ML20029A621
| ML20029A621 | |
| Person / Time | |
|---|---|
| Site: | 05000470 |
| Issue date: | 02/15/1991 |
| From: | Wambach T Office of Nuclear Reactor Regulation |
| To: | Erin Kennedy ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY |
| References | |
| PROJECT-675A NUDOCS 9102250363 | |
| Download: ML20029A621 (14) | |
Text
'
February 15, 1991 I
Project No,'675 Mr. E. H. KennedyI Manager i
Nuclear Systems L censing Combustion Engineering 1000 Prospect Hill Road Post Office Box 500 Windsor, Connecticut 06095-
Dear Mr. Kennedy:
i
~ StlBJECT:
REQUEST FOR ADDITIONAL. INFORMATION ON CESSAR-DC, SYSTEM 80+
Enclosed is a' request for additienti information based on a review by the i
Reactor Systems Branch of-Chapters 5 end 6 of CESSAR-DC.
Please respond within 90 days ef receipt of this request.
The reporting and/or recordkeeping requirements cont 61ned in this letter affect fewer then ten respondents;-therefore, OMB clearance is not required under P. ' L.96-511 Sincerely, Original: signed by Thomas V. Watach, Project Manager Standardization Project Directorhte Division of Advanced Reactors Office of: Nuclear Reactor Regulation:
Enclosure:
As-stated cc w/enclosuic:
See next p6pe DISTRIBUTION:
] Centr 41;Jped8 CLiang, BE23 i
NRC~PDR MFranovich, BE23
. PDCT. R/F--
1
- - PShea.
TWambach^
E EJordan, MNBB3701-
-OGC, 15B18-ACRS(10).-P-315 DCrutchfield,13A2-j WTravers t
15:
PM:PDS[)
D:PDST 4
PShe4 TWambach:cw-CM11 er f,g],gf75 D
. 02/y.
91-02//f/91 02/X /91 DOCUMENT NAME: MEMO C. MILLER /RAI 9102250363 910215 PDR PROJ PDR 6*7 5A-
_____.____-____,____._.__.__..___3
p@m mach, yk UNITED STATES i
NUCLEAR REGULATORY COMMISSION
'/
WASHINGTON. o C. 20065 o
%,, '.N ' /
February 15, 1991
...+
Project No. 675 Mr.E.H.KennedyIcensing Manager Nuclear Systems L Combustion Engineering 1000 Prosocct Hill Road Post Office Box 500 Wirdsor, Connecticut 06095
Dear Mr. Kennedy:
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION ON CESSAR-DC, SYSTEM 80+
Enclosed is a request for additional information based on a review by the Reactor Systems Branch of Chapters 5 and 6 of CESSAP-DC.
Please respond within 90 days of receipt of this request.
The reporting and/or recordketping requirenents contained in this letter affect fewer than ten respondents; therefore, OMB clearance is not required under P. L.96-511.
Sincerely, 4hvW
/
9 wc Thonas V. Wambach, Project Manager Standardization Project Directorate Division of Advanced Reacturs Office of Nuclear Reactor Regulation
Enclosure:
As stated cc w/ enclosure:
See next pbge l
-a-
4 Combustion. Engineering,Inc.
Project No. 675 f
t i
cc: Mr. A. E. Scherer, Vice President Nucle 6r Quality ABB Combustion Engineering Nuclear Power 1000 Prospect Hill Road-7 Post Office Box 500 Windsor, Corinecticut 06095-055-Mr. C. B. Brinkman, Manager Washington Nuclear Operations i
Combustion Engineering, Inc.
12300 Twinbrook Parkway.
Suite 330 Rockville, Maryland 20852 Mr. Stan Ritterbusch -
l Nuclear Licensing t
Combustion Engineering 1000 c respect 11111. Road -
Post Office Box-500 Windsor, Cor.riecticut _06095-0500
-t i
i M
?
v
?
u
'l n
.01.
2 _..,
ENCLOSURE REQUEST FOR ADDITIONAL INFORMATION ON THE DC APPLICATION FOR COMBUSTION ENGINEERING SYSTEM 80+ DESIGN PROJECT NtNBER 675 CESSAR-DC 440.39 CESSAR-DC Section 5.2.2 states that the design basis incident for (6.2.2) sizing the primary safety valves is a loss of turbine generator load.
Figure SA-3 indicates that for the worst case loss of load incident, the reactor trip signal is generated at 5.1 seconds following the event initiation. However, the type of signal which generates the reactor trip is not identified in CESSAR DC. Provide the results of the analysis, including the sequence of events, of the design basis incident for sizing the primary safety valves.
Confirm that the reactor trip is initiated by a second safety-grade signal from the reactor protection system per the acceptance criteria in the Standard Review Plan (SRP) Section 5.2.2.
440.40 The primary system safety valves may be subject to steem and/or (5.2.2) water discharge. Provide a discussion regardir:g the effects to the safety valves due to water relief including the passage of a water slug and water hanner. What w6ter relief rates were assumed in the loading analysis? Are these v61ves cc istent with test results obtained from similar valves?
440.41 CESSAR-DC Section 5.2.2.4.3.3 states that the main steam safety (5.2.2) valves (HS$Vs) are designed to operate in the environmental conditions with the maximum temperature of 330'F for 3 minutes following a main steam line break accident.
Provide a temperature profile for the compartment housing the MSSVs during a design basis I
main steam line break accident to support the assumptions made in the environmental conditions.
440.42 The statement made in CESSAR-DC Section 5.2.2.10.1.1 regarding (5.2.2)
. operator action for low temperature overpressure protection (LTOP)
l is not cicar. Discuss operator actions necessary during transients involving LTOP, including instrumentation and operating procedures avail 6ble that ensure proper operator 6ctions for mitigation of the transients.
440.43 CESSAR-DC Section 5.2.2.10.2.2 st6tes that during heatup, if the (5.2.2)
SCS suction isolettun valves are open and RCS pressure exceeds the LTOP pressure, an alarm will notify the operator that a pressurization transient is occurring during low temperature conditions. Clarify the definition of LTOP pressure mentioned above.
440.44 Provide the results of the analyses for the design bcsis a: ass addi.
(5.2.2.)
tion and energy addition transients including transient curves that du.onstrate the peak RCS pressures are within pressure tempt.rature limits determined for the Systems 804 design.
Instrument 6 tion uncertainties should be factored into your evaluation.
440.45 The following RAI clarifies the staff's position regarding inter-(5.2.2) system LOCA protection 6nd supersedes RAI 440.17 which should be deleted.
Future evolutionary ALWR designs should reduce the possibility of a loss-of-coolant accident (LOCA) cutside containment by designing to the extent practicable all systems end subsystens connected to the reactor coolent system (RCS) to an ultimate rupture strength at least equal to full RCS pressure.
The " extent pr6cticable" phrase is a realization that all systems must eventually interface with atmospheric pressure and th6t for certain large tanks and beat exchangers it would be difficult or prohibitively expensive to design such systems to an ultimate rupture strength equal to full reatter system pressure.
l
3 L.-
It should be noted that the degree of isolation or number of barriers I
(fore.smplethreeisolationvalves)isnotsufficientjustification for using low pressure components that can be practically designed to the ultinate rupture strength criteria.
For example, piping runs i
i should always be designed to sneet the ultimate rupture strength criteri6, as should all associated flanges, connector, packings including valve stem seals, pump seals, heat exchanges tubes, valve bor. nets and RCS drain and vent lines. The designer should make every effort to reduce the level of pressure challenge to all systems and subsystems connected to the RCS.
Our initial review of Systen 80+ design features, including proposed resolution of generic safety issue G1 105, does not provide adequate inforration on how these systens will satisfy the above staff position' for evolutionary ALWRs.
please provide a detailed discus-sion of how the System 80+ design meets the above criteri6. As part of the response include:
1 (1) an identification of all interfaces to the RC$ indicating design and ultimete pressure capabilities for these systems, (2) a color coded siroplified P&lD showing piping and component ultimate pressure capabilities, clearly identifying the interface junctions.
For 611 interfacing systems and components which dc rot meet the full RCS ultinate rupture strength criteria, justify, for each case, why it is not practichble to reduce the pressure challenge any further. This justification must be based upon engineering feasibility analysis and not solely risk benefit trade-offs.
i For those interfaces where acceptable justification on the j
impracticability of full RCS pressure capability has been provided, j.
there must be a denonstration of compensating isolation capability.
For exartple, it should be demonstrated fcr each interface that the l
l
...... -., _. _ _ _,... _ _... _ _ ~ _. _..-
4 l
t degree and quality of isolation or reduced severity of the potetitial pressure challenges compensate for and justify the safety of the 10w pressure interfacing system or component. Adequacy of pressure i
relief and piping of relief back to primary containment are possible l
consideretions.,%s identified in SECY 90 016 each of these high to low pressure interfaces must also include the following protection measures:
(1) the capability for leak testing of the pressure isolation valves (p!Ys),
(2) valve position indication that is available in the e rol room when isolation valve operators are deenergized, and (3) high-pressure alarms to warn control room operators when rising RCS prcssure approaches the design pressure of the attached low pressure systems and both isolation valves are closed.
440.46 Provide a discussion on the 6bility of the pressurizer surge line for (5.4.10) the System 804 design to withstand the effects of thermal strstifica-tion.
(Reference NRC Eulletin 80-11, December 20,1986.)
440.47 CESSAR-DC Section 5.4.7.1.2 states that the shutdown cooling system (5.4.7) is designed so that the SCS pumps can be tested at full flow condi-1 tions when the reactor is operdting at power. Discuss how this test could be achieved without overpressurization of the SCS.
l 440.48 provide a SCS pump characteristic curve for the staff review.
(5.4.7)-
Specify the available and required net positive suction head for the SCS pumps.
I-l'
-440.49 provide a discussion of the procedures and plant systems used to (5.4.7) take the plant from normal operating conditions to cold shutdown conditions.
This discussion should include, heat removal, depressurization, flow circulation, and reactivity control.
5 440.50 Provide a discussion on how the System 80+ design complies with the (5.4.7) staff position set forth in the Branch Technical Position (BTP) R$B E-1 ettached to SRP Section 5.4.7.
440.51 Provide the results of a thtnnal hydraulic analysis including (5.4.7) transient curves for a natural circulation cooldown following a reactor trip at full pcwer. Discuss the method to be used for preventing voids in the reactor upper head during the cooidown.
440.52 Per the staff position of BTP RSD 5-1, confirm that a boron mixing (5.4.7) and r.atural circulation cooldown test will be perforraed in the first plant with a System 80+ design.
440.53 Expand the SCS failure raodes and effects analysis including potentici (5.4.7) electrical single failures and failures of interlocks on the pressure isolation valves. Wt actions are necessary and where must they be performed under those conditions? Describe the procedures which the operators will need to use following a postulated failure such as discusst.d above.
440.54 Discuss the alarms and indications which would inform the operators (5.4.7) that the SCS suction line isolation valve has closed while the plant is in shutdown cooling? Is there any common mode failure which would result in isolation valves in both trains being closed w H le in shutdcwn cooling. Are there any manual maintenance valves whose closure could isolate the SCS suction, if so, describe procedures and controls to restrict this possibility?
440.55 Provide the following information related to pipe break or leaks in
.(5.4.7) high or moderate energy lines outside containment associated with the SCS when the plant is in a shutdown ecoling mode:
1.
Determine the maximum discharge rate from a pipe break in the systems outside containment used to matntain cure cooling.
i l
i 6
2.
Determine the time available for recovery based on these discharge rates and their effect on core cooling.
l 3.
Describe the alarm available to alert the operator to tht:
event, the recovery procedures to be utilized by the operator, and tirae available for operator action.
A single failure criterion consistent with SRP Section 3.6.1 and BTP ASB 3-1 should be applied in the evaluation of the recovery procedures utilized.
1 440.56 Indicate whether there are any systems or component needed for (5.4.7) shutdown cooling which are deenergized or have power locked out during plant operation.
If so, discuss what actions have to be taken to restore operability to the components or systems, and describe where the actions must be taken.
440.57 Regarding the power supplies to the SCS isolation valves, confirm (5.4.7) that a single failure cf a power supply will not prevent isolation of the.SCS when RCS pressure exceeds the design pressure of SCS.
Also, a single failure in power supplies cannot result in the inability to initiate at least one train of SCS. Confirm that the autoclosure interlock for the SCS isolation valves have been renoved for System 80+.
440.58 Discuss provision.. of pump protection available for SCS pumps from
.(5.4.7) potential low flow or no-flow operating conditions.
440.59 CESSAR-DC Section 6.3.1.2.1 states that the safety injection system (6.3)
(SIS)-is designed so that the SIS pumps can be tested at full-flow conditions with the reactor at power. Discuss how this functional design bases could be achieved in light of the fact that available pump head at full-flow is much lower than the RCS pressure at power.
L
440.60 CESSAR-DC Section 6.3.1.3 sutet that air for all SIS pneumatic (6.3) valve operators shall be clean, dry, ano vil-free.
Identify all cir-operated valves in SIS and discuss the safety classificction of the air supplies.
If a non-safety grade compressed air system is used, assess the consequences of the failure of these valves in the case of loss of air supply. Describe sizing analysis for safety grade air supplies and what test programs will be run to demonstrate adequacy of air supply to provide safety functioning throughout the time period they are required. Account for the credible air leaks or propose controb to 1.nsure a leak tight system, 440.61 The SCS and SIS pumps are used for long term cooling 411owir,g an (5.4.7, accident.
Discuss the capability of these purups to be opereted for 6.3) very long periods of time following a postulated severe accident.
Are there any test date available to support the above discussed capability? How will long term performance capebility be maintained throughout plant lifetimes?
(
440.62 The System 80+ design provides four high pres:ure safety irjection l
(6.3) pumps.
Discuss the pump performance following a safety injecticn signal due to a large break LOCA.
What provisiore, are availoble to prevent these HPSI pumps from runout doe to low RCS pressure?
l l
440.63 Provide the calculated available NPSH values for the HPSI pumps I
(6.3) during injection mode and long term cooling mode and demonstrbte l
that sufficient margins exist compared with the required NPSH l
values for these purips under various operating conditions.
440.64 Discuss the provisions available in the Sy; tem 80+ design to keep l
(6.3)
SIS piping 'illed with water to prevent potential weter hanyner.
440.65' CESSAR-DC Section 6.3.3.2.2 states that an evaluat$on of possible (6.3) single failure shows that no single failurc in SIS or the dietel-generator system is the worst condition for the large breat analysis.
This is because full SI flow would maximite the SI flow spilling to
g containment which minimizes the conteinment presst'1. This, it, turn, minimizes the core reflooding rate.
For the limiting break location (pump discharge leg), there is no single failure that results in en injection flow rate which cannot keep the downcomer filled to the elevation of the discharge leg. Thus maximum $1 flow rate from all four SI ptops are assumed in your large break LOCA analysis. The staff underst6nds the qualit6tive argument presented above. However, the results of a thermal hydraulic analysis should be provided to
'erify that with a single failure in the diesel-generator system, SI flow from only two S1 pumps would indeed keep the downcomer filled to the elevation of the pump dinharge leg in the same time f rune as that with all four Si pumps feeding the rebetor vessel.
440.06 Identify any lengths of ECCS piping which have nonna11y closed (6.3) v61ves that do not have pressure relief in the piping section between the isolation valves.
Verify that all pressure isolation check valves can be individually tested for back leakage.
440.67 provide a list of all active components which are required for (6.3) operation of the ECCS.
provide safety and seismic classification for each component and indicate what services such as cooling, lube oil, and air are necessary for the proper functioning of each component.
440.68 Discuss in detail how your interface rcouirement will specify the (6.3) preoperational test program for ECCS in conformance with RG 1,68 and 1.79.
Specifically, include the procedures which will be used to verify nominal and runcut ECCS flow, pump characteristics, piping losses and verification that each check valve in the systen is capable of performing both isolation and flow function.
440.69 stcribe the instrumentation for level indication in the IRWST.
(6.3) alss provide detailed design drawing of the IRWST, including the design provisions which preclude the formation of air entraining vorticos. Discuss the anti-vortex criteria which w6s utilized during
the design of IRWST. Discuss what testing has been performed to verify design objectives of no vortex formation.
440.70 Describe the means provided for ECCS pump protection including (6.3) instrumentation and alarms available to indicate degradation of ECCS pump performance. The staff's position is that suitable means should be provided to alert the operator promptly to possible degradation of ECCS pump performance. All instrumentation associated with monitoring the ECCS pump performance should be operable without offsite power, and should be ible to detect conditions of low discharge flow.
440.71 Discuss the percentage of safety injection (SI) flow capacity (in (6.3) terms of best efficiency flow) for SI pump minimum flow recircula-tion required to protect against hydraulic instability or impeller recirculation problems during extended SI pump low flow operations.
(Reference NRC Bulletin 88-04, May 5, 1988.)
440.72 Discuss the design criteria for the safety injection pumps, contain-(6.3, ment spray pumps, and the shutdown cooling pumps, and discuss 5.4.7) whether the pump design criteria includes pump operations at or near shutoff head conditions?
440.73 provida an analysis for the potential for pump-to-pump interaction (6.3, resulting in a pump dead-heading scenario for the safety injection 5.4.7) system, the containment spray system, and the shutdown cooling system.
This analysis should identify all pumps and piping configurations that are pathways for pump-to-pump interaction including all shared comon minimum flow recirculation lines and test lines.
(ReferenceNRC Information Notice 90-61, September 20,1990) 440.74 In PAID Figure 6.3.2-1A for SIS short term injection indicates that (6.3) velve SI-30S is normally closed for SI injection into hot leg f1; however, valve SI-304 is drawn in the open position for the SI injectior, path to hot leg #2. Should 51-304 be normally closed for 1
10 the initial phase of short term safety injection (less than two hours)? Provide clarification.
440.75 CESSAR-DC Section 6.3 addresses operator 4ctions for post-LOCA (6.3) operations. Which valves require manual closure on the injection path for SI-3 and SI-4 during valve realignment for simultaneous direct vessel injection (DVI) and hot leg injection? List ali valves requiring manual operation, administratin controls / procedures of these valves, and control room features ;i.e., key-operated control switches,valvepositionindication)thatminimizethepotentialfor valve n.isalignment.
440.76 What is the niinimum amount of boron concentration and mininom liquid (6.3) voluc.e assumed in the safety injection tanks (SITS) for the large
':ek LOCA analysis, and why is 2000 ppm, as cited in Table 6.3.2-1, ecceptable in lieu of the 4000 ppm required for the System 80 SITS?
440.77 Whet is the minimum amount of bor?.ted water assumed in the IRWST for (6.3) the large break LOCA analysis?
440.78 On page 6.3-3 under CESSAR-DC Section 6.3.1.3(A)(2)(f) states that (6.3)
SIT vent valves SI-331, SI-335, SI-320 SI-333,51-330
- 334, 6
SI-328, and SI-332 shall have provisions to remove power from these valves. According to Figure 6.3.2-1B, these valves are nitrogen supply valves. Should this statement address S!T vent valves51-320, SI-321, SI-322, SI-323, SI-324, SI-325, SI-326, and SI-327?
440.79 What type of " throttle valves" are S0-653 and S0-652 supposed to be?
(6.3, (Reference CESSAR-DC Sections 5.4.7, 6.3, ard Figure 5.4.7-3 and 5.4.7)
Figure 6.3.2-1A.)
440.80 CESSAR-DC Section 5.4.7.2.3. A(3) " Overpressure Protection" on page (5.4.7) 5.4-32 indicates that an alarm on the SCS system is provided to warn operetors of a pressurization transient during low RCS temperature e
-___m______
- 4..
11 conditions. This instrumentation should be appropriately reflected on the associated PalD. figure 5.4.7-3.
440.81 Provide an environmental qualification of ECCS equipment for post-(6.3)
LOCA conditions; i.e., Appendix 3.118.
440.82.
' Current editions of CESSAR-DC include Table 6.3.2-4e " SIS Flow Point h cor espond g f ow ding am doe nt a e f ow po t ata a eled
~for the location of the data readings Does this table reflect flow
= point-data for the System 80+ SI system?
i r
1a o
y v
-