ML19257B293

Forwards Initial Questions Re Facility Fsar.Questions Cover Sections 15.0-15.6,4.6,3.13 & 3.2.Requests Assistance in Obtaining Answers from Applicant

ML19257B293
Person / Time
Site:	Sequoyah
Issue date:	01/09/1980
From:	Taylor W E.I. DU PONT DE NEMOURS & CO., INC.
To:	Speis T Office of Nuclear Reactor Regulation
References
NUDOCS 8001150513
Download: ML19257B293 (27)

Text

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UCC: M. J. Sires, III, DOE E. l. ou PONT DE NEMouns & COMPANY J. D. Bilycu

~ ~ *'*= F. E. Krucsi, Wilmington ATOMIC ENERGY DIVIStoN J. E. Conaway, Wilmington SAVANNAH RIVER PLANT J. T. Granaghan - T. IIendrick AtMEN. south CARoLIN A 29808 A. E. Hadden

................,n........................... D C. Owen

!!. R. Reeve January 4, 1980 Revised January 9, 1980 Mr. Themis Speis, Chief Reactor Systems Branch Mail Stop 268 Phillips Building U.S. Nuclear Regulatory Commission Bethesda, MD 21609

Dear Mr. Speis:

Attached are initial questions relating to the Susquehanna Nuclear Station FSAR. The questions cover Sections 15.0, 15.1, 15.2, 15.3, 15.4, 15.5, 15.6, 4.6, 3.13, and 3.2.

Your assistance is requested in obtaining responses from the applicant.

Yours very truly,

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W. M. Taylor Superintendent E & I Department WMT:ssr ATT 1747 034 Xco3

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C211. Provide a rea?.istic range and permitted operating band for the (15.0) exposure dependent parameters in Tables 4.4-1 and 15.0-2. In SSES Table 15.0-2, provide assurance that values of parameters selected 1 yield the most conservative results.

Q211. Uncertainty exists on the correct value of APRM neutron flux scram (15.0) setpoint to be used in transient analyses. The value indicated as SSES input for transient analysis in Table 15.0-2 is 125% NBR.

2 However, a value of 120% NBR is indicated in Tables 7.2-4 and 7.6-5. Explain this discrepancy. For the correct value of setpoint used in transient analyses, provide a breakdown of any uncertainty allowances that are added to the nominal value.

Q211. Provide a listing of the transients ard accidents in Chapter 15 (15.0) for which operator action is required in order to mitigate the SS E.P consequences.

3 Q211. The response to question 211.113 does not provide sufficient (15.0) detail on non-safety grade equipnent and components which SSES mitigate transients and accidents. Provide a table of the non-4 safety grade equipment and components assumed to mitigate consequences for each transient and accident in Chapter 15.

Q211. The analysis of transients and accidents in Chapter 15.0 does (15.0) not state which of the RPS time response delays in Table 7.2-5 SSES is used in the REDY computer model (NEDO-10802). For each 5 transient and accident in Chapter 15.0, specify whether the sensor 1747 035

or overall delay time is used in the analysis and why the specified delay time is conservative.

Q211. Confirm the following items for all transients in Chapter 15.0 (15.0) which require control rod incertion to prevent or lessen plant SSES d am age .

6 a) All calculations were performed with the conservative scram reactivity curve No. 2 in Figure 15.0-2.

b) The slowest allowable scram insertion speed was used.

Q211. a) In Table 1 of Figure 5.1-3a (Nuclear Boiler), the relief (15.0) valve spring set pressure of 1130 psig for safety / relief SSES valves B and E does not agree with a corresponding value of 7 1146 psig in Table 5.2-2 of the FSAR and in Table 1 of Drawing M-141, Rev. 9. Correct this setpoint discrepancy for safety mode (mechanical) actuation.

b) For transient analysis, credit has been taken for safety /

relief valve actuation in the relief mode. A more conservative approach would be to take credit for s fe';/

relief valve actuation in the safety mode, resulting in higher peak vessel pressures.

1) What effect on MCPR and peak vessel pressure does credit for safety / relief valve actuation in the safety mode have on transients analyzed in Chapter 15?

2) Are all equipnent and components required for safety /

relief valve actuation in the relief mode safety grade?

Q211 Modify Table 15.0-1 as follows:

(15. 0 ) a) Give calculated values of MCPR instead of the entry >1.06.

SSES b) For the "feedwater controller failure at maximum demand" 8 transient, correct the discrepancy in values -ir maximum vessel pressure, maximum steam line pressure, and MCPR that exists between Table 15.0-1 and Section 15.1.2.3./.47 1 036

Q211. For transiencs and accidents in Chapter 15 in which it is stated (15.0) that the operator initiates some corrective action, provide SSES justification for any corrective actions by the operator prior to 9 20 minutes.

Q211. Discuss how the pre-operational and startup tests will be used to (15.0) confirm flow parameters used in Chapter 15 analyses. Provide SSES details of any previous test of components in test facilities 10 conducted to show satisfactory performance of the recirculation and feedwater flow control systems and respective pumps. Describe how this information was used in Chapter 15 analyses.

Q211. Analyze the turbine trip and generator load rejection transient (15.0) from a safe shutdown earthquake event. Credit should not be SSES taken for non-seismically qualified equipment which include:

11 a) Any equipnent contained in a non-seisnic structure; and b) Any equipment which is not seismically qualified.

0211. On page 4-7 of NED0-10802, it is stated that the difference in (15. 0 ) trend of flow coastdown versus initial rower between the SSES analytical and experimental coastdown curves for Dresden Unit 12 No. 2 (a EWR/3) in Figure 4-11 was due in part to differences between actual and computed jet pump efficiencies, a) How has this effect been treated in analysis of SSES transients involving flow coastdown with two recirculation pump trip (RPT)?

b) Is this treatment applicable to Susquehanna which is a EWR/47 1747 037

J211. For 'be " loss of feedwater heating" transient, the sequence of (15.1.1.2.1) es -nts a- Tables 15.1-1 and 15.1-2 f:r the automatic and manual SSES flow control modes, respectively, are not described in sufficient 13 detail to permit evaluation of transient results in Figures 15.1-1 and 15.1-2 and comparison with NSOA events in Figure 15A.6-21.

For the more limiting manual flow control mode, no detail is presented in Table 15.1-2 between 2 and 40 plus seconds. Revise Table 15.1-2 to include NSOA events in Figure 15A.6-21 and additional detail between 2 and 40 plus seconds.

Q211. The thermal power monitor (TPM) is not included in the Susquehanna (15.1.1.2.3) design per response to question 211.118. However, it is SSES indicated as the primary protection system for mitigating the 14 consequences of the " loss of feedwater heating" transient in Section 15.1.1.2.2. What was used to scram the reactor in the manual mode? Modify Figure 15A.6-21 and Sections 15.1.1. 2. 2 and 15.1.1.2.3 accord ingly .

Q211. This section states that input parameters and initial plant (15.1.1.3.2) conditions for the " loss of feedwater heating" transient are in SSES Table 15.0-1. This should be changed to Table 15.0-2 in this 15 section and in the corresponding sections of the remaining transients in Chapter 15 where this discrepancy occurs.

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0211. Correct discrepancies between events in Table 15.1-3 and NSOA (15.1.2.?.1) Figure 15A.6-22 for the "feedwater controller failure at maximum SSES demand" transient . Table 15.1-3 does not include the initial core 16 cooling and reactor vessel isolation events indicated in Figure 15A.6-22.

Q211. Explain the basis for the assumed feedwater flow controller

( 15.1. 2. 3.1 ) failure at 135% flew. Is the indicated failure initiated at 0 SSES secorils or does the failure begin at 0 seconds and increase to 17 135% flow at a later time. If the former is true, correct Figure 15.1-3 accordingly.

C211 Please correct the inadvertent combination of Section 15. 1.2. 3. 2,

( 15.1. 2. 3.1 ) beginning on page 15.1-7, with Section 15.1.2.3.1 SSES 18 Q211. Provide justification that analysis of this transient at 105%

(15.1.2.3 3) NBR steam flow is more restrictive than at low power. If so, SSES delete reference to " low power" for NSOA event No. 22 in Table 19 15A.2-2. If not, re-analyze and make appropriate corrections.

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Q211. a) It is not app 3 rent from the text whether the " pressure (15.1.3.3.3) regulator failure-open" transient is terminated by a low SSES turbine-inlet pressure trip or a L8 trip. Trips indicated in 20 various sections of the text are summarized below:

Section Trip 15.1.3.2.1.1 Low pressure at the turbine inlet 15.1.3.3.2 Low pressure at the turbine inlet 15.1.3.3.3 L8 trip Table 15.1-4 Low pressure at the turbine inlet Specify which trip is most restrictive on thermal margins and revise applicaole tables, sections, and figures of the FSAR.

b) It appears that less than the assumed 115% NBR steam flow in Section 15.1.3.3.2 was simulated at the beginning of the transient in Figure 15.1-4 Explain this discrepancy and make corrections, if necessary.

c) Safety / relief valve (SRV) actuation for this transient in the relief mode is not included in Tables 15.0-1 and 15.1 4 and Figure 15.1-4 for decay heat removal. Please explain.

Q211 In Table 15.1-4, (15.1.3.2.1) a) Include safety / relief valve actuation times for the " pressure SSES regulator failure-open" transient.

21 b) Indicate the value of steam flow simulated at time = 0, presumably 115% NBR per Section 15.1.3.3.2, 1747 040

Q211. Specify the assumed operating mode (manual or automatic) of the (15.1.3.3.2) recirculation riew control system for the " pressure regulator SSES failure-open" transient and prov ide justification that the most 22 conservative results on core thermal margins are obtained with the assumed operating mode.

1747 041

Q211. A qualitative presentation of results for the- " inadvertent (15.1.4.3.1) safety / relief valve opening" transient is given because analyses SSES from earlier FSAR's indicated this event is not limiting from a 23 thermal margin standpoint.

a) Provide supoorting data that justifies this condition (i.e.,

referenced plant and FEPR).

b) The discussion in Section 15.1.4.3.2 implies a quantitative analysis was made. A statement similar to that in Section 15.2.1.3.2 would be more appropriate.

1747 042

9 Q211 For the " pressure regulator failure-closed" transient, correct the (15.2.1.2.1) discrepancy that exists between the 5 psi setpoint difference for SSES the backup pressure regulator in Sections 15.2.1.1.1 and 24 15.2.1.2.1 and a corresponding 10 psi setpoint difference in Section 10 3.2.

Q211. It is stated that the pressure diaturbance in the reactor vessel (15.2.1.3.3) from failure of the primary pressure regulator ia the closed mode SSES is not expected to exceed flux or pressure scram trip setpoints.

25 Is this conclusion based on quantitative results in earlier FSARs?

If so, reference appropriate sections of these FSARs or provide a summary of the results.

1747 043

0211. In the evaluation of the " generator load rejection" transient, a (15.2.2.3.2) full-stroke closure time of 0.15 seconds is assumed for the SSES turbine control valve.a ( T';V ) . Section 15.2.2.3.4 states that the 26 assumed closure time is conservative compared to an actual closure time of more like 0.20 seconds. However , in Figure 10.2-2 Turbine Control Valve Fast Closure Characteristic, an acceptable TCV closure time of 0.08 seconds is implied. Explain this apparent non-conservative discrepancy and the effect it has on analyses in Chapter 15 requiring TCV closure.

Q211 Explain why vessel steam and bypass flows in Figure 15.2-1 (15.2.2.3.3.1) drop to zero at approximately 37 seconds instead of zero at SSES 45-plus seconds from a L2 vessel level isolation in Table 27 15.2-1.

Q211. Durind the " generator load rejection with bypass" transient, it is (15.2.2.4.1) stated that peak pressure remains within normal operating range.

SSES Explain how this is accomplished since safety / relief valve 28 actuation in the relief mode occurs from the pressure increase.

1747 044

0211. Correct NSOA Figure 15A.6-31 Protection Sequence Main Turbine (15.2.3.2.1.2) Trip - With Bypass Failure, by rettesing the indicated power SSES le v el s . This error occurred during revision of this figure per 29 Question 211.110.

Q211. Would a turbine trip coupled with failure of the operator to put (15.2.3.2.1.3)the mode switch in the startup position before reactor pressure SSES decays to <850 psig (action (5)] be more restrictive on thermal 30 margir.s than the " turbine trip with bypass failure" transient analyzed in Section 15.2.3.3.3.2?

Q211 This section addresses the effect of single failures and operator (15.2.3.2.3.1) errors for turbine trips at power levels >67%.

SSES a) What is the basis for power levels >67%?

31 b) Explain the discrepancy with NSOA Figures 15A.6-25 and 15A.6-31 which refer to power levels >30%.

Q211 In the evaluation of the turbine trip transients, 0.10 second is

( 15. 2. 3. 3. 2 ) assumed for full-stroke closure time of the turbine stop valve.

SSES Demonstrate that turbine stop valve closure times smaller than 32 0.10 second do not result in unacceptable increases in MCPR and reactor peak pressure or provide either (1) justification that smaller closure time cannot occur or (2) a minimun closure time to be incorporated in the Technical Specifications.

Q211. During the " turbine trip with bypass" transient. explain (15.2.3.3.3.1)why vessel steam and bypass flows in Figure 15.2-3 drop to SSES zero at approximately 37 seconds instead of zero at 45-plus 33 seconds crom a L2 vessel level isolation in Table 15.2-3 1747 045

Q211. This section includes a detailed discussion of activity above the

( 15. 2. 4.5 ) suppression pool, activity releases to the environs, and offsite SSES radiological doses. Explain why this information was not included 34 in corresponding sections of other events in Chapter 15 requiring S3V actuation. For instance, the " generator load rejection with bypass failure" transient clearly has a higher peak vessel pressure and lo'ger blowdown.

Q211. Table 15.2-5 does not list all significant events up to 40 (15.2.4.2.1) seconds for the " closure of all MSIV" transient. Include the SSES following items:

35 a) Significant actions associated with attainment of applicable vessel level setpoints, b) Recirculation pump runback if it was simulated in the anal ysis .

1747 046

0211. Include the time at which the turbine stop valves are closed in (15.2.5.2.1) Table 15.2-40.

SSES 36 0211. This section states that the turbine bypass valve and main (15.2.5.3.3) steam isolation valve closure would follow the main turbine SSES and feedwater turbine trip about 5 seconds after they initiate 37 during the transient. Based on this, the bypass valves should close at approximate ~.y 5.01 seconds instead of 12.1 seconds in Table 15.2-10 and Figure 15.2-6. Explain this apparent discrepancy.

1747 047

Q211. Add the following items to Table 15.2-12 to be consistent with (15.2.6.2.1.1) Figure 15A.6-28 for the " loss of auxiliary power transformer" SSES transient:

38 a) Safety / relief valve actuation b) Reactor vessel and containment isolaticn.

Q211. Add the following items to Table 15.2-12 to te consistent with (15.2.6.2.1.2) Figure 15A.6-29 for the " loss of all grid connections" transier,t:

SSES a) Reactor vessel and contair. ment isolstion 39 b) Initiation of the standby AC power system.

1747 048

Q211. It is indicated that credit is taken for safety / relief valve (15.2.7.2.2) operation with " low setpoints" to remove decay heat since SSES bypass valves become ineffective with MSIV isolation. Does this 40 mean use of relief mode setpoints that are lower than the safety mode setpoints or does this imply use of setpoints lower than the relief mode values in Table 15.0-2.

1747 049

0211. For the " failure of lillit chutouwn i"esling" trannient, the I:;AR (15.2.9.2.1.1) considers alternate shutdown cooling methods in the event the SSES residual heat removal (RHR) system in the suction line may not be 41 used because of valve failure. In the analysis, valves in the automatic depressurization system ( ADS) were used to transfer fluid (steam, water or a ecmbination of these) from the reactor vessel to the suppression pool. The RHR system removes the added heat by removing cooling water from the suppression pool and injecting it into the reactor vessel. We require that you perform a test or cite previous test results to demonstrate that the ADS valves can discharge the fluid flow under the most limiting conditions when the fluid is all water. Show that this alternace method is a viable means of shutdown cooling by comparing the system hydraulic losses with the available pump head. Hydraulic losses should be pr ovided for each system canponent and, wherever possible, should be derived from ex perimental results.

1747 050

Q211. Table !S.3-2 indicates that zero vessel steam flow does not occur (15.3.1.3.3.2) until arter 46 seconds. However, Figure 15.3-2 indicates zero SSES steam flow occurs at approximately 36 seconds. Explain this 42 discrepancy.

0211. In the analysis of one and two recirculstion pump trip events in (15.3.1.3.2) Sections 15.3.1. a minimum design rotating inertia was used to SSES obtain a predicted rate of decrease in core flow greater than 43 ex pect ed. Specify the type inertia value (minimum, average, or maximum) used for each transient in Chapter 15 and the basis for selection of each. In the selection basis, include the effect (increase, decrease, or no change) on MCPR and reactor vessel pressure.

Q211. Inclode relief valve flow in Figure 15.3-2.

(15.3.1.3.3.2)

SSES 44 1747 051

0211. a) Table 15.3-3 indicates that zero steca flow should not occur (15.3.3.3.3) until after 41.7 seconds. However. Figure 15.3-3 indicates SSES zero steam flow at approximately 35 seconds. Explain this 45 d iscrepancy ,

b) Include relief valve flow in Figure 15.3-3.

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C211. The narrative on page 15.4-13 states, "The water level does not (15.4.4) reach either the high or low level set points." Table 15.4.3 SSES indicates a low level trip occurs 22.0 seconds after pump start.

'$ Sector 3 of Figure 15.4-6 indicates a low level trip occurs approximately 23.5 seconds af ter pump start. Further:

a) Table 15.4-6 indicates a low level alarm 10.5 seconda after pump start while Sector 3 of Figure 15.4.6 indicates this alarm occurs about 11.5 seconds after the pump starts.

b) Table 15.4-6 indicates vessel level beginning to stabilize 50.0 seconds after the pump starts. Sector 3 of Figure 15.4-6 sho 6,3 no such indication.

Resolve these discrepancies.

Q211. Please identify the diffuser flow units in Sector 2 of Figure (15.4.4) 15.4-6 (and also in Sector 2 of Figure 15.4-7). If this is %

SSES flo w, explain why diffuser flow 1. drops to zero about 30 seconds 47 after the pump starts.

1747 0453

Q211. 1he narrative on page 15.5-3 and Table 15.5-1 both indicate full (15.5.1) HPCI flow is established at approximately 19% of rated feedwater SSES flow in one second. Explain why the curve of feedwater flow in 1;8 Sector 1 does not show this change.

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Q211. The FSAR indicates this transient in analyzed in Subsection (15.6.1) 15.1.4 However, no analytical data (curves) are provided in SSES Subsection 15.1.4 Please supply necessary int'ormation so that 49 this transient can be evaluated concerning a decrease in reactor coolant inventory.

1747 055

0211. A number of inconsistencies exist among narrative descriptions, (4.6) tables, and figures in Appendix 15A relative to Control Rod Drive SSES System. Please resolve.

50 a) Table 15A 6-2 indicates event 7 can occur in stetes C & D.

Figure 15A.6-7 indicates applicability to states A, B, C, u.

Narrative on page 15A-35 indicates any state.

b) Table 15A.6-2 indicates event 16 can occur in states A, 8, &

C. Narrative and Figure 15 A.6-16 indicate applicability in states A & B only, c) Figure 15A.6-17 and narrative on page 15A-39 indicate event 17 is applicable in states C & D. Darinition indicates not applicable in state C.

d) Figure 15A.6-25 does not indicate event 25 applicable to state D only, e) Figure 15A.6-28 Table 15A.6-2 and narrative page 15A-44 for event 23 are inconsistent for applicable states.

f) Narrative page 15A-50, Table 15A.6-4, Figure 15 A . 6-40 f or event 40 are inconsistent for applicable state.

1747 056

Q211 Regulatory Guide 1.29, Section C.1.e, specifies that portions of (3.13.1) the steam systems of boiling water reactors extending from the SSES outermost containment isolation valve up to but not including the 51 turbine stop valve, and connected piping of 2 1/2 inches or larger nominal pipe size up to and including the first valve that is either normc11y closed er capable of automatic closure during all modes of normal reactor operation, be classified Seismic Category I. You state on page 3.13-10 that 3our equivalent portion of the steam system is non-Seismic Category I. Justify your design deviation from the above requirements.

Q211. Your response to Question 211.1 indicates that GE is currently (3.13.1) preparing a Licensing Topical Report to provide an analytical SSES basis for recirculation pump seal leakage. Provide this report.

52 Q211. a) Item (5) on page 3.13-11, discusses those portionr. of (3.13.1) structures, systems, or components (SSC) whose continued SSES function is not required but whose failure could reduce the 53 functioning of items important to safety. Provide a list of these SSC.

b) Regulatory Guide 1.29 Section C.4, requires that Appendix B of 10 CFR 50 should be applied to the above SSC. Provide justification for not including such items in the 10 CFR 50 Appendix B Quality Assurance Program.

Q211 a) Provide a list of those structures, systems and components 1747 057

(3.13.1) which form interfaces between Seismic Category I and SSES non-Seismic Category I features.

54 b) Provide justification for not adhering to 10 CFR 50, Appendix B fo: such items (item (6), page 3.13-11).

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9

m. -

Q211. Regulatory Guide 1.29, Section C.1.b, requires that all reactor (3.2.1) vessel internals to be seismic Category I. Table 3.2-1 indicates SSES that reactor vessel internals other than engineered safety 55 features are not Seismic Category I. Please justify.

Q211. In Table 3.2-1, fill in the following information, where missing:

(3.2.1) (1) Principal construction codes and standards (most pages) .

(3.2.2) (2) Page 18, Main Steam System: Pressure vessels, heat exchangers SSES (all information) .

56 (3) Page 1, Nuclear Boiler System: Air supply check valves (safety class) .

Q211. The RHR pump return line as shown on P & I Diagram M-151 (Fieure (3.2.2) 5.4-13) penetrates into the Suppression Chcaber as a Safety Class SSES 2. Quality Group B line ( pipe 18"-CDD-109 ) . After penetration, 57 the quality group classification is changed to D. Standard Review Plan Section 3.2.2 states that changes in quality group classification are usually permitted only at valve locations, with the valve assigned the higher classification. Demonstrate that the safety function of the system is not impaired due to the fact that quality group classification changes at a point 5.here no valve was located.

Q211 The RHR containment spray line piping (within isolation valve) is (3.2.2) listed as Quality Group A, Safety Class 1, Seismic CateEory I SSES (Table 3.2-1, page 4 ) . In Figure 5.4-13 (P & ID M-151) this 58 line is indicated as 12" GEB-118, i .e. Quality Group B. Resolve 1747 059

this inconsistency.

Q211. Table 3.2-1, page 10, lists piping and valves forming a part of (3.2.2) containment boundary of the Reactor Puilding Closed Cooling Water SSES System as Quality Group B, Safety Class 2, Seismic Category I.

59 Penetration of primary containment for this piping is not shown on any of the relevant P & I Diagrams. Show the above piping and valves on appropriate P & I Diagrams and indicate the classification of thi.9 piping.

Q211. Confirm that the piping from the condensate storage tank to the (3.2.1) HPCI pump suction is safety grade and seismically classified.

SSES 60 C211. Regulatory Guide 1.29 states that systems required for post (3.2.1) accident containment heat removal should be Seismic Category I.

SSES Justify why the Reactor Building Closed Cooling Water System pumps 61 and heat exchangers are not Seismic Category I.

I747 060