. 1-2, Onsite Organization (attached), the

173 Page 2 nager Environmental & Radiation Control has direct and open lines of communication with the Plant General Manager. The Manager Environmental and Radiation Control also has lines of communication with the Manager Corporate Health Physics and the Manager Radiological & Chemical Support for matters regarding the radiological health arid safety of employees and the public. The Manager Environmental & Radiation Control is charged with the responsibility and is encouraged to use these lines of communications to ensure that the com-pany policy for Health Physics and ALARA are implemented. A figure reflecting these lines of communication will be added to FSAR Section 13. 1.2 in a future amendment.

Carolina Rwer & Light Company feels that the current organization meets the i.ntent of NRC's guidance, since the Manager E&RC does not report to the Manager Operations or Manager Maintenance but reports to the next level of management, the Manager Plant Operations. This organization includes the additional check and balance of encouraging the Manager E&RC to communicate directly to the Plant General Manager and Corporate Managers on Health Physics matters and unresolved concerns.

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Shearon Harris Nuclear Power Plant DSER Open Item 212 Question 440.83 (Section 15.3.31)

Revised Response The analyses of a locked reactor coolant pump rotor and a sheared reactor pump shaft in Section 15.3 of the FSAR assumes the availability of offsite power throughout the event. In accordance with Standard review Plant 15.3.3, 15.3.4, and GDC 17, we require that this be analyzed assuming turbine trip and consequential loss of offsite power to the plant auxiliaries and resulting coastdown of all undamaged pumps. Appropriate delay times may be assumed for loss of offsite power if suitably justified.

The event should also be analyzed asuming the worst single failure of a safety system active component. Maximum technical specification primary system activity and steam generator tube leakage at the rate specified in the Technical Specifications should be assumed. The results of the analyses should demonstrate that offsite doses following the accident are less than 10 CFR100 guideline values.

Response

The following is a revised response to Open Item 212 concerning the locked rotor/break incident. This response supercedes the previous response dated July 15, 1983.

Accident Scenario: The locked rotor followed by a loss of offsite power transient is postulated to occur in the following manner.

1. Reactor coolant pump rotor locks (or shears) and flow in that loop begins to coast down.

2. The reactor is tripped on low RCS flow in one loop.

3. Turbine/generator trips.

4. Offsite power is lost.

5. The loss of offsite power causes the two remaining reactor coolant pumps to coast down.

Methods of Analysis: The method of analysis used is the same as the cases presented in Section 15.3.3. (The worst single failure for this analysis is the loss of one protection train. This failure has no impact, since there are two trains.) The following is analyzed: three loops operating, one rotor locks/shears, followed by a coastdown of the other two reactor coolant pumps.

Results: The calculated sequence of events for the locked rotor/shaft break without offsite power is shown in Table 1. The results of temperature, flow, heat flux, and clad temperature are shown in Figures I-1 through I-7. These may be compared with the results of the transient shown in Section 15.3. As can be seen from the figures, the case without offsite power results in an

212 Page 2 increase of aproximately 20 F clad temperature and 30 psi reactor coolant pressure. This reflects the effects of reduced flow due to the loss of power, since the peak values occur after the coastdown begins.

==

Conclusion:==

The transient presented here is the most limiting for the locked of the rods.

The radiological doses have been calculated and found to be within the guidelines values of 10 CFR100. The doses are presented in Table 2.

212 Page 3 TABLE 1 Sequence of Events Locked Rotor Without Offsite Power Event Time (Seconds)

Rotor on one pump locks 0. 00 Low RCS flow trip setpoint reached 0.07 Rods begin to drop 1. 07 maximum RCS pressure occurs 2. 80 Remaining reactor coolant pumps begin coastdown 3.07 Maximum clad temperature occurs 4.01

212 Page 4 TABLE 2 Radiological Consequences of Locked Reactor Coolant Pump Rotor Without Offsite Power Doses (Rem)

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Shearon Harris Nuclear Power Plant Draft SER Open Item 214 Supplemental Information Provide additional clarification of the low temperature overpressure protection system design basis and setpoints.

Response

As discussed in the CP&L-NRC Reactor Systems Branch meeting of September 22, 1983, and stated in FSAR Section 5.2.2. 11. 1, the design basis mass input event for the low temperature overpressure protection system (LTOP) is caused by the inadvertent actuation of a single safety injection pump. A single safety injection pump will deliver approximately 650 gpm with reactor coolant system fully depressurized. The setpoints for the LTOP are determined such that the reactor vessel's Appendix G curve is not exceeded. These setpoints will be based on the design basis event caused by the inadvertent actuation of a single safety injection pump. These setpoints will be submitted for NRC revie~ with the submittal of proposed Technical Specifications presently scheduled for mid-1984.

Shearon Harris Nuclear Ebwer Plant Draft SER Open Item 335 Supplemental Information .

The reviewer requested verification that there is no spring disengagement in the one-way spring model used for the bedrock representation by the STARDYNE code in the stability analysis. In addition, the reviewer requested that copies of the time history plots for the reactor auxiliary building be included in the response.

Response

The factor of safety against overturning was computed by dividing the resisting moments by the overturning moments, as stated in FSAR Section 3.8.5.5.

The overturning moment for Shearon Harris was calculated by the square root of the sum of the squares (SRSS) of the horizontal acceleration only in both the E-W and N-S directions. The vertical seismic overturning moment was added directly, as 0.4 times the actual moment.

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A study was also conducted for Shearon Harris to compute the overturning moment by the SRSS method by taking the square root of the sum of the square of the overturning moments due to the horizontal acceleration in both the E-W and N-S direction, plus the overturning moments due to the vertical seismic uplift forces.

2 2 2 M MEW+MNS+M (2)

The resisting moment is the vertical load of the building and mat (reduced by buoyancy) multiplied by the perpendicular distance to the axis of rotation through the corner of the building.

The factor of safety against overturning computed by both the methods described above resulted in higher factors of safety than the minimum required by FSAR Section 3.8.5.5. The factor of safety for the RAB Unit 1 calculated in accordance with (1) above is 1.25 and with method (2) is 1.34. Both are higher than the acceptable limit of 1.1.

Further details on the determination of the stability of the Reactor Auxiliary Building against overturning are given below.

335 Page 2 The Ebasco computer program Dynamic 2037 was used to obtain time histories of moments and vertical forces. These time histories, and the dead loads and buoyancies in effect for the building, were then used as input for the Ebasco computer program THPLOT 2524 to obtain time history plots of eccentricity (the location of the point of application of resultant load at the foundation level due to earthquake loads). The plots are given in Figures OI-335-1 and OI-335-2, both dated October 18, 1983, and entitled, respectively, "SHNPP RAB DBE EW

& VERT., F'C ~ 4000 PSI, RIGID FDN METHOD ECCENTRICITY-TIME HISTORY CURVE FOR MAT," and "SHNPP RAB DBE NS 6 VERT., F'C = 4000 PSI RIGID FDN METHOD ECCENTRICITY-TIME HISTORY CURVE FOR MAT" (see attachments).*

A line was drawn on each plot to show the initial eccentricity e1, the distance of the center of gravity of the building loads from the neutral axis of the mat, for the axis covered in the figure. A second set of lines, one for each direction of motion about the axis covered in the figure, was added to the plots showing the maximum eccentricity e2 from the neutral axis line at which the mat would still retain 100% contact with the supporting media below.

The peaks that cross the 100% contact lines locate the times when the mat is not in 100% contact with the supporting media. The percentage of mat still in contact was calculated for the peaks of maximum extension beyond the 100%

contact lines, and the values for percent of mat remaining in contact are given at those locations.

• Due to their size, the drawings are being sent only to the reviewer.