ML20235G114
| ML20235G114 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 06/04/1987 |
| From: | Ball S, Moses D OAK RIDGE NATIONAL LABORATORY |
| To: | NRC |
| Shared Package | |
| ML20235G082 | List: |
| References | |
| CON-FIN-A-9478 NUDOCS 8707140107 | |
| Download: ML20235G114 (18) | |
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TECHNICAL EVALUATION REPORT FORT ST. VRAIN NUCLEAR GENERATING STATION DOCKET 50-267 LICENSEE: PUBLIC SERVICE CO. OF COLORADO FORT ST. VRAIN SAFE SHUTDOWN FROM 82% POWER
- PREPARED BY: .
S. J. Ball D. L. Moses k Oak Ridge National Laboratory Oak Ridge, TN. 37831 June 4, 1987 NRC Lead Engineer: K. L. Heitner Project: Selected Operating Reactors Issues (FIN A9478) a ff14 97 870 Jdc8 osolp
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NOTICE This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, or any of their employees, makes any warranty, expressed or implied, or assumed any legal liability or responsibility for any third party's use, or
" the results of such use, of any information, apparatus, product or process disclosed in this report or represents that its use by such third party would .
not infringe privately owned rights.
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l MT Technical Evaluation Report - Fort St.' Vrain Safe Shutdown From 82% Power S. J. Ball D. L. Moses
- 1. Introduction ORNL has provided technical assistance to NRC in their evaluations and analyses of issues raised in Fort St. Vrain's 1987 Power Ascension ORNL had previously performed analyses which confirmed the Plan (P-87038).
Public Service Co. of Colorado (PSC) assertion that for long-term operation at 35* power, a safe shutdown could be- achieved even if all equipment exposed to a harsh environment in a worst-case high-energy line break (HELB) scenario were assumed to fail. From these studies it was concluded that, for a worst-case permanent loss of forced circulation (LOFC) event, a long term cooldown using only firewater coolant in the PCRV liner cooling system (LCS) Inas the addition, ultimate heat sink would not result in significant fuel failure. i' restart of coolant to the LCS after postulated prolonged downtimes would be both feasible and acceptable.
Upon successful implementation of the steam line rupture detection and isolation system (SLRDIS) at FSV, it was determined by analysisThis that the long is due to term reactor power should be limited to values below 100t.
previously undetected limitations in the shutdown cooling systems designed for use in emergencies. Because of the inability to simultaneously flood all six of the reheater sections of the steam generators in a loop (at least with the present flow capacity and piping arrangements), reheaters can no longer be counted on for emergency cooldowns following postulated 90 min duration LOFC events. If hot helium coolant from the core (following restart of the .
circulation) were not cooled by the reheater bundles in several modules, hot gas could impinge on the economizer-evaporator-superheater (EES) sections' tube bundles and cause failure of the tubing. Consequently, PSC requested a change in the Technical Specifications to eliminate reliance on the use of the reheater sections for Safe Shutdown Cooling (P-87002).
A series of PSC ane. lyses were submitted to NRC that Justified long-term power operation at powcr levels between 82% and 87.5%, depending on which accident scenario was postulated. In the Environmental Qualification (EQ) l case, it was claimed that 87.5% power operation could be Justified by relying on only one firewater pump supplying coolant to only one of two (redundant)
EES sections (i.e., all six EES modules in one loop) following a 90 min LOFC event. The EQ event postulates only the use of equipment which is qualified to withstand the design basis earthquake, the maximum tornado and the most limiting HELB. In order to meet this goal, it was necessary to install two (one per loop) 6 in. vent pipes from the EES sections to provide for single The other failure proof venting during the once-through cooling mode.
scenarios involved only the use of equipment qualified to survive and operate J
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L > DWI following fires in non-congested cable areas per 10CFR50, Appendix R. For such fires, the emergency cooling flow sources can successfully accommodate cooldowns from power levels up to 83.2%. -In January 1987 PSC formally requested permission to operate at power levels up to 82% (P-87038).
The ORNL technical assistance provided was'mainly in two areas. The .
I first involved confirmatory analyses of various scenarios using accident codes developed under both RES and NRR sponsorship. The second involved a detailed re"lew of the ability of the existing systems to supply sufficient water flow to the helium circulator Pelton wheel drives and the EES sections of the steam generators during the cooldown scenarios. N o other smaller tasks were included, investigations of possible structural and metallurgical failures in !
the steam generators -(R. C. Gwaltney), and en assessment of the potential for water hammer upon restart of coolant flow to the hot tube bundles (C. R.' Hammonds). - A brief letter report on the latter of these smaller tasks is' provided in an attachment.
- 2. Accident Scenario Analyses 2.1 ORECA Code )
The original version of the ORNL ORECA code for siriulating HTGR core dynamics is described in ORNL/1M-5159 (1976). Subsequent updates appeared periodically in RES quarterly progress reports, and a detailed writeup on the -
ORECA code family. was submitted to NRC -in February 1985 as part of an assessment report of all ORNL HTGR accident codes. This latter report described three versions of ORECA: severe-accident and " verification" versions
.s of FSV simulations, and a source term study version for the 2240-MW(t) H1GR
' design. The ORECA-FSV versions have been used extensively by many different users .in many different types of analyses, and have been verified for a variety of transients by comparisons with both plant data and other
~ independent analyses. ORECA is considered to be a "best estimate" code; conservatism are accounted for by means of sensitivity analysis.
The ORECA code used for this task was a modification of the severe
- accident sequence analysis (SASA) version. The modifications included an addition of a model of the flooded steam generator EES section, and a model to predict stagnation or reverse flows within the worst-case refueling regions.
The latter feature had been developed as part of ORNL's assessment of PSC's proposed changes to the Tech Spec L.C.O. 4.1.9, which deals with concerns about fuel overheating in low-flow, low-power operating modes.
2.2 Model and Parameter Assumptions Specific input data for the analyses were provided by PSC and GA. The major assumptions used for the reference case were as follows:
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- 1. Equilibrium core region peaking factors (Max. = 1.83).
l 2. Maximum allowable region outlet temperature dispersions (during L operation) consistent with LCO 4.1.7. (The maximum mismatch value provided fl00 F) was increased by 50 F to allow for region outlet temperature measurement error.) Orifice positions were assumed fixed for the duration.
! This corresponds
- 3. . Long-term operating power (before shutdown) of 87.5%.
to the EQ case.
- 4. PSC-supplied estimate of EES cooling water flow (940 gpa) following the 90 min.LOFC. The helium coolant flow was assumed to be " manually controlled" to limit EES water outlet temperatures to 250 F to prevent boiling. The capabilities of the Pelton wheel drive were not limiting.
- 5. EES (water-cooled) initial performance crwracteristics per GA's The subsequent EES performance (with varying inlet
- l. SUPERHEAT code.
temperatures and flows) was calculated using the ORNL code AWHEXI.
- 6. Flow to the LCS is not available for the duration.
2.3 Analysis Results ORNL analysis of the reference EQ case accident scenario resulted in predicted temperatures quite close to those provided by PSC. The ORNL predicted maximum fuel temperatures were somewhat lower than those. of PSC, while the mean core outlet temperatures were slightly higher. The ORECA maximum fuel temperature was 2560 F (1405 C) vs 2858 F (1570 C) per PSC, hence
-indicating less likelihood of fuel damage. The ORECA maximum average core outlet gas temperature was 1625 F vs. approximately 1500 F per PSC. The f corresponding primary system pressures predicted were 425 psia (ORNL) vs.
4 j 325 psia (PSC) at the time of maximum core outlet temperature, so overall, the potential for damage to the (dry) reheater tube bundles would be somewhat greater for the ORNL predictions. Comparisons of the results for maximum fuel and average core outlet gas temperatures are shown in Fig. 1, and.for primary system pressure response in Fig. 2. The " manually-controlled" values of primary helium flow are shown in Fig. 3, where the ORBCA predictions give somewhat lower allowable flows to prevent boiling in the EBS tubes.
Calculations of intra-region flow redistributions in the worst-case regions showed that there was no intra-region flow stagnation (or flow reversals) during the cooldown.
Sensitivity analyses were also done to assess both calculational and k
) operational margins for error. For an assumed 20% reduction in the available l EES. cooling water flow, the predicted maximum fuel temperature was only
$. 2685 F (1475 C), still well below temperatures causing significant failure rates. Variations in the heat transfer perfonsance of the water-cooled EES f, were made within reasonable bounds, and had very little impact on the resulting peak temperatures predicted, as did the assumption of ICS failure.
Various assumptions about the response of the operators in controlling helium E flow were also found to be unimportant, although the assumption of a failure F
to avoid boiling and subsequent choking of the flow was not evaluated. It is 3
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. EfI reasonable to assume that the recovery time from such a complication (shutting off the helium flow temporarily to reestablish EES water flow) would not be excessive.
Another assessment of the " safety margin" available was done by compounding several uncertainties and/or conservative assumptions to see what it would take for the predicted maximum fuel temperature to approach 2900 F (1600 C). In this case, power for the EQ case was increased by 5% (to 91.9%) to account for measurement error, EES cooling water flow was decreased by 25* (to 705 gpe), and the restart of forced cooling was delayed an additional 20 min. (to 110 min). The resulting peak fuel temperature was 2830 F (1555 C), with some refueling regions experiencing reversed flow. The .
maximum average core outlet temperature predicted was 1730 F. This indicates l that substantial margin for error is available before there would be any concern for significant fuel damage.
Calculations of the " Appendix R" reference case, with power limited to 83.2% and EES open loop cooling water flow limited to 700 gpa (P-87158),
showed even milder transients than did the EQ cases. The maximum predicted fuel temperature was 2460 F (1350 C) and the maximum average core outlet gas temperature was 1650 F.
- 3. Water Supply to the Helium Circulator Pelton Wheels and Steam Cenerators 3.1 Introduction and Background This portion of the technical evaluation is directed to resolving restart issues arising out of Reportable Occurrence (RO) 50-267/86-026 and related to the adequacy of the seismically-qualified Class I Safe Shutdown Cooling System at Fort St. Vrain. The adequacy of the Class I water supply for Safe Shutdown Cooling was initially brought into question as the result of deficiencies identified during startup testing as reported in Unusual Event l Report (UER) 50-267/76-05. The particular deficiencies identified in both l UER 50-267/76-05 and RO 50-267/86-026 relate to the provision of facility I firewater to Class I piping in the Fort St. Vrain secondary cooling system. (
To accomplish safe Shutdown Cooling, Class I firewater has to be supplied with sufficient prescure and flow rate to drive one helium circulator water turbine at required speed and to flood one section of the steam generator !
economizer-evaporator-superheater (EES) with sufficient cooling capacity to I maintain the maximum fuel temperature below 2900 F. In addition, Safe i Shutdown Cooling should prevent the surface temperatures of the steam generator tubes and of the helium circulator components from reaching values l that could challenge reactor coolant boundary integrity or the capability to i perform long-term cooling. Similarly, Safe Shutdown Cooling should preclude primary system pressure from increasing to the point of lifting the safety relief valves. The sufficiency of the Class I firewater supply to the steam i generator EES is dependent on the backpressure in the steam generator, and the 1
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. EFT EES backpressure is a function of the downstream piping resistance to EES outlet flow and of whether boiling occurs in the EES tubing after forced cooling of the reactor core is restored.
The two steam generator EES sections are the only primary system heat exchangers capable of supporting Safe Shutdown Cooling. This observation is I based on the recent finding, as reported in DO 50-267/86~020, that the j resistance in the firewater flow path to the steam generator reheater sections l was too high to permit adequate flow for Safe Shutdown Cooling. Previously, 1 Updated FSAR Section 10.3.10 had claimed that the two reheater sections each j provided a Class I firewater cooldown capability from full power conditions and that was fully redundant to the EES cooldown capability on firewater.
Recent analyses reported in Attachment 4 of P-86682 indicate that firewater cooling through a single reheater section module (where one module of six in each section is the nost that can be flooded) will provide effective Safe Shutdown Cooling capability only from 39% of rated reactor power and with a 90 minute interruption of forced cooling.
As reported in UER 50-267/76-05A (final supplecent) and more recently amplified and clarified in the licernee's response to a request for additional information (PSC Response 6, Attachment to P-87110), testing in 1976 demonstrated that each circulator could apparently be driven at speeds exceeding 700 rps with 1000 gpm or more supplied to the steam generator EES.
This was achieved by: ,
(1) fully opening the bypass valves PV-22129 and PV-2229 in the lines from the EES outlet to the bypass flash tank and pre-flash tank, respectively, thereby reducing EES backpressure to 50-60 psig; and (2) by physically modifying the discharge flow paths in several valves (HV-21257 through HV-21260) upstream of the Pelton wheel water turbine to reduce pressure drops and thereby increase flow capacity to the respective water turbines.
No testing was performed to quantify reheater flooding with simulated firewater in the 1976 tests. Preoperational testing in 1973 had identified limitations to reheater flooding on condensate, but the results of the preoperational tests and analysis were not effectively documented. Therefore, the simulated firewater tests that are reported in UER 50-267/76-05 did not include tests or analysis of reheater flooding because the potential problem had not been recognized by PSC.
4 However, a related event, described in UER 50-267/76-09, identified the fact that the initial analysis of the firewater cooldown had used nominal instead of design power peaking factors. To prevent flow reversal or flow stagnation in the reactor core during the firewater cooldown from power levels greater than 70% of rated, UER 50-267/76-09B indicated the need to install emergency 5
water booster pumps in the piping immediately epstream of the circulator water turbine so as to increase primary system helium flow from an apparently achievable 2% to between 3 and 4%. Subsequently, NRC imposed the requirement j to assume a 90 minute delay in establishing firewater flow and Safe Shutdown j Cooling following the Design Basis Earthquake or Maximum Tornado (the Class I '
events).
In September 1986, as noted in RO 50-267/86-026, a reanalysis of the design i basis Class I Safe Shutdown Cooling transient (Updated FSAR Sections 10.3.9 and 14.4.2.2) was performed in support of Environmental Qualification. Per the licensee event report, the reanalysis indicated:
..,that restart of forced circulation at 3% of rated, approximately 1-1/2 hours after a scram from 105* power, will cause steaming and a large pressure increase in the steam generator EES and discharge piping. Under I these conditions, secondary flow would be significantly reduced due to the limited capacity of the firewater pumps. Due to this reduction in secondary coolant flow, the heat removal rate computed in the FSAR analysis would not be achievable.
More recently, the confirmatory analysis (Attachment I to P-87053) of other FSAR cooldown transients has shown that the original FSAR analysis of the Class I Safe Shutdown cooling transient assuming an immediate start of firewater (i.e., without assuming the 90 minute delay as currently required) .
also overpredicts the firewater heat removal capacity through the EES by at least a factor of two, specifically, an assumed 180 x 108 BTU /hr in the original analysis versus an apparently achievable 86.7 x 108 BTU /hr from the most recent analysis. For the case of the 90 minute delay in restoring forced cooling (i.e., the current design basis Class I safe Shutdown Cooling transient), the limiting heat load is stated to be 73.5 x 108 BTU /hr in attachment I to P-87053; however, the achievable firewater heat removal capacity is not provided for the 90 minute delay in restarting forced cooling.
It should be the same once the steam generator EES is pre-cooled to prevent boiling in the EES tubes.
Thus, the attempted resolution ta UERs 50-267/76-05 and 50-267/76-09 were inadequate because the associated testing and analyses did not adequately address secondary side pressure and flow conditions that must be attained in the EES during both the original and the current version of the Class I Safe Shutdown Cooling scenario. Also, the attempted resolutions to both of the 1976 UERs failed to address the cooling capability of the reheater sections using firewater. Therefore, this portion of the technical evaluation assesses the adequacy of the recent calculations submitted by PSC and the need for further analysis or testing. As indicated previously, this technical evaluation focuses on the perfonnance of the Class I Safe Shutde <n Cooling system in accommodating the Design Basis Earthquake and Maximum Tornado (Updated FSAR Sections 10.3.9. and 14.4.2.2).
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i 3 Review of the New Class I Firewater Flowpath Calculations in Recent PSC The results of PSC's recent analysis have been documented in several !
submittals. The cover letters to the PSC submittals use the terminology " Safe Shutdown Cooling" to refer to the scenarios addressed in the various sets of {
l subcontractor calculations that are attached to these documents. These scenarios include emergency core cooling following high energy line breaks in either the reactor or turbine building and following major fires in the noncongested cable areas. Emergency core cooling following the design basis seismic event is incorporated with the analysis to address Environmental l i
Qualification per 10CFR50.49. J Attachment 4 to P-87002 provides the safety analysis supporting a proposed change to the Fort St. Vrain Technical Specifications. This change is to i I
eliminate reliance on the reheater section of the steam generator fo- Safe Shutdown Cooling and to require instead full reliance on the EES section for the firewater cooldown from power levels up to 82% of rated reactor power.
Adequate firewater flow to the EES section of one steam generator and to one !
circulator water turbine drive in the same loop is to be achieved by reducing !
EES backpressure with the use of seismically-qualified six inch vent lines to {
the atmosphere from each EES discharge header. In addition, as shown in j Figure III-l of Attachment 4 to P-87002, a new seismically qualified (Class I) .
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firewater flow path has been installed between the firewater discharge header j and the emergency condensate header. The new flow path has only two isolation valves that have been installed for remote operation outside the turbine building in order to avoid a harsh environment in case of a high energy line break in that building. The new firewater piping to the emergency condensate
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header meets both the seismic and environmental (10CTR50.49) qualification j requirements for Safe Shutdown Cooling.
l Analyses presented in Attachment I to P-86683, Attachment 2 to P-87055, and the Attachments to P-87171 have been submitted by PSC to demonstrate the {
firewater supply capability for the new Class I firewster flow path. The Attachment to P-87171 include the results of a simulated firewater flow test -
using condensate to drive a circulator water turbine. This test and an l accompanying sensitivity analysis were submitted by PSC to demonstrate an adequate capability to produce a primary system helium flow of 3.8% of full {
l load at a firewater flow that is less than that which is estimated to be i available. Sensitivity analyses show that increasing flow to the EES section {
does not significantly degrade the suction pressure and flow available to the emergency water booster pump upstream of the circulator water turbine drive.
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s The firewater flow analysis and the flow sensitivity studies were performed using a computer code developed by Proto-Power Corporation and described in Attachment 6 to P-87055. The theory and methods used in the pressure drop 7
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program have been reviewed and appear to be both reasonable and consistent with other steady-state methods for flow and pressure drop calculations.
The same Proto-Power Corporation computer code described in Attachment 6 to P-87055 has also been modified and used, as reported in Attachment 9 to P-86682, to calculate the EES flooding time for pre-cooling the EES to subcooled outlet flow conditions prior to restart of the helium circulators on water turbine drive. As discussed in Attachment 9 to P-86682, the firewater flow and pressure drop computer code was solved iteratively in a quasi-static manner with a heat transfer code simulating the EES tube thermal performance.
The quasi-static analysis yielded a 14 minute flooding time for the EES. The flooding time was reported to have been conservatively confirmed by a calculation by GA Technologies as cited in PSC response 7 in Attachment I to P-87055. As discussed in the Attachments to P-86682, P-86683, and P-87055, CA Technologies reportedly checked the steam generator steady-state flow and pressure drop results with the SUPERHEAT code. Also, as discussed in Attachment 5 to P-86682, GA performed limited checking of other steady-state firewater flow calculations, but not the Safe Shutdown Cooling flow path, using the SNIFFS flow network computer code and hand calculations. Since no complete set of alternate calculations are available for either the steady-state firewater flow calculations or the EES flooding (pre-cooling) time calculations, we recommend that NRC should perform an audit of the Proto-Power Corporation calculations. Such an audit is needed to assure that the type of problems with inadequate verification checking encountered in .
UERs 50-267/76-05 and 50-267/76-09 does not recur for RO 50-267/86-026.
In Attachment 4 to P-87002, PSC states that the new Class I flowpath also meets the single failure criterion for both active and passive failures. To accommodate a single passive failure, Attachment 4 to P-87002 addresses reliance on the original firewater flow path in System 45 (the Fire Protection j l System) that is the basis for the Safe Shutdown Cooling results reported in FSAR Section 10.3.9, 10.3.10, and 14.4.2.2. In Attachment 4 to P-87002, the firewater system flow is assumed to be routed to the emergency feedwater header to accommodate a single passive failure, but the requirement to use j this flow path is based on a single passive failure that, consistent with NRC l Policy given in SECY-77-439, is assumed not to occur until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after {
emergency cooling has been initiated. This alternate flow path has not been l analyzed in calculations submitted by PSC.
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Active single failures in the preferred new flow path, as illustrated in Figure III-l of Attachment 4 to P-87002, are assumed to be Ilmited to " failure to function of an electrical component of the two valves (HV-4518 and HV-4519) in the (new) line (that) can be compensated for quickly by a manual action in a mild environment." Mechanical failures of these valves failing to open have not been addressed.
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L If either HV-4518 or HV-4519 were to fail to open, alternate flow paths exist via the firewater system to either the emergency condensate header supply or the cuergency feedwater header, which was alluded to above. The alternate flow path to the emergency condensate header is the original preferred flowpath for Safe Shutdown Cooling as described in FSAR Section 10.3.9 and 14.4.2.2. A review of the recent PSC analyses that use this original flow path is presented in the following section of this report.
3.3 Review of the Original Class Flowpath Calculations in Recent PSC Submittals There was no attempt to clarify in the recent PSC submittals that the original FSAR Class I cooling configuration ir most closely analogous either to the
" emergency cooling with EES using firewater pump (open loop)," as shown in Fig. A-1 of Attachment 5 to P-86682, or to the " firewater flow diagram for Appendix R Train B case, open loop," as shown in Fig. 2.4 of Attachment 2 to P-85055. Both of these configurations represent firewater supply via the emergency condensate header as assumed in the Class I performance and safety analysis reported in Updated FSAR Sections 10.3.9 and 14.4.2.2.
In the configuration cited in Attachment 5 to P-86682, about 100 gpm of fluid is assumed to be vented to the atmosphere via an electromatic valve (PV-22162) located downstream of the steam generator EES, and the rest of the flow is vented to the condenser through the desuperheater and flash tank. Per FSAR .
Section 10.3.9, the original Cless I event scenario assumed venting from ruptured non-Class I piping downstream of the steam bypass valves to the desuperheater. The licensee did not specify which of these configurations presents the higher backpressure to the steam generator EES and thereby produces the lower flow from the firewater supply system; however, use of the newly installed and seismically qualified six inch vent lines off the EES outlet piping to the atmosphere allows for lower EES backpressure. The maximum fuel temperature calculated in the recent analysis (Attachment 5 to P-86682) is 2511 F for a cooldown from 77.9% of rated reactor power. The firewater flow rate to the EES is calculated to be 829 gpm with an EES backpressure of 107 psia at the steam generator ring header and 87.3 psia at the mainstream bypass valve (PV-22129). Water exiting the EES is subcooled.
Similarly, in the configuration cited in Attachment 2 to p-85055, flow through the open loop (i.e., atmospheric venting via the newly installed six inch vent ;
lines) lasts for only five hours, after which the loop is assumed to be closed i in the Appendix R Train B Case (See Fig. 2.5 in Attachment 2 to P-85055). In the open loop configuration, the firewater flow is 996 gpm, but is reduced to 789 gpm in the closed loop. The steam generator EES backpressure at the j outlet is " controlled" to 76 psia (open loop) for the first five hours and i 97.8 psia (closed loop) thereafter; presumably, the quoted pressures are at the steam bypass valve. The msximum fuel temperature is calculated to be 2644 F following shutdown cooling from a power level of 87.5%.
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mfI If either HV-4518 or HV-4519 were to fail to open, alternate flow paths exist via the . firewater system to either the emergency condensate header supply or the' emergency feedwater header, which was alluded to above. The alternate flow path to .the emergency condensate header is the original preferred flowpath for ' Safe Shutdown Cooling as described in FSAR Section 10.3.9 and 14.4.2.2. A review of the recent PSC analyses that use this original flow path is presented in the following section of this report.
3.3 Review of the original Class Flowpath Calculations in Recent PSC Subafttals There was no attempt to clarify in the recent PSC submittals that the original FSAR Class I cooling configuration is most closely analogous either to the
- " emergency cooling with EES using firewater pump (open loop)," as shown in Fig._ A-1 of Attachment 5 to P-86682, or to the " firewater flow diagram for Appendix R Train B case, open loop," as shown in Fig. 2.4 of. Attachment 2 to P-85055. Both of these configurations represent firewater supply via- the.
emergency condensate header as assumed in the Class I performance and safety analysis reported in Updated FSAR Sections 10.3.9 and 14.4.2.2.
In the configuration cited in Attachment 5 to P-86682, about 100 gpm of fluid is assumed to be vented to the atmosphere via an electromatic valve (PV-22162) located downstream of the steam generator EES, and the rest of the flow is vented to the condenser through the desuperheater and flash tank. Per FSAR .
Section 10.3.9, the original Class I event scenario assumed venting from ruptured non-Class I piping downstream of the steam bypass valves to the desuperheater. The licensee did not specify which of these configurations ;
presents the higher backpressure to the steam generator EES and thereby j produces the lower flow from the firewater supply system; however, use of the i newly installed and seismically qualified six inch vent lines off the EES f outlet piping to the atmosphere allows for lower EES backpressure. The ]
maximum fuel temperature calculated in the recent analysis (Attachment 5 to 4 P-86682) is 2511 F for a cooldown from 77.9% of rated reactor power. The firewater flow rate to the EES is calculated to be 829 gpa with an EES backpressure of 107 psia at the steam generator ring header and 87.3 psia at the mainstream bypass valve (PV-22129). Water exiting the EES is subcooled.
Similarly, in the configuration cited in Attachment 2 to P-85055, flow through the open loop (i.e., atmospheric venting via the newly installed six inch vent lines) lasts for only five hours, after which the loop is assumed to be closed in the Appendix R Train B Case (See Fig. 2.5 in Attachment 2 to P-85055). In the open loop configuration, the firewater flow is 996 gps, but is reduced to 789 gpa in the closed loop. The steam generator EES backpressure at the outlet is " controlled" to 76 psia (open loop) for the first five hours and j 97.8 psia (closed loop) thereafter; presumably, the quoted pressures are at the steam bypass valve. The maximum fuel temperature is calculated to be 2644 F following shutdown cooling from a power level of 87.5%.
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L MAFT l f 1 In addition, as reported in Attachment I to P-87053, the analysis of the !
firewster cooldown from 83.2% of rated reactor power without an interruption I in forced cooling indicated that the FSAR Section 14.4.2.1 values for maximum helium temperature (and thereby maximum fuel temperature) can be achieved with 795 gpm of 80 F firewater flow against a steam generator backpressure of 83.1 psia at the steam bypass valve. This result is analogous to that for the original Class I firewster cooldown scenario with no delay in restoring forced cooling, as assumed by PSC prior to the NRC's imposing the assumed 90 minute delay in 1978. As r.oted in Section 3.1 above, the firewater cooling capacity for this event should be the same as the current Class I firewater cooldown scenario with the 90 minute delay in restoring forced cooling. The major difference between the analysis in Attachment I to P-87053 and the current class.I event scenario would be core conditions and maximum helium temperature obtained after the 90 minute delay in restoring forced cooling of the core.
3.4 Findings and Recommendations The new Class I flow path, which was installed to meet the requirements for environmental qualifications of electrical equipment por 10CFR50.49, has been shown by calculations to provide a sufficient flow of firewster following shutdown from 87.5% of rated reactor power to drive one helium circulator at up to 3.8% of rated helium flow and to flood one EES section of the steam generator with subcooled flow. This flow of firewster is capable of flooding ,
the EES in an estimated 14 minutes prior to restart of forced cooling of the reactor core following a 90 minute interruption of forced cooling. This flow of firewater has been calculated to maintain the maximum temperature below 2900 F and to prevent lifting of the primary safety valves. However, the analysis of the new flow path has not adequately addressed single active failures due to a mechanical failure to open either of the two new valves in the new Class I flow path. Only electrical failures have been addressed.
Thus, the original Class I flow path must also be addressed for Safe Shutdown Cooling.
Although the original Class I Safe Shutdown Cooling scenario (i.e., 90 minute delay in restoring forced cooling via the emergency condensate supply header following the Design Basis Earthquake or Maximum Tornado) is not exactly duplicated in any of the new analyses, the results from similar analyses allow us to conclude that the original configuration for accommodating the Class I event can be utilized from 82% of rated reactor power without exceeding a maximum fuel temperature of 2900 F. PSC should perform an explicit confirmatory analysis of the Class I scenario to demonstrate that this is indeed the case for accommodator.g single active failures in the new Class I ,
flowpath. I Also, with regard to meeting the single failure criterion, we conclude that a confirmatory analysis is needed to demonstrate the capability of the alternate 10
a flow path via the emergency feedwater header. As shown in Figure 4-la, Attachment 2 to p-87055, the maximum fuel temperature is calculated to peak below 2900 F at about five hours into the firewater cooldown using the new Class I flow path. By extrapolation of the data in the cited figure, the maximum fuel temperature would still exceed 2200 F at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> into the firewater cooldown. This tendency to cool slowly is confirmed by the results given in Updated FSAR Figure 14.4-6 where these results are now recognized to be non conservative. Therefore, confirmatory analyses are needed to assure adequate flow and continued decrease in the fuel temperature after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> assuming a single passive failure.
In addition, since the failure to resolve issue.s raised initially by UERs 50-267/76-05 and 50-267/76-09 can apparently be traced to unreviewed or inadequately reviewed analyses of the firewater supply capability, assurance should be obtained that current analyses comply with the provisions for verification checking in design control per 10CFR50, Appendix B Section III and ANSI /ASME N45.2.11. Since design reviews are apparently the least ;
effective measure for verification checking, as evidenced by the findings of RO 50-267/86-026, the use of independent checks by alternate calculations and of checking by comparison to applicable plant test data should be emphasized.
As indicated in Section 3.2 above, NRC should perform an audit of the calculations performed by PSC and its subcontractors, particularly if design reviews have been the primary mechanism to achieve nominal compliance with 10CFR50, Appendix B and ANSI /ASME N45.2.11.
- 4. Conclusions Results of the accident analyses indicate that for both the "EQ" and
" Appendix R" scenarios. (which are both low probability events), there is substantial margin in the existing emergency cooling systems to provide for a safe shutdown. The analyses assumed that suitable procedures and training would be in place to assure that the operators would first implement the appropriate cooling water supply, and then manually control the primary coolant flow to avoid steaming and choking in the EES. We recommend that NRC confirms that suitable procedures and training are established.
PSC's analyses of the seismically and environmentally qualified firewater flowpath for Safe Shutdown Cooling indicate that adequate firewater flow can be obtained to avoid fuel damage and component damage following shutdown from 82% of rated reactor power and assuming a 90 min delay in restart of forced cooling of the reactor core. Pre-cooling of the steam generator EES section with firewater is calculated to take about 14 min to assure subcooled firewater flow at the EES outlet after forced cooling is restarted. Based on results of other recent analyses, single active failures in the new Class I flowpath are judged to be accommodated by using the original Class I firewater flow path to the emergency condensate header. Single passive failures, which are assumed not to occur until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the firewater cooldown is 11 L--__--------_-----.-- - - - - -
y
[4 l-(i' J
initiated, are to'be accommodated by an alternate flowpath through the i i
emergency feedwater header,'but specific analyses of this capability have not been presented by PSC. Based on our review of the PSC analyses and of. ;
previous reportable events related to the adequacy of the Class I firewster ,
cooldown capability, we recommend that the following confirmatory actions be i taken:
o- NRC should perfom an audit of the PSC calculations to confirm the adequacy of verification checking per 10CFR50, Appendix B.
Section III. This action is necessitated by findings of inadequate verification checking from previous reportable events that should-have been resolved by analysis of the firewater flow capability for Safe Shutdown Cooling.
o . PSC should perform an explicit confirmatory analysis to demonstrate that the original Class I-firewater flow path accommodates a single
' active failure in the new Class -I firewster flow path. The confirmatory analysis needs to address EES pre-cooling . times and long tern cooling capability, o PSC should perfom an explicit confirmatory analysis to demonstrate-that the alternate flow path via the emergency feedwater header accommodates a single passive failure in the preferred flow path via )
' the emergency condensate header. , j a
1, 1
I i
12
ATTACIDiENT Internal Correspondence MARTIN MARIETTA ENERGY SYSTEMS,INC.
March 25, 1987
[S.yJ. Ball Fire Water Cooldown Induced Water Hammer in Fort St. Vrain Steam Generators
Reference:
" Analysis of the Capability of the Fort St. Vrain Steam Generators to Withstand the Fire Water Cooldown Transient Following an Appendix 'R Fire," Rev.1, NLI-86-0649 As you requested, Jack Dixon and I reviewed the portion of the reference document regarding structural effects of water hammer. The report was qualitative and brief, but we agree with the following two assertions.
- 1. Water hammer caused by the collapse of a pocket of steam surrounded by incoming cold water is unlikely due to the steam generator design.
- 2. Water hammer forces in the steam generator tubes are significantly reduced by the pressure wave restriction at the entrance to the tubes.
However, we do not have enough information to comment on the water hammer-like forces produced by the cooling water sweeping through the steam generator nor on the effect on the generator of an external pressure wave being largely absorbed at the generator tube entrance.
We have experience in calculating structural response to water hammer loads including fluid entering a dry system and in performing confirmatory analyses of power reactor piping systems for the NRC. If the NRC would like to pursue this matter further, we will discuss schedules and cost with you.
If you have any questions, do not hesitate to contact me.
W N C. R. Hammond, 1000, MS-332, ORNL (4-6499) - NoRC cc: J. R. Dixon R. W. Glass T. W. Pickel W. C. Stoddart File - CRH jNb4
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