ML20216D828

Responds to NRC 980206 RAI on Util 970915 License Amend Request to Revise TS Requirements for Reactor Vessel Pressure & Temp Limits

ML20216D828
Person / Time
Site:	Catawba
Issue date:	03/05/1998
From:	Gordon Peterson DUKE POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
TAC-M99629, TAC-M99630, NUDOCS 9803170285
Download: ML20216D828 (14)

Text

.. e-Duke Power Company ,

A Duke Emvp Compney 1 Catawba Nuclear Station ys-a .wa,c 4800 Concord Road hk, SC 29745 Gary R. kerson (BW) 831-4251 oma Vicelhsident - (803) 831-3426m l

l l

March 5, 1998 l 1

U. S. Nuclear Regulatory Commission Washington, DC 20555-0001 i

' ATTENTION: Document Control Desk

SUBJECT:

Duke Energy Corporation l Catawba Nuclear Station Units 1 and 2 Docket Nos. 50-413, 50-414 Response to NRC Request for Additional Information I on Technical Specifications Amendment Regarding Pressure / Temperature Limits TAC NOS. M99629 and M99630 By letter to the NRC dated September 15, 1997, Duke Energy Corporation (" Duke") submitted a license amendment request (LAR) to revise the Technical Specifications requirements for reactor vessel pressure and temperature limits, primarily Technical Specification 3.4.9.3 and Figures 3.4-2 and 3.4-3. By letter dated February 6, 1998, the NRC transmitted a request for additional information (RAI) on this LAR. The NRC's RAI contained six questions. Duke's responses to these six questions are contained in the attachment to this letter.

Additionally, on February 3, 1998, a conference call between an NRC Staff official and Duke representatives was held to  !

discuss the September 15 LAR submittal, and the forthcoming l February 6 RAI. During this conference call, the NRC Staff j official suggested that additional new Technical '

Specifications may be required at Catawba as a precondition l

to NRC approval of the revised pressure and temperature ,

limits. The NRC Staff official provided no basis for this j suggestion. Duke is very concerned about the position taken  !

by this NRC Staff official.

h.

Since this conference call with the NRC staff, Duke has reviewed the contents of the September 15 LAR. In Duke's view, the changes proposed in this LAR are conservative in nature and are consistent with the content of the current Catawba Technical Specifications. This LAR, in Duke's view, does not warrant the imposition of additional new Technical ,

h: l 9803170285 98u305 PDR ADOCK 05000413-  !

-.vv.v

~'

h

.4 -

U.'S. Nuclear ~ Regulatory Commission March 5, 1998 Page 2 Specifications. If the NRC staff takes a different view, please notify us immediately of the basis for that view.

Moreover, it is Duke's position that any non-conservative Technical Specifications or discrepancies believed to exist in the current Catawba Technical Specifications were not created by the proposed changes contained in the LAR presently undergoing NRC review. So, it is not reasonable to withhold' approval of the Sr tember 15 LAR pending an agreement to address such unrelated issues.

To the extent such perceived discrepancies may exist, these are being addressed through: 1) Catawba's May 27,:1997 LAR for conversion to the Improved Technical Specifications; and

2) another Technical Specifications amendment to which Duke committed in the corrective actions for Licensee Event Report No. 413/97-004. Both'of these amendments are

' discussed in Duke's response to NRC Question 1 contained in the. attachment to,this letter.

k For all the above reasons, timely NRC approval of this LAR should not be n.ade dependent upon prior or concurrent approval of any additional new Technical Specifications.for Catawba.

The service period of the existing heatup and cooldown curves contained in the current Catawba Technical Specifications expires after 10 EFPY. This expiration date occurs in less than 170 days for Catawba Unit 1, insufficient time to process any new proposed Technical Specifications. Additionally, Duke has previously requested that the NRC grant an implementation grace period of 60 days from the date of approval of this LAR. Duke has determined that the NRC's normal allowance of 30 days will not allow sufficient time to complete the document revisions necessary to implement this amendment in an orderly manner.

The above considerations make it imperative that the issues related_to this LAR be resolved without delay in order to support the continued operation of the Catawba units. We again urge the NRC to complete its review and subsequent approval of this LAR in a timely manner. To this end,'the appropriate Duke representatives are available to further

U .* S . Nuclear Regulatory Commission March 5, 1998 Page 3 discuss this matter with the NRC by means of a telephone conference call or a meeting in White Flint.

Inquiries on this matter should be directed to M. S. Kitlan, Jr. at (803) 831-3205.

Very truly yours,

~

D G. R. Peterson GRP/JSW Attachment l

~

es

• U .' S . Nuclear Regulatory Commission March 5, 1998 Page 4 xc w/att.:

L. A. Reyes U. S. Nuclear Regulatory Commission Regional Administrator, Region II Atlanta Federal Center 61 Forsyth St., SW, Suite 23T85 Atlanta, GA 30303 P. S. Tam NRC Senior Project Manager (CNS)

U. S. Nuclear Regulatory Commission Mail Stop O-14H25 Was'nington, DC 20555-0001 1

D. J. Roberts Senior Resident Inspector (CNS)

U. S. Nuclear Regulatory Commission Catawba Site l

t I

l  !

l i

1 i

L Attachment l

Duke Energy Corporation l Response to NRC Request for Additional Information

! Catawba Nuclear Station Units 1 and 2 l Proposed Amendment on Technical Specifications l

Pressure / Temperature Limits The Duke responses to the NRC's six additional questions are provided in the following paragraphs. Each NRC question is restated and shown in bold print.

1. Your current Technical Specifications (TS) allows one safety j injection pump and one centrifugal charging pump to be operable in Modes 4, 5, and 6 with the reactor vessel head installed. Your LTOP analysis accounts for injection from a l

cingle safety injaction pump. This is non-conservative from an LTOP perspective since you can potentially have injection l from both the safety injection pump and the charging pump.

) Provide analysis consistent with current TS, or justify the l discrepar.cy. In addition, identify your current TS

! restrictions on accumulator discharge valves and reactor coolant pump operation with respect to LTOPS.

This issue with the non-conservative TS was identified by plant

personnel and reported to the NRC (LER 413/97-004). Plant

! procedures were modified to ensure that current analysis and TS l surveillance 4.5.3.2 are fully satisfied. TS 4.5.3.2 specifically l requires:

l l All charging pumps and Safety Injection pumps, except l the above required OPERABLE centrifugal charging pump, L

I shall be demonstrated inoperable by verifying that the ,

motor circuit breakers are secured in th- t pen position I

[ or the discharge of each pump has been isolated from j the Reactor Coolant System by at least two isolation valves with power removec' from the valve motor operators at least once I~r 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> whenever the j temperature of one or more of the Reactor Coolant '

System cold legs is less than or equal to 285 F.  !

The footnote for TS 3.5.3.a (Page 3/4 5-8) that speaks to i allowing one Centrifugal Charging (NV) pump and one Safety '

Injection (NI) pump is in contrast to the required surveillance of 4.5.3.2. When this issue was identified steps were taken to ensure that surveillance 4.5.3.2 is met since it was the most conservative option. Additionally, the LTOPS analysis was 1

8

. Attachment reevaluated and concludes-that the multiple pump combination of l one'NV pump.and one NI pump is acceptable for LTOPS providing the '

Residual Heat Removal System (ND) suction valve (s) are credited  ;

for in LTOPS. As a result of the LER a TS amendment will be J developed, crediting the use of the ND suctlon relief valves to I provide LTOP protection, in order to allow the operability of a NI' pump during LTOP Modes.

- Catawba's Improved Technical Specifications (ITS) submittal, made by Duke letter dated May 27, 1997, resolves the issue with this note-that is in contrast to Surveillance requirement 4.5.3.2 requiring only one centrifugal charging pump. At this time Catawba restricts the number of pumps capable of injecting to the one (1) centrifugal charging pump during LTOP modes.

Notwithstanding these two LARs, described above, the September 15, 1997, submittal should.be considered on its own merits.

In its response to Generic Letter 88-17, Loss of Decay Heat Removal, dated January 3, 1989, Duke responded'to eight "

recomhr'.ded actions. Recommended action #6 of Generic Letter 88-17 states:

Provide at least two available or operable means of adding inventory to the RCS that are in addition to f pumps that are a part of the normal DHR system. These j should include at least one high pressure injection pump. The water addition rate capable of being provided by each of the means should be at least sufficient to keep the core covered. Procedures for use of these systems during loss of DHR events should be provided.  !

The path of water addition must be specified to assure the flow does not bypass the reactor vessel before exiting any opening in the RCS.

Duke's Response.was:

I To ensure required flow path availability during outages, several means of inventory makeup have been

. identified. At least two of these flow paths will be

" protected" to' ensure their availability during all phases of an outage.

I One high head injection pump is currently provided for as part of the TS required boration flow path in Mode 5 and 6.-The low head injection path consists of a driving head and inventory from the FWST, which is required by TS. The Loss of Residual Heat Removal i 2

)

4 8

. Attachment abnormal procedure provides guidance to the operator on the use of the available makeup flow paths. Further administrative' controls will be implemented to ensure the required flow paths are maintained while the unit 4 is in a reduced inventory condition.  !

Duke's response does not mention the additional NI pump being I needed te satisfy the requirements of Generic Letter 88-17.

Current TS do not imposa restrictions on the Cold Leg Accumulator )

discharge valves during LTOP modes. However, the ITS submittal does include this surveillance. ]

For Reactor Coolant System (NC) Pump operations, current Catawba i TS 3.4.1.3 and 3.4.1.4.1 impose the requirement: i A reactor coolant pump shall not be started with one or more of the Reactor Coolant System cold leg temperatures less than or equal to 285*F unless the secondary water temperature of each steam generator is less than 50 F above each of the Reactor Coolant System cold leg temperatures.

2. On page 4 of Attachment 3, you proposed to relocate the instrument uncertainty margin from the Catawba Technical Specifications to licensee-controlled procedures. It is the staff's position that Technical Specification set points (specifically the LTOP setpoints for this case) must account for instrument uncertainties. Please account for instrument l uncertainty and provide a calculation of the setpoints identifying all of the separate components l pressure / temperature (P/T) limit, analytical left set point, overshoot, etc.

The proposed amendment only removes instrument uncertainty from the Pressure / Temperature (P/T) curves published in the TS. This is consistent with the latest amended curves approved for use in the current Oconee and McGuire TS.

The instrument uncertainty applied to LTOPS is unaffected by this amendment.

3

Attachment

3. Provide an evaluation / analysis of the adequacy of the required vent area of 4.5 square inches (TS 4.4. 9.3) with the new limits.

Vent Paths (See UFSAR Fig 5-1 for valve locations) Area (square inch)

1. Head vent-paths through NC-298 and NC-299 4.524

2. Head vent paths through NC-250,251,232,253 0.35 or NC-223,224 and 225 open to the PRT

3. NC Safety Valves - when removed 21.10

4. NC PORV - when removed 5.40

5. NC PORV - blocked open 4.155 The LTOP analysis shows that one NC PORV is capable of mitigating a LTOP event. The area of a NC PORV is 4.155 square inches. Thus any area larger than this would'be sufficient for LTOP purposes.

The 4.5 square inches mentioned in TS conservatively requires a slightly larger area than the area of an open PORV.

Even though Catawba's LTOP analysis concludes that a single PORV can mitigaPe a LTOP event, the single active failure criteria still must be applied. Thus it is required that two (2) NC PORVs be fully-OPERABLE in LTOP modes. Of course, with valves removed or blocked open, the single active failure criteria would not apply.

l 4. On page 3 of Attachment 3 you stated "This evaluation demonstrated that the 285 *F enable temperature used for both units remains conservative relative to the RTNDT + 90 F l enable temperature." The enable temperature should account j for temperature instrument uncertainty and the temperature difference from the inside wall of the ressel to the quarter thickness location. Provide a calculation of the enable temperature, for each unit, which accounts for these effects.

The Low Temperature Overpressure Protection System is required to be operable at a temperature of RTmn + 90 degrees F (for the T location.in accordance with Standard Review Plan, Section 5.2.2

" Overpressure Protection", Rev. 2, November, 1988). Additional adjustments for instrument uncertainty and temperature differences between the inside wall of the vessel and the kT location are also taken into account.

l 4

, Attachment 4

For Catawba Unit 1, the limiting RTmn at 15 EFPY is 43 F, (WCAP i

- 13720, " Analysis of Capsule Y from the Duke Power Company ]

Catawba Unit 1 Reactor Vessel Radiation Surveillance Program, 1 June 1993," Table B-3). The Catawba instrumentation uncertainty )

calculation was determined to be i7.1 F. This calculation conservatively uses i10 F. A Westinghouse review of the available data resulted in the maximum delta-wall temperatures  !

(i.e., delta-T between the water temp. and crack tip). The kT from the inside su. face is 19.1 F. Thus, the LTOP system enable temperature is required to be 43 + 90 + 10 + 19.1 = 162.1 F.

Therefore, the Catawba Unit 1 LTOP Enable Temperature of 285 F for the new TS data would continue to be valid up to 15 EFPY.

Fo* Catawba Unit 2, the limiting RTmn at 15 EFPY is 112.6 F, (WCAP-13875, " Analysis of Capsule X from the Duke Power Company Catawba Unit 2 Reactor Vessel Radiation Surveillance Program, February 1994," Table 3-4). Using the same uncertainty value and the maximum Delta T between the surface of the vessel and the HT location, the LTOP system enable temperature is required to be 112.6 + 90 + 10 + 19.1 = 2s1.7 F. Therefore the Catawba Unit 2 LTOP Enable Temperature of 285 F for the new TS data would continue to be valid up to 15 EFPY.

5. Provide your LTOP analyses or a summary of the LTOP analyses, and the LTOP methodology for staff review.

l Purpose The purpose of the Duke LTOP analysis is to verify that the pressurizer (Pzr) power operated relief valve (PORV) setpoint is acceptable for preventing the violation of the 10CFR50 Appendix G pressure limitations (shown on Figures 3.4-2 and 3.4-4 of the Catawba TS) during pressure transients at low temperatures.

Three possible transients exist, a mass input from an operable NI ]

pump, a mass input from an NV pump, and a heat input from a 50 F ,

temperature difference between the steam generators (S/G) and NC.  !

These three transients are evaluated for a 3.0 and a 2.0 second  !

PORV opening time. I This calculation also evaluates:

1. the impact of the BWI replacement S/Gs; 5

, Attachment

2. the impact of the latest WCAP-13720 with the l BWI S/Gs (Unit 1);

3. the impact of the latest WCAP-13875 with the D5 S/Gs  !

(Unit 2);

4. the impact of ASME Code Case N-514 with data for WCAP-13875 (not yet included in Reg. Guide 1.147); and

5. the NC pressure response assuming operation of the ND suction line relief valves 1(2)ND3, 1 (2 ) ND38.

General LTOP Analytical Methodology Employed at Catawba:

The technique used in determining valve accumulation is presented in the Westinghouse document, " Pressure Mitigating System Transient Analysis Results," prepared by Westinghouse Electric Corporation for the Westinghouse Owners Group On Reactor Coolant System Overpressurization, July 1977, and " Supplement to the July 1977 2eport, Pressure Mitigating System Transient Analysis Results," prepared by Westinghouse Electric Corporation for the Westinghouse Owners Group On Reactor Coolant System j Overpressurization, September 1977. I I

The peak pressure (setpoint plus valve accumulation) is combined with corrections to convert indicated pressure and temperature to actual conditions, including instrumentation margins and calculated reactor vessel differential pressure. Both the Pzr PORVs and the ND System suction line relief valves (not yet credited for LTOP protection) are evaluated in response to bounding mass and energy addition transients. Also, corrections for S/G tube plugging are considered. The PORV setpoint is verified acceptable by comparison to the P/T curves. Acceptable heatup and cooldown limits are determined based on comparison of j peak pressure and P/T curve limits. l The acceptability of the Pzr PORV LTOP setpoint is verified as follows:

1. The NC peak pressure is calculated using the previously mentioned methodology for limiting transients (inadvertent l start of a NI pump, charging letdown mismatch, and overheating from a S/G). This is done for both the DS S/Gs l (Unit 2) and the new BWI S/Gs (Unit 1). This calculation is performed assuming both a 3.0 and a 2.0 second PORV opening time.

2. The difference between the indicated pressure (signal actuating Pzr PORV's) and the actual reactor vessel beltline pressure is calculated. This includes clevation differences between the reactor vessel beltline and the NC pressure 6

s 4

, Attachment transmitters, and the differential pressure across the I reactor core due to hydraulic losses with varying number of NC pumps operating (the transmitters actuating the PORV's are on the hot legs, while'the reactor vessel beltline is on the cold leg side of the core).

3. The impact of S/G tube plugging on the peak pressure is evaluated. (Not a significant contributor.)

4. The instrument error for the pressure and temperature instrumentation associated with LTOP is.' input from the uncertainty calculation.

5. The information gathered in steps 1, 2, _3, and 4 are then combined for the peak vessel pressure at the critical weld

' location.

6. The ASME Section III, Appendix G heatup and cooldown limits (calculated by Westinghouse) .are adjusted for temperature uncertainty and extrapolated to the boltup temperature.

7. The peak pressures in step 5.are then compared to the-l adjusted Appendix G limits of step 6 above.

Assumptions Employed

1. The Pzr is water solid.

2. NC is at a temperature cf 100 F for the purpose of calculating friction factors.

?.

3. The AP across the RV is the same at hot or cold temperatures.  ;

4. It is anticipated that instrumentation will be re-calibrated following an earthquake. But since the Branch Technical Position RSB 5-2, requires the overpressure protection systems to function during an " Operating Basis Earthquake" ,

the Post-Seismic instrumentation uncertainty analysis is used.

1 l

5. NC volume for Catawba 2 is taken as 12,000 ft3 and the S/G I h'eatLtransfer area is taken as 48,000 ft2'. The volume closely _ approximates the actual value (12,040 ft3) and the results of the calculations are insensitive to small deviations. For Catawba 1 the actual volume is 13,050 ft3 l i

7 l

'i l .

, ' Attachment l

(the BWI S/Gs assume 13,000 ft3 ) with 79,800 ft2 heat transfer surface. area.

t

6. The In7' System Positive. Displacement Pump is considered as p .being not capable of' injecting into'NC.

L

7. The NC loop flow increases 11% in the active loop with one (1) NC Pump operating due to.back flow through .the inactive

[ loops. It.is also assumed that for 2 NC pumps the flow increase is'7.3% and for 3 RCP'the flow increase is 3.7%

8 ', Unit 2 parameters are calculated at an (old) S/G heat transfer surface area of 48,000 ft2, even though D5 S/Gs have a surface area of 48,165 ft2. The difference in the-heat transfer surface area between the d S/G (48,000 f t 2) and the DS S/G (48,165 f t2 ) is insignificant since .the

! ' calculation is insensitive to small deviations.

9. The Pzr PORV LTOP setpoint is 400 psig (currently controlled administratively) .

10. The BWI S/Gs are assumed to have the same thermal conductivity cs the original D3 S/Gs. In reality, the BMI L S/Gs use alloy 690 which has a thermal conductivity lower l -than that of the original tube material. This assumption

! will make the results conservative for the Heat Input

! Transient results.

l

11. For the purpose of proceduralizing the NC Pump start process with respect to LTOP, steady state temperature is defi'ed as AT/At < 10 F/hr, where T is the maximum variation of NC temperature (T) in the previous one hour. This definition is consistent with that of ASME Section XI Article E-1000 for steady state conditions.

12. The steady state Appendix G limits are used for the verification of acceptability of the Pzr PORV and ND suction relief valves for LTOP protection in situations where multiple combinations of NI and NV pumps are capable of i injecting into~NC. )

13. The. actual Appendix ~G Heatup, Cooldown and Steady State limits (depending on plant status) are used to verify acceptability of a Pzr'PORV to, mitigate 1the consequences of 1 a single NI or NV pump or secondary side heat input LTOPS' ^

transient..The use'of non-steady-state Appendix G curves 8 l

.. . 1

, Attachment provides additional conservatism since LTOP type events are basically acknowledged as being a steady-state event.

14. The NI and NV pump flow rates have been calculated based on a conservative set of operating parameters. The flowrates selected for LTOP analysis are based on th~ flowrate of the specific pump at the LTOP relief valve setpoint (i.e., 400 psig). The' limiting flow rate trom a single NI pump is 550 gpm. The limiting flow rate from a single NV pump is 475 gpm.

15. The instrumentation uncertainty determined in accordance with the Duke uncertainty analysis is 151.8 psi and i7.1 F.

The LTOP calculation conservatively used 155 psi and i10 F.

These values were calculated using the guidance provided by "ISA-S67.04, Part I, Setpoints for Nuclear Safety-Related Instrumentation, Dated 1994," and "ISA-RP67.04, Part II, Methodologies for the Determination of Setpoints for Nuclear Safety Related Instrumentation, Approved September 1994." A

' noteworthy point: these values were found to be very similar to the original Westinghouse supplied _ uncertainty numbers of f 60 psi and 110 F.

6. Provide and justify your minimum temperature for boltup/ pressurization with respect to LTOP (i.e. show that the lower and temperature at which the system can potentially pressurize is bounded by your LTOP analysis). This must also l account for instrument uncertainty.

The boltup temperature used at Catawba Nuclear Station is 60 F

! and is conservatively applied to both units. Unit 1 would not be

{

restricted based on actual material properties which result in a difference in their initial RTyy (Unit 1 Closure Head Flange = -4 F, Vessel Flange = -31 F, Nozzle Shell = -13"F from TS Bases Table B 3/4.4-1) (Unit 2 Closure Head Flange = 10 F, Vessel Flange = 10 F, Nozzle Shell = 50 F) . The verification of the minimum temperature is performed locally on the flange region prior to tensioning or detensioning of the reactor head.  !

Additionally, the operators in the control room have remote i instruments to use to determine NC temperatures. The uncertainty discussed in Question 4 concerning tne LTOP temperature uncertainty, is conservatively assumed to be 10 F. Therefore, 9

m

, Attachment with the highest of the above mentioned initial RT,y of 50 *F, thus the boltup temperature becomes 50*F + .10 F = 60 F.

The current.and prcposed TS P/T curves do not provide data down to the bolt-up temperature. The WCAPs (WCAP - 13720, " Analysis of Capsule Y from the Duke Power Company Catawba Unit 1 Reactor Vessel Radiation. Surveillance Program, June 1993," and WCAP-13875, " Analysis of Capsule X from the Duke Power Company Catawba Unit ~2 Reactor Vessel Radiation Surveillance Program, February 1994") also do not provide data down to the bolt-up temperature.

.Since:the data supplied from Westinghouse do not continue down to the bolt up temperature, the data have been extrapolated to 60 F. .This extrapolation is based on additional information previously provided by Westinghouse, and recently reconfirmed by Westinghouse letter No. NSD-E-MSI-98-045, dated February 23, 1 1998. The extrapolation is calculated based upon using the last I two Westinghouse data points (flange limit data points excluded) and this provides a conservative plot. Westinghouse has agreed that future WCAP capsule analyses will contain the necessary information, .such that subsequent TS P/T curves will continue down to the bolt up temperature.

.The extrapolated plots are provided to Catawba operations.

personnel in the form of a composite graph.in the plant's Data .I Book, which is located in the control room. The operators are procedurally required to use these plots during heatups'and

.cooldowns. This composite graph provides such information as-the LTOP setpoint, the instrument heatup and cooldown limits (the P/T curves extrapolated to the bolt-up temperature), NC pump number

1. 1' delta P requirement, NC pump NPSH, saturation curves, the L minimum temperature for boltup, etc.). Upon implementation of the i L new P/T curves, this graph will also display the uncertainty l

values'that are being proposed for removal from the TS.

With incorporation of the proposed P/T curves, there will be a I need to require that the minimum temperature during LTOP mode be no lower than 66 F on Unit 1 and 65 F on Unit 2. These numbers are based on the peak pressure from the limiting steady state j condition mass addition transient from a NI pump, with no NC l pumps running. j 1

A review of past-refueling outage data that were takt during NC refill (low decay heat mode), shows that NC temperature is typically operated between 95 F and 115 F.

1 10